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Sample records for coolant accidents bruchmechanische

  1. Loss-of-coolant accident analysis of the Savannah River new production reactor design

    SciTech Connect

    Maloney, K.J.; Pryor, R.J.

    1990-11-01

    This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs.

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

    SciTech Connect

    Su'ud, Zaki; Anshari, Rio

    2012-06-06

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

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

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Anshari, Rio

    2012-06-01

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

  4. Cladding embrittlement during postulated loss-of-coolant accidents.

    SciTech Connect

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  5. Analysis of Loss-of-Coolant Accidents in the NBSR

    SciTech Connect

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

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  6. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Yoder, G.L. ); Wendel, M.W. )

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs.

  7. Comparison of passive safety and the safety injection systems under loss of coolant accident

    NASA Astrophysics Data System (ADS)

    Tahir, M.; Chughtai, I. R.; Lodhi, M. A. K.

    2009-04-01

    A Passive Safety Injection System (PSIS) and a Safety Injection System (SIS) with reference to a typical pressurized water reactor have been studied. The performance of the PSIS has been analyzed for a large break Loss of Coolant Accident (LOCA) in one of the cold leg of reactor coolant system. The SIS is a huge system consisting of many active components needing electrical power to perform its role of core cooling as high head safety injection system under designed accidents. The PSIS consist of passive components and performs its function automatically under gravity. In a reactor transient simulation, the PSIS and the SIS are tested for large break LOCA under the same boundary conditions. Critical thermal hydraulic parameters of both the systems are presented. Results obtained are approximately similar in both cases. Nevertheless, the PSIS would be a better choice for handling such scenarios due to its reduced and passive components.

  8. JAEA Studies on High Burnup Fuel Behaviors during Reactivity-Initiated Accident and Loss-of-Coolant Accident

    SciTech Connect

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nagase, Fumihisa; Suzuki, Motoe

    2007-07-01

    The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of reactivity-initiated accident (RIA) studies including pulse-irradiation experiments in the NSRR and cladding mechanical tests, loss-of-coolant accident (LOCA) tests including integral thermal shock test and oxidation rate measurement, development and verification of computer codes FEMAXI-6 and RANNS, and so on. In addition to an overview of the fuel safety research at JAEA, most recent progresses in the RIA and LOCA tests programs and the codes development are described and discussed in the paper. (authors)

  9. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    SciTech Connect

    Al-Falahi, A.; Haennine, M.; Porkholm, K.

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  10. The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents

    SciTech Connect

    Terrani, Kurt A.; Wang, Dean; Ott, Larry J.; Montgomery, Robert O.

    2014-05-01

    The effect of variation in thermal conductivity of light water reactor fuel elements on core response during loss-of-coolant accident scenarios is examined. Initially, a simplified numerical analysis is utilized to determine the time scales associated with dissipation of stored energy from the fuel into the coolant once the fission reaction is stopped. The analysis is then followed by full reactor system thermal-hydraulics analysis of a typical boiling and pressurized water reactor subjected to a large break loss-of-coolant accident scenario using the TRACE code. Accordingly, sensitivity analyses to examine the effect of an increase in fuel thermal conductivity, up to 500%, on fuel temperature evolution during these transients are performed. Given the major differences in thermal-hydraulics design aspects of boiling and pressurized water reactors, different fuel and temperature responses during the simulated loss-of-coolant transients are observed.

  11. Large break loss of coolant severe accident sequences at the HFIR (High Flux Isotope Reactor)

    SciTech Connect

    Simpson, D.B.; Greene, S.R.

    1990-01-01

    An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs.

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

    SciTech Connect

    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, 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.

  13. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    SciTech Connect

    Heames, T.J. ); Williams, D.A.; Johns, N.A.; Chown, N.M. ); Bixler, N.E.; Grimley, A.J. ); Wheatley, C.J. )

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  14. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    SciTech Connect

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-07-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO{sub 2} volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  15. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    SciTech Connect

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  16. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    SciTech Connect

    Kao, S.P.; Chang, S.K.; Huang, H.C.

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

  17. Cobalt-60 simulation of LOCA (loss of coolant accident) radiation effects

    SciTech Connect

    Buckalew, W.H.

    1989-07-01

    The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs.

  18. An investigation of core liquid level depression in small break loss-of-coolant accidents

    SciTech Connect

    Schultz, R.R.; Watkins, J.C. ); Motley, F.E.; Stumpf, H. ); Chen, Y.S. . Div. of Systems Research)

    1991-08-01

    Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs.

  19. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  20. Feedwater transient and small break loss of coolant accident analyses for the Bellefonte Nuclear Plant

    SciTech Connect

    Bayless, P D; Dobbe, C A; Chambers, R

    1987-03-01

    Specific sequences that may lead to core damage were analyzed for the Bellefonte nuclear plant as part of the US Nuclear Regulatory Commission's Severe Accident Sequence Analysis Program. The RELAP5, SCDAP, and SCDAP/RELAP5 computer codes were used in the analyses. The two main initiating events investigated were a loss of all feedwater to the steam generators and a small cold leg break loss of coolant accident. The transients of primary interest within these categories were the TMLB' and S/sub 2/D sequences. Variations on systems availability were also investigated. Possible operator actions that could prevent or delay core damage were identified, and two were investigated for a small break transient. All of the transients were analyzed until either core damage began or long-term decay heat removal was established. The analyses showed that for the sequences considered the injection flow from one high-pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were able to prevent core damage in the S/sub 2/D sequence; no operator actions were available to prevent core damage in the TMLB' sequence.

  1. Analysis of an AP600 intermediate-size loss-of-coolant accident

    SciTech Connect

    Boyack, B.E.; Lime, J.F.

    1995-09-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  2. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  3. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    SciTech Connect

    Nelson, C.F.

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  4. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    SciTech Connect

    Nelson, C.F.; Gauthier, G.; Carlin, F.

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40{degrees}C or 70{degrees}C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased.

  5. Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

    SciTech Connect

    Revankar, S.T.; Xu, Y.; Yoon, H.J.; Ishii, M.

    2002-07-01

    The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance. (authors)

  6. 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

  7. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    SciTech Connect

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  8. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    SciTech Connect

    Heams, T J; Williams, D A; Johns, N A; Mason, A; Bixler, N E; Grimley, A J; Wheatley, C J; Dickson, L W; Osborn-Lee, I; Domagala, P; Zawadzki, S; Rest, J; Alexander, C A; Lee, R Y

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  9. Dose to man from a hypothetical loss-of-coolant accident at the Rancho Seco Nuclear Power Plant

    SciTech Connect

    Peterson, K.R.; Greenly, G.D.

    1981-02-01

    At the request of the Sacramento Municipal Utilities District, we used our computer codes, MATHEW and ADPIC, to assess the environmental impact of a loss-of-coolant accident at the Rancho Seco Nuclear Power Plant, about 40 kilometres southeast of Sacramento, California. Meteorological input was selected so that the effluent released by the accident would be transported over the Sacramento metropolitan area. With the release rates provided by the Sacramento Municipal Utilities District, we calculated the largest total dose for a 24-hour release as 70 rem about one kilometre northwest of the reactor. The largest total dose in the Sacramento metropolitan area is 780 millirem. Both doses are from iodine-131, via the forage-cow-milk pathway to an infant's thyroid. The largest dose near the nuclear plant can be minimized by replacing contaminated milk and by giving the cows dry feed. To our knowledge, there are no milk cows within the Sacramento metropolitan area.

  10. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    SciTech Connect

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

    1995-09-01

    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 of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  11. 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.

  12. Experimental investigation on the chemical precipitation generation under the loss of coolant accident of nuclear power plants

    SciTech Connect

    Kim, C. H.; Sung, J. J.; Chung, Y. W.

    2012-07-01

    The PWR containment buildings are designed to facilitate core cooling in the event of a Loss of Coolant Accident (LOCA). The cooling process requires water discharged from the break and containment spray to be collected in a sump for recirculation. The containment sump contains screens to protect the components of the Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) from debris. Since the containment materials may dissolve or corrode when exposed to the reactor coolant and spray solutions, various chemical precipitations can be generated in a post-LOCA environment. These chemical precipitations may become another source of debris loading to be considered in sump screen performance and downstream effects. In this study, new experimental methodology to predict the type and quantity of chemical precipitations has been developed. To generate the plant-specific chemical precipitation in a post-LOCA environment, the plant specific chemical condition of the recirculation sump during post-LOCA is simulated with the experimental reactor for the chemical effect. The plant-specific containment materials are used in the present experiment such as glass fibers, concrete blocks, aluminum specimens, and chemical reagent - boric acid, spray additives or buffering chemicals (sodium hydroxide, Tri-Sodium Phosphate (TSP), or others). The inside temperature of the reactor is controlled to simulate the plant-specific temperature profile of the recirculation sump. The total amount of aluminum released from aluminum specimens is evaluated by ICP-AES analysis to determine the amount of AlOOH and NaAlSi{sub 3}O{sub 8} which induce very adverse effect on the head loss across the sump screens. The amount of these precipitations generated in the present experimental study is compared with the results of WCAP-16530-NP-A. (authors)

  13. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    SciTech Connect

    Ruggles, A. E.; Cheng, L. Y.; Dimenna, R. A.; Griffith, P.; Wilson, G. E.

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis.

  14. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    SciTech Connect

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  15. Assessment of a large break loss of coolant accident scenario requiring operator action to initiate safety injection

    SciTech Connect

    Grendys, R.C.; Nissley, M.E.; Baker, D.C.

    1996-11-01

    As part of the licensing basis for a nuclear power plant, the acceptability of the Emergency Core Cooling Systems (ECCS) following a postulated Loss-of-Coolant Accident (LOCA) as described in the Code of Federal Regulations (CFR), Title 10, Chapter 1, Part 50.46, must be verified. The LOCA analysis is performed with an acceptable ECCS Evaluation Model and results must show compliance with the 10 CFR 50.46 acceptance criteria. Westinghouse Electric Corporation performs Large and Small Break LOCA and LOCA-related analyses to support the licensing basis of various nuclear power plants and also performs evaluations against the licensing basis analyses as required. Occasionally, the need arises for the holder of an operating license of a nuclear power plant to submit a Licensee Event Report (LER) to the US Nuclear Regulatory Commission (USNRC) for any event of the type described in the Code of Federal Regulations, Title 10, Chapter 1, Part 50.73. To support the LER, a Justification for Past Operation (JPO) may be performed to assess the safety consequences and implications of the event based on previous operating conditions. This paper describes the work performed for the Large Break LOCA to assess the impact of an event discovered by Florida Power and Light and reported in LER-94-005-02. For this event, it was determined that under certain circumstances, operator action would have been required to initiate safety injection (SI), thus challenging the acceptability of the ECCS. This event was specifically addressed for the Large Break LOCA by using an advanced thermal hydraulic analysis methodology with realistic input assumptions.

  16. Neutron Imaging Investigations of the Secondary Hydriding of Nuclear Fuel Cladding Alloys during Loss of Coolant Accidents

    NASA Astrophysics Data System (ADS)

    Grosse, M.; Roessger, C.; Stuckert, J.; Steinbrueck, M.; Kaestner, A.; Kardjilov, N.; Schillinger, B.

    The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrogen and zirconium was determined by testing calibration specimens with known hydrogen concentrations. Hydrogen enrichments located at the end of the ballooning zone of the tested tubes were detected in the inner rods of the test bundles. Nearly all of the peripheral claddings exposed to lower temperatures do not show such enrichments. This implies that under the conditions investigated a threshold temperature exists below which no hydrogen enrichments can be formed. In order to understand the hydrogen distribution a model was developed describing the processes occurring during loss of coolant accidents after rod burst. The general shape of the hydrogen distributions with a peak each side of the ballooning region is well predicted by this model whereas the absolute concentrations are underestimated compared to the results of the neutron tomography investigations. The model was also used to discuss the influence of the alloy composition on the secondary hydrogenation. Whereas the relations for the maximal hydrogen concentrations agree well for one and the same alloy, the agreement for tests with different alloys is less satisfying, showing that material parameters such as oxidation kinetics, phase transition temperature for the zirconium oxide, and yield strength and ductility at high temperature have to be taken into account to reproduce the results of neutron imaging investigations correctly.

  17. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    SciTech Connect

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio.

  18. Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors

    SciTech Connect

    Not Available

    1993-07-01

    An interfacing systems loss of coolant accident (ISLOCA) involves failure or improper operation of pressure isolation valves (PIVs) that compose the boundary between the reactor coolant system and low-pressure rated systems. Some ISLOCAs can bypass containment and result in direct release of fission products to the environment. A cost/benefit evaluation, using three PWR analyses, calculated the benefit of two potential modifications to the plants. Alternative 1 is improved plant operations to optimize the operator`s performance and reduce human error probabilities. Alternative 2 adds pressure sensing devices, cabling, and instrumentation between two PIVs to provide operators with continuous monitoring of the first PIV. These two alternatives were evaluated for the base case plants (Case 1) and for each plant, assuming the plants had a particular auxiliary building design in which severe flooding would be a problem if an ISLOCA occurred. The auxiliary building design (Case 2) was selected from a survey that revealed a number of designs with features that provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The results were judged not to provide sufficient basis for generic requirements. It was concluded that the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs).

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

    SciTech Connect

    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.

  20. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    SciTech Connect

    Ghan, L.S.; Ortiz, M.G.

    1991-12-31

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B&W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission`s (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B&W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions.

  1. 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

    ... accident. II. Further Information DG-1234 was published in the Federal Register on July 15 2010, (75 FR... atmosphere cleanup systems. RG 1.82 provides guidelines for evaluating the adequacy and the availability of... containment atmosphere cleanup systems. RG 1.82 provides guidelines for evaluating the adequacy and...

  2. Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients

    SciTech Connect

    Ratnayake, Ruwan K.; Ergun, S.; Hochreiter, L.E.; Baratta, A.J.

    2002-07-01

    In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and 5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena. (authors)

  3. 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.

    SciTech Connect

    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 provided 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.

  4. RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

    SciTech Connect

    Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

    1990-05-01

    A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

  5. A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX thermal hydraulic testing facility.

    SciTech Connect

    Wachs, D. M.

    1998-11-04

    Thermal stratification, which has been linked to the occurrence of pressurized thermal shock (PTS), is observed to occur during the early stages of simulated loss of coolant accidents (LOCAS) in the Oregon State University Advanced Plant Experiment (OSU APEX) Thermal Hydraulic Test Facility. The OSU APEX Test Facility is a scaled model of the Westinghouse AP600 nuclear power plant. Analysis of the OSU APEX facility data has allowed the determination of an onset criteria for thermal stratification and has provided support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer that described the phenomena occurring within them. Some mixing phenomena were predicted that lead to non-uniformity between the two cold legs attached to the steam generator on the side of the facility containing the Passive Residual Heat Removal (PRHR) injection system. The stratification was found to be two phase and unlikely to be a factor in PTS.

  6. Coolant injector

    SciTech Connect

    Stevenson, D.L.; Seleno, F.M.

    1984-04-03

    Apparatus for injecting coolant into the fuel-air mixture supplied to a turbocharged internal combustion engine in which coolant is stored in a reservoir that is connected to the outlet side of a turbocharger compressor and is pressurized during times that the turbocharger is generating positive pressures above a predetermined level. The pressure in the reservoir forces coolant to the inlet side of the turbocharger to cool fuel and injection is interrupted when the pressures are below that level and cooling of the fuel-air mixture is not required.

  7. 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.

  8. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 4: External Pressurizer Surge Line Break Near Inlet Header

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  9. APT Blanket System Loss-of-Coolant Accident (LOCA) Analysis Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

    SciTech Connect

    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 (HR) system. These simulations were performed for the Preliminary Safety Analysis Report.

  10. APT Blanket System Loss-of-Coolant Accident Based on Initial Conceptual Design - Case 5: External RHR Break Near Inlet Header

    SciTech Connect

    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.

  11. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    NASA Astrophysics Data System (ADS)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  12. Coolant line hydrometer

    SciTech Connect

    Barber, M.D.; Kipp, W.G.

    1987-03-17

    This patent describes a hydrometer unit for connection in an automobile coolant flow line comprising: a tubular fitting adapted to be connected to the coolant flow line; a coolant receiving chamber means connected to the tubular fitting for receiving coolant from the tubular fitting; and indicating float elements contained within the coolant receiving chamber means and adapted to rise therein individually as a function of the specific gravity of the coolant. The coolant receiving chamber means includes a closure cap which when connected to the tubular fitting forms a coolant receiving chamber, retaining means for retaining the indicating float elements within the coolant receiving chamber, a viewing window member of a substantially clear material through which the float elements can be visually observed within the coolant receiving chamber means, and air venturi means located within the coolant receiving chamber means for automatically removing air which may collect within the coolant chamber means.

  13. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  14. Modeling Reactor Coolant Systems Thermal-Hydraulic Transients

    1999-10-05

    RELAP5/MOD3.2* is used to model reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transients without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal-hydraulic systems. Control system and secondary system components are included to allow modeling of themore » plant controls, turbines, condensers, and secondary feedwater systems.« less

  15. Machine coolant waste reduction by optimizing coolant life. Project summary

    SciTech Connect

    Pallansch, J.

    1995-08-01

    The project was designed to study the following: A specific water-soluble coolant (Blasocut 2000 Universal) in use with a variety of machines, tools, and materials; Coolant maintenance practices associated with three types of machines; Health effects of use and handling of recycled coolant; Handling practices for chips and waste coolant; Chip/coolant separation; and Oil/water separation.

  16. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolytic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 5% of beryllium or magnesium dispersed in the hydrocarbon.

  17. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolitic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 10% of an alkall metal dispersed in the hydrocarbon.

  18. MACHINE COOLANT WASTE REDUCTION BY OPTIMIZING COOLANT LIFE

    EPA Science Inventory

    Machine shops use coolants to improve the life and function of machine tools. hese coolants become contaminated with oils with use, and this contamination can lead to growth of anaerobic bacteria and shortened coolant life. his project investigated methods to extend coolant life ...

  19. The IRIS Spool-Type Reactor Coolant Pump

    SciTech Connect

    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 safety 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)

  20. Environmentally Friendly Coolant System

    SciTech Connect

    David Jackson Principal Investigator

    2011-11-08

    Energy reduction through the use of the EFCS is most improved by increasing machining productivity. Throughout testing, nearly all machining operations demonstrated less land wear on the tooling when using the EFCS which results in increased tool life. These increases in tool life advance into increased productivity. Increasing productivity reduces cycle times and therefore reduces energy consumption. The average energy savings by using the EFCS in these machining operations with these materials is 9%. The advantage for end milling stays with flood coolant by about 6.6% due to its use of a low pressure pump. Face milling and drilling are both about 17.5% less energy consumption with the EFCS than flood coolant. One additional result of using the EFCS is improved surface finish. Certain machining operations using the EFCS result in a smoother surface finish. Applications where finishing operations are required will be able to take advantage of the improved finish by reducing the time or possibly eliminating completely one or more finishing steps and thereby reduce their energy consumption. Some machining operations on specific materials do not show advantages for the EFCS when compared to flood coolants. More information about these processes will be presented later in the report. A key point to remember though, is that even with equivalent results, the EFCS is replacing petroleum based coolants whose production produces GHG emissions and create unsafe work environments.

  1. Reactor coolant pump flywheel

    SciTech Connect

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  2. F. E. L. coolant

    SciTech Connect

    Schumann, D.; Lepper, J.

    1987-11-20

    The compatibility of a number of organic and inorganic materials with Freon-113 in ETA-II is studied. The stability of Freon-113 in the simulated electrical environment is also discussed. It would be desirable to use Freon-113 as the dielectric coolant because it is a factor of twenty less expensive than the currently used Flourinert-75. No substantial problems were observed with any of the materials of interest. Polycarbonate will craze under certain conditions of exposure to Freon-113, mechanical stress, temperature and time. Extruded polymethylmethacrylate crazes on exposure to Freon-113. This is assumed to result from residual stress or processing aids because the cast polymer shows no effects. The polysiloxane swelled severely in the Freon-113 which is no surprise since solubility parameters are close. The Freon-113 did not break down under electrical arcing which simulated service conditions to produce HF or HCl within the detection limits of a Karl Fisher titration even with substantial added water.

  3. Treatment of mixed waste coolant

    SciTech Connect

    Kidd, S.; Bowers, J.S.

    1995-02-01

    The primary processes used at Lawrence Livermore National Laboratory (LLNL) for treatment of radioactively contaminated machine coolants are industrial waste treatment and in situ carbon adsorption. These two processes simplify approaches to meeting the sanitary sewer discharge limits and subsequent Land Disposal Restriction criteria for hazardous and mixed wastes (40 CFR 268). Several relatively simple technologies are used in industrial water treatment. These technologies are considered Best Demonstrated Available Technologies, or BDAT, by the Environmental Protection Agency. The machine coolants are primarily aqueous and contain water soluble oil consisting of ethanol amine emulsifiers derived from fatty acids, both synthetic and natural. This emulsion carries away metal turnings from a part being machined on a lathe or other machining tool. When the coolant becomes spent, it contains chlorosolvents carried over from other cutting operations as well as a fair amount of tramp oil from machine bearings. This results in a multiphasic aqueous waste that requires treatment of metal and organic contaminants. During treatment, any dissolved metals are oxidized with hydrogen peroxide. Once oxidized, these metals are flocculated with ferric sulfate and precipitated with sodium hydroxide, and then the precipitate is filtered through diatomaceous earth. The emulsion is broken up by acidifying the coolant. Solvents and oils are adsorbed using powdered carbon. This carbon is easily separated from the remaining coolant by vacuum filtration.

  4. 1996 Coolant Flow Management Workshop

    NASA Technical Reports Server (NTRS)

    Hippensteele, Steven A. (Editor)

    1997-01-01

    The following compilation of documents includes a list of the 66 attendees, a copy of the viewgraphs presented, and a summary of the discussions held after each session at the 1996 Coolant Flow Management Workshop held at the Ohio Aerospace Institute, adjacent to the NASA Lewis Research Center, Cleveland, Ohio on December 12-13, 1996. The workshop was organized by H. Joseph Gladden and Steven A. Hippensteele of NASA Lewis Research Center. Participants in this workshop included Coolant Flow Management team members from NASA Lewis, their support service contractors, the turbine engine companies, and the universities. The participants were involved with research projects, contracts and grants relating to: (1) details of turbine internal passages, (2) computational film cooling capabilities, and (3) the effects of heat transfer on both sides. The purpose of the workshop was to assemble the team members, along with others who work in gas turbine cooling research, to discuss needed research and recommend approaches that can be incorporated into the Center's Coolant Flow Management program. The workshop was divided into three sessions: (1) Internal Coolant Passage Presentations, (2) Film Cooling Presentations, and (3) Coolant Flow Integration and Optimization. Following each session there was a group discussion period.

  5. Long life coolant pump technology

    NASA Technical Reports Server (NTRS)

    1976-01-01

    Design concepts were investigated to improve space system coolant pump technology to be suitable for mission durations of two years and greater. These design concepts included an improved bearing system for the pump rotating elements, consisting of pressurized conical bearings. This design was satisfactorily endurance tested as was a new prototype pump built using various other improved design concepts. Based upon an overall assessment of the results of the program it is concluded that reliable coolant pumps can be designed for three year space missions.

  6. Chemical factors affecting fission product transport in severe LMFBR accidents

    SciTech Connect

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly.

  7. Polonium release from an ATW burner system with liquid lead-bismuth coolant

    SciTech Connect

    Li, N.; Yefimov, E.; Pankratov, D.

    1998-04-01

    The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods.

  8. INHIBITING THE POLYMERIZATION OF NUCLEAR COOLANTS

    DOEpatents

    Colichman, E.L.

    1959-10-20

    >The formation of new reactor coolants which contain an additive tbat suppresses polymerization of the primary dissoclation free radical products of the pyrolytic and radiation decomposition of the organic coolants is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to 5% of a powdered metal hydride chosen from the group consisting of the group IIA and IVA dispersed in the hydrocarbon.

  9. Effects of coolant volatility on simulated HCDA bubble expansions. Technical report No. 10

    SciTech Connect

    Tobin, R.J.

    1980-09-01

    The effects of coolant volatility on the expansion dynamics and cover loading of hypothetical core disruptive accidents (HCDA) were studied by performing experiments with a transparent 1/30-scale model of a typical demonstration size loop-type liquid metal fast breeder reactor (LMFBR). Freon 113 and Freon 11 were used as coolant simulants of increasing volatility. High-pressure nitrogen gas (1450 psia) or flashing water (1160 psia) were used to simulate the qualitative features of sodium vapor or molten fuel expansions. To validate the use of constant mass and constant geometry experiments as a means of evaluating the effects of coolant volatility, a set of baseline experiments was performed in these configurations with the room temperature nitrogen bubble source. In all the experiments, the expanding HCDA bubbles, the motion of the coolant simulant, and the vessel loads were monitored by pressure transducers, a thermocouple in the bubble, and high-speed photography. Results of the constant mass experiments with the flashing water source show that higher volatility results in higher pressure driving the coolant slug and therefore higher impact loads. The Freon experiments had about 50% higher pressure in the upper core and bubble, a 30% larger slug impact impulse, and 25% greater expansion work done on the coolant slug. The higher pressure in the Freon experiments is believed due to vaporization of some of the Freon that mixes with the hot flashing water in the upper core very early in the expansion. Entrainment of coolant within the bubble and the bubble shape were comparable in the Freon and water experiments. Entrainment at slug impact varied between 20 and 40% of the bubble volume. The presence of internal vessel structures attenuated the slug impact impulse by about 50%, whether the coolant was Freon 113 or water. 79 figures, 23 tables.

  10. Cleaning of uranium vs machine coolant formulations

    SciTech Connect

    Cristy, S.S.; Byrd, V.R.; Simandl, R.F.

    1984-10-01

    This study compares methods for cleaning uranium chips and the residues left on chips from alternate machine coolants based on propylene glycol-water mixtures with either borax, ammonium tetraborate, or triethanolamine tetraborate added as a nuclear poison. Residues left on uranium surfaces machined with perchloroethylene-mineral oil coolant and on surfaces machined with the borax-containing alternate coolant were also compared. In comparing machined surfaces, greater chlorine contamination was found on the surface of the perchloroethylene-mineral oil machined surfaces, but slightly greater oxidation was found on the surfaces machined with the alternate borax-containing coolant. Overall, the differences were small and a change to the alternate coolant does not appear to constitute a significant threat to the integrity of machined uranium parts.

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

    SciTech Connect

    Crosswait, K.M.

    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 as 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.

  12. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    SciTech Connect

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  13. Influence of coolant injector configuration on film cooling effectiveness for gaseous and liquid film coolants

    NASA Astrophysics Data System (ADS)

    Shine, S. R.; Sunil Kumar, S.; Suresh, B. N.

    2012-05-01

    An experimental investigation is conducted to bring out the effects of coolant injector configuration on film cooling effectiveness, film cooled length and film uniformity associated with gaseous and liquid coolants. A series of measurements are performed using hot air as the core gas and gaseous nitrogen and water as the film coolants in a cylindrical test section simulating a thrust chamber. Straight and compound angle injection at two different configurations of 30°-10° and 45°-10° are investigated for the gaseous coolant. Tangential injection at 30° and compound angle injection at 30°-10° are examined for the liquid coolant. The analysis is based on measurements of the film-cooling effectiveness and film uniformity downstream of the injection location at different blowing ratios. Measured results showed that compound angle configuration leads to lower far-field effectiveness and shorter film length compared to tangential injection in the case of liquid film cooling. For similar injector configurations, effectiveness along the stream wise direction showed flat characteristics initially for the liquid coolant, while it was continuously dropping for the gaseous coolant. For liquid coolant, deviations in temperature around the circumference are very low near the injection point, but increases to higher values for regions away from the coolant injection locations. The study brings out the existance of an optimum gaseous film coolant injector configuration for which the effectiveness is maximum.

  14. Loss of coolant analysis for the tower shielding reactor 2

    SciTech Connect

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs.

  15. Coolant mass flow equalizer for nuclear fuel

    DOEpatents

    Betten, Paul R.

    1978-01-01

    The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

  16. Bi-coolant flat plate solar collector

    NASA Astrophysics Data System (ADS)

    Chon, W. Y.; Green, L. L.

    The feasibility study of a flat plate solar collector which heats air and water concurrently or separately was carried out. Air flows above the collector absorber plate, while water flows in tubes soldered or brazed beneath the plate. The collector efficiencies computed for the flow of both air and water are compared with those for the flow of a single coolant. The results show that the bi-coolant collector efficiency computed for the entire year in Buffalo, New York is higher than the single-coolant collector efficiency, although the efficiency of the water collector is higher during the warmer months.

  17. 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.

  18. Radiotherapy Accidents

    NASA Astrophysics Data System (ADS)

    Mckenzie, Alan

    A major benefit of a Quality Assurance system in a radiotherapy centre is that it reduces the likelihood of an accident. For over 20 years I have been the interface in the UK between the Institute of Physics and Engineering in Medicine and the media — newspapers, radio and TV — and so I have learned about radiotherapy accidents from personal experience. In some cases, these accidents did not become public and so the hospital cannot be identified. Nevertheless, lessons are still being learned.

  19. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... without rust inhibitors. (c) For coolants allowed in paragraphs (a) and (b) of this section, you may use rust inhibitors and additives required for lubricity, up to the levels that the additive...

  20. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... without rust inhibitors. (c) For coolants allowed in paragraphs (a) and (b) of this section, you may use rust inhibitors and additives required for lubricity, up to the levels that the additive...

  1. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... without rust inhibitors. (c) For coolants allowed in paragraphs (a) and (b) of this section, you may use rust inhibitors and additives required for lubricity, up to the levels that the additive...

  2. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... without rust inhibitors. (c) For coolants allowed in paragraphs (a) and (b) of this section, you may use rust inhibitors and additives required for lubricity, up to the levels that the additive...

  3. Worldwide trends in engine coolants, cooling system materials and testing

    SciTech Connect

    Not Available

    1990-01-01

    This book contains the proceedings on worldwide trends in engine coolants, cooling systems, materials and testings. Topics covered include: Internationalization of the Automotive Industry - Global Responses in the Functional Fluid Area; Analysis of Coolants from Diesel Engines; Cavitation Damage of Diesel Engine Wet- Cylinder Liners; and Development of Test Stand to Measure the Effect of Coolant Composition on Engine Coolant Pump Seal Leakage.

  4. Lead Coolant Test Facility Development Workshop

    SciTech Connect

    Paul A. Demkowicz

    2005-06-01

    A workshop was held at the Idaho National Laboratory on May 25, 2005, to discuss the development of a next generation lead or lead-alloy coolant test facility. Attendees included representatives from the Generation IV lead-cooled fast reactor (LFR) program, Advanced Fuel Cycle Initiative, and several universities. Several participants gave presentations on coolant technology, existing experimental facilities for lead and lead-alloy research, the current LFR design concept, and a design by Argonne National Laboratory for an integral heavy liquid metal test facility. Discussions were focused on the critical research and development requirements for deployment of an LFR demonstration test reactor, the experimental scope of the proposed coolant test facility, a review of the Argonne National Laboratory test facility design, and a brief assessment of the necessary path forward and schedule for the initial stages of this development project. This report provides a summary of the presentations and roundtable discussions.

  5. Full-length fuel rod behavior under severe accident conditions

    SciTech Connect

    Lombardo, N J; Lanning, D D; Panisko, F E

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

  6. Coolant monitoring apparatus for nuclear reactors

    DOEpatents

    Tokarz, Richard D.

    1983-01-01

    A system for monitoring coolant conditions within a pressurized vessel. A length of tubing extends outward from the vessel from an open end containing a first line restriction at the location to be monitored. The flowing fluid is cooled and condensed before passing through a second line restriction. Measurement of pressure drop at the second line restriction gives an indication of fluid condition at the first line restriction. Multiple lengths of tubing with open ends at incremental elevations can measure coolant level within the vessel.

  7. On-Line Coolant Chemistry Analysis

    SciTech Connect

    LM Bachman

    2006-07-19

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level.

  8. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  9. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Engine Fluids, Test Fuels, Analytical Gases and Other Calibration Standards § 1065.745 Coolants... your engine in use. (b) For laboratory testing of liquid-cooled engines, you may use water with...

  10. Design criteria for Waste Coolant Processing Facility and preliminary proposal 722 for Waste Coolant Processing Facility

    SciTech Connect

    Not Available

    1991-09-27

    This document contains the design criteria to be used by the architect-engineer (A-E) in the performance of Titles 1 and 2 design for the construction of a facility to treat the biodegradable, water soluble, waste machine coolant generated at the Y-12 plant. The purpose of this facility is to reduce the organic loading of coolants prior to final treatment at the proposed West Tank Farm Treatment Facility.

  11. Severe accident sequence assessment for boiling water reactors: program overview

    SciTech Connect

    Fontana, M. H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case.

  12. MELCOR accident analysis for ARIES-ACT

    SciTech Connect

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

    We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

  13. 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.

  14. Coolant passage heat transfer with rotation

    NASA Astrophysics Data System (ADS)

    Hajek, T. J.; Wagner, J.; Johnson, B. V.

    1986-10-01

    In current and advanced gas turbine engines, increased speeds, pressures and temperatures are used to reduce specific fuel consumption and increase thrust/weight ratios. Hence, the turbine airfoils are subjected to increased heat loads escalating the cooling requirements to satisfy life goals. The efficient use of cooling air requires that the details of local geometry and flow conditions be adequately modeled to predict local heat loads and the corresponding heat transfer coefficients. The objective of this program is to develop a heat transfer and pressure drop data base, computational fluid dynamic techniques and correlations for multi-pass rotating coolant passages with and without flow turbulators. The experimental effort is focused on the simulation of configurations and conditions expected in the blades of advanced aircraft high pressure turbines. With the use of this data base, the effects of Coriolis and buoyancy forces on the coolant side flow can be included in the design of turbine blades.

  15. Evaluation of Salt Coolants for Reactor Applications

    SciTech Connect

    Williams, David F

    2008-01-01

    Molten fluorides were initially developed for use in the nuclear industry as the high-temperature fluid fuel for the Molten Salt Reactor (MSR). The U.S. Department of Energy Office of Nuclear Energy is exploring the use of molten salts as primary and secondary coolants in a new generation of solid-fueled, thermal-spectrum, hightemperature reactors. This paper provides a review of relevant properties for use in evaluation and ranking of salt coolants for high-temperature reactors. Nuclear, physical, and chemical properties were reviewed, and metrics for evaluation are recommended. Chemical properties of the salt were examined to identify factors that affect materials compatibility (i.e., corrosion). Some preliminary consideration of economic factors for the candidate salts is also presented.

  16. Radiation accidents

    SciTech Connect

    Saenger, E.L.

    1986-09-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity.

  17. Radiation accidents.

    PubMed

    Saenger, E L

    1986-09-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity. PMID:3526994

  18. Evaluation of engine coolant recycling processes: Part 2

    SciTech Connect

    Bradley, W.H.

    1999-08-01

    Engine coolant recycling continues to provide solutions to both economic and environmental challenges often faced with the disposal of used engine coolant. General Motors` Service Technology Group (STG), in a continuing effort to validate the general practice of recycling engine coolants, has conducted an in-depth study on the capabilities of recycled coolants. Various recycling processes ranging from complex forms of fractional distillation to simple filtration were evaluated in this study to best represent the current state of coolant recycling technology. This study incorporates both lab and (limited) fleet testing to determine the performance capabilities of the recycled coolants tested. While the results suggest the need for additional studies in this area, they reveal the true capabilities of all types of engine coolant recycling technologies.

  19. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J; Wilson, C L

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  20. Porous coolant tube holder for fuel cell stack

    DOEpatents

    Guthrie, Robin J.

    1981-01-01

    A coolant tube holder for a stack of fuel cells is a gas porous sheet of fibrous material adapted to be sandwiched between a cell electrode and a nonporous, gas impervious flat plate which separates adjacent cells. The porous holder has channels in one surface with coolant tubes disposed therein for carrying coolant through the stack. The gas impervious plate is preferably bonded to the opposite surface of the holder, and the channel depth is the full thickness of the holder.

  1. CFD analyses of coolant channel flowfields

    NASA Technical Reports Server (NTRS)

    Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

    1993-01-01

    The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

  2. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Liquid Cooling § 23.1063... specimen liner must be conducted with the coolant at operating temperature. Induction System...

  3. 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.

  4. Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    SciTech Connect

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.; Xiang, J.Y.

    1995-12-31

    Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment.

  5. Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    SciTech Connect

    Casino, William A. Jr.

    2006-07-01

    The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

  6. Advanced sodium fast reactor accident source terms :

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  7. Advanced light water reactor requirements document: Chapter 3, Reactor coolant system and reactor non-safety auxiliary systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Coolant System and the Reactor Non-safety Auxiliary Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the reactor coolant system and reactor non-safety auxiliary systems for Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Non-safety auxiliaries include systems which are required for normal operation of the plant but are not required to operate for accident mitigation or to bring the plant to a safe shutdown condition. For PWRs, the reactor coolant system, steam generator system, chemical and volume control system and boron recycle system are included. For BWRs, the reactor coolant system and reactor water cleanup system are included. The chapter also includes requirements for the above systems which are common to BWRs and PWRs and requirements for process sampling for BWRs and PWRs.

  8. ANS severe accident program overview & planning document

    SciTech Connect

    Taleyarkhan, R.P.

    1995-09-01

    The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

  9. Severe Accident Test Station Activity Report

    SciTech Connect

    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 Accident 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.

  10. Coupled thermal analysis applied to the study of the rod ejection accident

    SciTech Connect

    Gonnet, M.

    2012-07-01

    An advanced methodology for the assessment of fuel-rod thermal margins under RIA conditions has been developed by AREVA NP SAS. With the emergence of RIA analytical criteria, the study of the Rod Ejection Accident (REA) would normally require the analysis of each fuel rod, slice by slice, over the whole core. Up to now the strategy used to overcome this difficulty has been to perform separate analyses of sampled fuel pins with conservative hypotheses for thermal properties and boundary conditions. In the advanced methodology, the evaluation model for the Rod Ejection Accident (REA) integrates the node average fuel and coolant properties calculation for neutron feedback purpose as well as the peak fuel and coolant time-dependent properties for criteria checking. The calculation grid for peak fuel and coolant properties can be specified from the assembly pitch down to the cell pitch. The comparative analysis of methodologies shows that coupled methodology allows reducing excessive conservatism of the uncoupled approach. (authors)

  11. 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

  12. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    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, 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

  13. Power module assemblies with staggered coolant channels

    DOEpatents

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  14. Transpiration cooling using air as a coolant

    SciTech Connect

    Kikkawa, Shinzo; Senda, Mamoru; Sakagushi, Katsuji; Shibutani, Hideki )

    1993-02-01

    Transpiration cooling is one of the most effective techniques for protecting a surface exposed to a high-temperature gas stream. In the present paper, the transpiration cooling effectiveness was measured under steady state. Air as a coolant was transpired from the surface of a porous plate exposed to hot gas stream, and the transpiration rate was varied in the range of 0.001 [approximately] 0.006. The transpiration cooling effectiveness was evaluated by measuring the temperature of the upper surface of the plate. Also, a theoretical study was performed and it was clarified that the effectiveness increases with increasing transpiration rate and heat-transfer coefficient of the upper surface. Further, the effectiveness was expressed as a function of the blowing parameter only. The agreement between the experimental results and theoretical ones was satisfactory.

  15. Testing of organic acids in engine coolants

    SciTech Connect

    Weir, T.W.

    1999-08-01

    The effectiveness of 30 organic acids as inhibitors in engine coolants is reported. Tests include glassware corrosion of coupled and uncoupled metals. FORD galvanostatic and cyclic polarization electrochemistry for aluminum pitting, and reserve alkalinity (RA) measurements. Details of each test are discussed as well as some general conclusions. For example, benzoic acid inhibits coupled metals well but is ineffective on cast iron when uncoupled. In benzoic acid inhibits coupled metals well but is ineffective on cast iron when uncoupled. In general, the organic acids provide little RA when titrated to a pH of 5.5, titration to a pH of 4.5 can result in precipitation of the acid. Trends with respect to acid chain length are reported also.

  16. 73. View of line of stainless steel coolant storage tanks ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    73. View of line of stainless steel coolant storage tanks for bi-sodium sulfate/water coolant solution at first floor of transmitter building no. 102. - Clear Air Force Station, Ballistic Missile Early Warning System Site II, One mile west of mile marker 293.5 on Parks Highway, 5 miles southwest of Anderson, Anderson, Denali Borough, AK

  17. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Coolant tank tests. 23.1063 Section 23.1063 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Liquid Cooling § 23.1063 Coolant tank tests. Each...

  18. Automatic coolant flow control device for a nuclear reactor assembly

    DOEpatents

    Hutter, E.

    1984-01-27

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  19. Progress in accident analysis of the HYLIFE-II inertial fusion energy power plant design

    SciTech Connect

    Reyes, S; Latkowski, J F; Gomez del Rio, J; Sanz, J

    2000-10-11

    The present work continues our effort to perform an integrated safety analysis for the HYLIFE-II inertial fusion energy (IFE) power plant design. Recently we developed a base case for a severe accident scenario in order to calculate accident doses for HYLIFE-II. It consisted of a total loss of coolant accident (LOCA) in which all the liquid flibe (Li{sub 2}BeF{sub 4}) was lost at the beginning of the accident. Results showed that the off-site dose was below the limit given by the DOE Fusion Safety Standards for public protection in case of accident, and that his dose was dominated by the tritium released during the accident.

  20. 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.

  1. Radiant energy receiver having improved coolant flow control means

    DOEpatents

    Hinterberger, H.

    1980-10-29

    An improved coolant flow control for use in radiant energy receivers of the type having parallel flow paths is disclosed. A coolant performs as a temperature dependent valve means, increasing flow in the warmer flow paths of the receiver, and impeding flow in the cooler paths of the receiver. The coolant has a negative temperature coefficient of viscosity which is high enough such that only an insignificant flow through the receiver is experienced at the minimum operating temperature of the receiver, and such that a maximum flow is experienced at the maximum operating temperature of the receiver. The valving is accomplished by changes in viscosity of the coolant in response to the coolant being heated and cooled. No remotely operated valves, comparators or the like are needed.

  2. World commercial aircraft accidents

    SciTech Connect

    Kimura, C.Y.

    1993-01-01

    This report is a compilation of all accidents world-wide involving aircraft in commercial service which resulted in the loss of the airframe or one or more fatality, or both. This information has been gathered in order to present a complete inventory of commercial aircraft accidents. Events involving military action, sabotage, terrorist bombings, hijackings, suicides, and industrial ground accidents are included within this list. Included are: accidents involving world commercial jet aircraft, world commercial turboprop aircraft, world commercial pistonprop aircraft with four or more engines and world commercial pistonprop aircraft with two or three engines from 1946 to 1992. Each accident is presented with information in the following categories: date of the accident, airline and its flight numbers, type of flight, type of aircraft, aircraft registration number, construction number/manufacturers serial number, aircraft damage, accident flight phase, accident location, number of fatalities, number of occupants, cause, remarks, or description (brief) of the accident, and finally references used. The sixth chapter presents a summary of the world commercial aircraft accidents by major aircraft class (e.g. jet, turboprop, and pistonprop) and by flight phase. The seventh chapter presents several special studies including a list of world commercial aircraft accidents for all aircraft types with 100 or more fatalities in order of decreasing number of fatalities, a list of collision accidents involving commercial aircrafts, and a list of world commercial aircraft accidents for all aircraft types involving military action, sabotage, terrorist bombings, and hijackings.

  3. Test Data for USEPR Severe Accident Code Validation

    SciTech Connect

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  4. Designing an Experimental "Accident"

    ERIC Educational Resources Information Center

    Picker, Lester

    1974-01-01

    Describes an experimental "accident" that resulted in much student learning, seeks help in the identification of nematodes, and suggests biology teachers introduce similar accidents into their teaching to stimulate student interest. (PEB)

  5. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    NASA Technical Reports Server (NTRS)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  6. Severe Accident Test Station Design Document

    SciTech Connect

    Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

    2015-09-01

    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 to 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.

  7. Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Joppen, F.; Van den Berghe, S.

    2009-10-01

    In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% 235U enriched UAl x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% 235U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE. Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900-950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.

  8. Method for removing cesium from a nuclear reactor coolant

    DOEpatents

    Colburn, Richard P.

    1986-01-01

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium

  9. Steam as turbine blade coolant: Experimental data generation

    SciTech Connect

    Wilmsen, B.; Engeda, A.; Lloyd, J.R.

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  10. Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

    SciTech Connect

    Moretti, Fabio; Melideo, Daniele; Terzuoli, Fulvio; D'Auria, Francesco

    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 of 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)

  11. Corrosion problems with aqueous coolants, final report

    SciTech Connect

    Diegle, R B; Beavers, J A; Clifford, J E

    1980-04-11

    The results of a one year program to characterize corrosion of solar collector alloys in aqueous heat-transfer media are summarized. The program involved a literature review and a laboratory investigation of corrosion in uninhibited solutions. It consisted of three separate tasks, as follows: review of the state-of-the-art of solar collector corrosion processes; study of corrosion in multimetallic systems; and determination of interaction between different waters and chemical antifreeze additives. Task 1 involved a comprehensive review of published literature concerning corrosion under solar collector operating conditions. The reivew also incorporated data from related technologies, specifically, from research performed on automotive cooling systems, cooling towers, and heat exchangers. Task 2 consisted of determining the corrosion behavior of candidate alloys of construction for solar collectors in different types of aqueous coolants containing various concentrations of corrosive ionic species. Task 3 involved measuring the degradation rates of glycol-based heat-transfer media, and also evaluating the effects of degradation on the corrosion behavior of metallic collector materials.

  12. Reactor coolant pump monitoring and diagnostic system

    SciTech Connect

    Singer, R.M.; Gross, K.C.; Walsh, M. ); Humenik, K.E. )

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs.

  13. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    SciTech Connect

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  14. Heat transfer processes during intermediate and large break loss-of-coolant accidents (LOCAs)

    SciTech Connect

    Vojtek, I

    1986-09-01

    The general purpose of this project was the investigation of the heat transfer regimes during the high pressure portion of blowdown. The main attention has been focussed on the evaluation of those phenomena which are most important in reactor safety, such as maximum and minimum critical heat flux and forced convection film boiling heat transfer. The experimental results of the 25-rod bundle blowdown heat transfer tests, which were performed at the KWU heat transfer test facility in Karlstein, were used as a database for the verification of different correlations which are used or were developed for the analysis of reactor safety problems. The computer code BRUDI-VA was used for the calculation of local values of important thermohydraulic parameters in the bundle.

  15. INVESTIGATION OF CLEANER TECHNOLOGIES TO MINIMIZE AUTOMOTIVE COOLANT WASTES

    EPA Science Inventory

    The US Environmental Protection Agency in cooperation with the State of New Jersey evaluated chemical filtration and distillation technologies designed to recycle automotive and heavy-duty engine coolants. These evaluations addressed the product quality, waste reduction and econo...

  16. Optimized Coolant-Flow Diverter For Increased Bearing Life

    NASA Technical Reports Server (NTRS)

    Subbaraman, Maria R.; Butner, Myles F.

    1995-01-01

    Coolant-flow diverter for rolling-element bearings in cryogenic turbopump designed to enhance cooling power of flow in contact with bearings and thereby reduce bearing wear. Delivers jets of coolant as close as possible to hot spots at points of contact between balls and race. Also imparts swirl that enhances beneficial pumping effect. Used with success in end ball bearing of high-pressure-oxidizer turbopump.

  17. Study of extended life coolant with suspended carbon nanotubes

    NASA Astrophysics Data System (ADS)

    Overturf, Logan

    Utilizing an experimental facility which was prepared to conduct performance tests on heat exchangers; experiments were completed in an attempt to see verifiable improvements in overall heat transfer coefficient in engine coolant with nanoparticles suspended at different weight percentages. The different fluids tested were: base ELC (Extended Life Coolant), ELC with 0.002 wt% CNT (Carbon Nanotubes), ELC with 0.02 wt% CNT, ELC with 0.02 wt% MWNT's (Multiwalled Nanotubes) and water. The volume percents range from 0.00164 volume% to 0.0164 volume% which seemed quite small, but according to Caterpillar representatives, were the best concentration. These fluids were tested at standard flowrates which this type of heat exchanger would be used in as well as a higher air flowrate and lower coolant flowrates in an attempt to gather more verifiable data. Results were obtained regarding the change in heat transfer ability of engine coolant with suspended nanoparticles. For this system under these specific conditions, there was verifiably no increase in UA as nanoparticles were added to the coolant. The benefits of adding nanoparticles to engine coolant have potential to be great, but the cost of nanoparticles and difficulty keeping them suspended may outweigh any benefits obtainable in this type of set up.

  18. Visualization of Traffic Accidents

    NASA Technical Reports Server (NTRS)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

    Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

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

    SciTech Connect

    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 through 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)

  20. Sealing of a shrouded rotor-stator system with pre-swirl coolant

    NASA Astrophysics Data System (ADS)

    El-Oun, Z. B.; Neller, P. H.; Turner, A. B.

    1987-05-01

    Experimental results for a modeled gas turbine rotor-stator system using both preswirled blade coolant and radially outward flowing disc coolant are presented. Although the preswirled coolant flow is found to have little effect on the pressure distribution below the preswirl nozzles, it is shown that considerable contamination of the preswirled coolant by the frictionally heated disc coolant can occur. A clear pressure inversion effect was found when coolant was provided by the preswirl nozzles alone, while the pressure under the rim seal increased with increasing rotational speed. Blade coolant flow increases the sealing flow requirement, except at the lowest flow rates.

  1. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    DOEpatents

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  2. Laser accidents: Being Prepared

    SciTech Connect

    Barat, K

    2003-01-24

    The goal of the Laser Safety Officer and any laser safety program is to prevent a laser accident from occurring, in particular an injury to a person's eyes. Most laser safety courses talk about laser accidents, causes, and types of injury. The purpose of this presentation is to present a plan for safety offices and users to follow in case of accident or injury from laser radiation.

  3. Reactor coolant seal testing under station blackout conditions

    SciTech Connect

    Marsi, J.A.

    1988-01-01

    Failures of reactor coolant pump (RCP) seals that could result in a significant loss-of-coolant inventory are of current concern to the US Nuclear Regulatory Commission. Particular attention is being focused on seal behavior during station blackout conditions, when failure of on-site emergency diesel generators occurs simultaneously with loss of all off-site alternating current power. Under these conditions, both seal injection flow and component cooling water flow are lost, and the RCP seals are exposed to full reactor coolant temperature. Overheating of elastomeric components and flashing of coolant across the sealing faces can cause unacceptably high leakage rates, with potential catastrophic consequences. A test program has been conducted that subjects full-scale seal cartridges to typical pressurized water reactor (PWR) coolant system steady-state and transient operation conditions including associated dynamic shaft motions. A special test segment was developed to evaluate seal operation under station blackout conditions. The test program successfully mirrored the severity of an actual loss-of-seal cooling water event under station blackout conditions, and the Byron Jackson{reg sign} N-9000 seal cartridge maintained its integrity.

  4. Stagnation region gas film cooling: Effects of dimensionless coolant temperature

    NASA Technical Reports Server (NTRS)

    Bonnice, M. A.; Lecuyer, M. R.

    1983-01-01

    An experimental investigation was conducted to mode the film cooling performance for a turbine vane leading edge using the stagnation region of a cylinder in cross flow. Experiments were conducted with a single row of spanwise angled (25 deg) coolant holes for a range of the coolant blowing ratio and dimensionless coolant temperature with free stream-to-wall temperature ratio approximately 1.7 and Re sub D = 90000. the cylindrical test surface was instrumented with miniature heat flux gages and wall thermocouples to determine the percentage reduction in the Stanton number as a function of the distance downstream from injection (x/d sub 0) and the location between adjacent holes (z/S). Data from local heat flux measurements are presented for injection from a single row located at 5 deg, 22.9 deg, 40.8 deg, from stagnation using a hole spacing ratio of S/d = 5. The film coolant was injected with T sub c T sub w with a dimensionless coolant temperature in the range 1.18 or equal to theta sub c or equal to 1.56. The data for local Stanton Number Reduction (SNR) showed a significant increase in SNR as theta sub c was increased above 1.0.

  5. Nuclear criticality safety assessment of the proposed CFC replacement coolants

    SciTech Connect

    Jordan, W.C.; Dyer, H.R.

    1993-12-01

    The neutron multiplication characteristics of refrigerant-114 (R-114) and proposed replacement coolants perfluorobutane (C{sub 4}F{sub 10}) and cycloperfluorobutane C{sub 4}F{sub 8}) have been compared by evaluating the infinite media multiplication factors of UF{sub 6}/H/coolant systems and by replacement calculations considering a 10-MW freezer/sublimer. The results of these comparisons demonstrate that R-114 is a neutron absorber, due to its chlorine content, and that the alternative fluorocarbon coolants are neutron moderators. Estimates of critical spherical geometries considering mixtures of UF{sub 6}/HF/C{sub 4}F{sub 10} indicate that the flourocarbon-moderated systems are large compared with water-moderated systems. The freezer/sublimer calculations indicate that the alternative coolants are more reactive than R-114, but that the reactivity remains significantly below the condition of water in the tubes, which was a limiting condition. Based on these results, the alternative coolants appear to be acceptable; however, several follow-up tasks have been recommended, and additional evaluation will be required on an individual equipment basis.

  6. Safety analysis results for cryostat ingress accidents in ITER

    SciTech Connect

    Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

    1996-12-31

    Accidents involving the ingress of air or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  7. Radiological Impact Assessment (RIA) following a postulated accident in PHWRS

    SciTech Connect

    Soni, N.; Kansal, M.; Rammohan, H. P.; Malhotra, P. K.

    2012-07-01

    Radiological Impact Assessment (RIA) following postulated accident i.e Loss of Coolant Accident (LOCA) with failed Emergency Core Cooling System (ECCS), performed as part of the reactor safety analysis of a typical 700 MWe Indian Pressurized Heavy Water Reactor(PHWR). The rationale behind the assessment is that the public needs to be protected in the event that the postulated accident results in radionuclide release outside containment. Radionuclides deliver dose to the human body through various pathways namely, plume submersion, exposure due to ground deposition, inhalation and ingestion. The total exposure dose measured in terms of total effective dose equivalent (TEDE) is the sum of doses to a hypothetical adult human at exclusion zone boundary by all the exposure pathways. The analysis provides the important inputs to decide upon the type of emergency counter measures to be adopted during the postulated accident. The importance of the various pathways in terms of contribution to the total effective dose equivalent(TEDE) is also assessed with respect to time of exposure. Inhalation and plume gamma dose are the major contributors towards TEDE during initial period of accident whereas ingestion and ground shine dose start dominating in TEDE in the extended period of exposure. Moreover, TEDE is initially dominated by I-131, Kr-88, Te-132, I-133 and Sr-89, whereas, as time progresses, Xe-133,I-131 and Te-132 become the main contributors. (authors)

  8. Civil aircraft accident investigation.

    PubMed

    Haines, Daniel

    2013-01-01

    This talk reviews some historic aircraft accidents and some more recent. It reflects on the division of accident causes, considering mechanical failures and aircrew failures, and on aircrew training. Investigation results may lead to improved aircraft design, and to appropriate crew training. PMID:24057309

  9. Anatomy of an Accident.

    ERIC Educational Resources Information Center

    Mobley, Michael

    1984-01-01

    The findings of industrial safety engineers in the areas of accident causation and prevention are wholly applicable to adventure programs. Adventure education instructors can use safety engineering concepts to assess the risk in a particular activity, understand factors that cause accidents, and intervene to minimize injuries and damages if…

  10. Farm accidents in children.

    PubMed Central

    Cameron, D.; Bishop, C.; Sibert, J. R.

    1992-01-01

    OBJECTIVE--To examine the problem of accidental injury to children on farms. DESIGN--Prospective county based study of children presenting to accident and emergency departments over 12 months with injuries sustained in a farm setting and nationwide review of fatal childhood farm accidents over the four years April 1986 to March 1990. SETTING--Accident and emergency departments in Aberystwyth, Carmarthen, Haverfordwest, and Llanelli and fatal accidents in England, Scotland, and Wales notified to the Health and Safety Executive register. SUBJECTS--Children aged under 16. MAIN OUTCOME MEASURE--Death or injury after farm related accidents. RESULTS--65 accidents were recorded, including 18 fractures. Nine accidents necessitated admission to hospital for a mean of two (range one to four) days. 13 incidents were related to tractors and other machinery; 24 were due to falls. None of these incidents were reported under the statutory notification scheme. 33 deaths were notified, eight related to tractors and allied machinery and 10 related to falling objects. CONCLUSIONS--Although safety is improving, the farm remains a dangerous environment for children. Enforcement of existing safety legislation with significant penalties and targeting of safety education will help reduce accident rates further. PMID:1638192

  11. Method for removing cesium from a nuclear reactor coolant

    DOEpatents

    Colburn, R.P.

    1983-08-10

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium inventory thereof further into the carbon matrix while simultaneously redispersing a portion into the regeneration system for absorption at a reduced temperature by the secondary trap.

  12. Actively controlling coolant-cooled cold plate configuration

    SciTech Connect

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  13. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    SciTech Connect

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  14. The nature and behavior of particulates in PWR primary coolant

    SciTech Connect

    Bridle, D.A.; Butter, K.R.; Cake, P.; Comley, G.C.W.; Mitchell, C.R. )

    1989-12-01

    A study of particle size distributions, nature and behavior of insoluble species carried by PWR coolants has been carried out over a four year period in Belgian reactors. Comparative data was obtained by the use of improved sampling systems designed to overcome the inadequacies of standard facilities. Coolant data is presented from commissioning and early operation of new plant to that in established PWR circuits. Results arising from reactors transients are also included, which refer to shutdown and start-up phases, power changes and scram situations. The information obtained includes chemical and radiochemical characteristics of particulates and their contribution to total activity carried by reactor coolant. The implications for plant operations are discussed. 16 refs., 55 figs., 24 tabs.

  15. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    NASA Astrophysics Data System (ADS)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  16. Code System to Calculate Reactor Coolant System Leak Rate.

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report ofmore » the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.« less

  17. Development of Figure of Merits (FOMs) for Intermediate Coolant Characterization and Selection

    SciTech Connect

    Eung Soo Kim; Piyush Sabharwall; Nolan Anderson

    2011-06-01

    This paper focuses on characterization of several coolant performances in the IHTL. There are lots of choices available for the IHTL coolants; gases, liquid metals, molten salts, and etc. Traditionally, the selection of coolants is highly dependent on engineer's experience and decisions. In this decision, the following parameters are generally considered: melting point, vapor pressure, density, thermal conductivity, heat capacity, viscosity, and coolant chemistry. The followings are general thermal-hydraulic requirements for the coolant in the IHTL: (1) High heat transfer performance - The IHTL coolant should exhibit high heat transfer performance to achieve high efficiency and economics; (2) Low pumping power - The IHTL coolant requires low pumping power to improve economics through less stringent pump requirements; (3) Low amount of coolant volume - The IHTL coolant requires less coolant volume for better economics; (4) Low amount of structural materials - The IHTL coolant requires less structural material volume for better economics; (5) Low heat loss - The IHTL requires less heat loss for high efficiency; and (6) Low temperature drop - The IHTL should allow less temperature drop for high efficiency. Typically, heat transfer coolants are selected based on various fluid properties such as melting point, vapor pressure, density, thermal conductivity, heat capacity, viscosity, and coolant chemistry. However, the selection process & results are highly dependent on the engineer's personal experience and skills. In the coolant selection, if a certain coolant shows superior properties with respect to the others, the decision will be very straightforward. However, generally, each coolant material exhibits good characteristics for some properties but poor for the others. Therefore, it will be very useful to have some figures of merits (FOMs), which can represent and quantify various coolant thermal performances in the system of interest. The study summarized in this

  18. Progress in the studies of passive heat removal in the next European torus under accident conditions

    SciTech Connect

    Soria, A. ); Renda, V.; Papa, L. . Joint Research Centre); Fenoglio, F. )

    1989-12-01

    Within the framework of safety analysis for the next European torus, a decay heat hazards assessment is under way in Ispra. Undercooling accidents (loss-of-coolant and loss-of-flow accidents (LOCAs and LOFAs)) due to pump failure have been investigated assuming an automatic plasma shutdown in both cases. The passive heat removal mechanisms considered include radiation between components and residual cooling by the thermosyphon effect in the main cooling circuits. Conservative thermohydraulic calculations have been made to determine coolant velocity and temperature transients to avoid water boiling int he circuits. Results show that during a LOFA, water boiling can be avoided provided that the water inertia is large enough, and material melting temperatures are not reached during a LOCA.

  19. Persistence of airline accidents.

    PubMed

    Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

    2010-10-01

    This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation.

  20. Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

    SciTech Connect

    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 ceramic 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, Fe

  1. MELCOR code analysis of a severe accident LOCA at Peach Bottom Plant

    SciTech Connect

    Carbajo, J.J. )

    1993-01-01

    A design-basis loss-of-coolant accident (LOCA) concurrent with complete loss of the emergency core cooling systems (ECCSs) has been analyzed for the Peach Bottom atomic station unit 2 using the MELCOR code, version 1.8.1. The purpose of this analysis is to calculate best-estimate times for the important events of this accident sequence and best-estimate source terms. Calculated pressures and temperatures at the beginning of the transient have been compared to results from the Peach Bottom final safety analysis report (FSAR). MELCOR-calculated source terms have been compared to source terms reported in the NUREG-1465 draft.

  2. Modular high-temperature gas-cooled reactor core heatup accident simulations

    SciTech Connect

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs.

  3. Estimate of radionuclide release characteristics into containment under severe accident conditions. Final report

    SciTech Connect

    Nourbakhsh, H.P.

    1993-11-01

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented.

  4. 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.

  5. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Coolant tank tests. 23.1063 Section 23.1063 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Liquid Cooling §...

  6. Coolant Characteristics and Control in Direct Chill Casting

    SciTech Connect

    2001-10-01

    This project focuses on understanding the fundamentals of coolant behavior and developing strategies to control the cooling rate of DC casting of aluminum ingots. Project partners will conduct a fundamental study to identify various parameters affecting critical heat flux and boiling transition and evaluate the effects of various additives (impurity particulates, sodium and calcium salts, carbonates, bicarbonates, surfactants, etc.).

  7. Integral coolant channels supply made by melt-out method

    NASA Technical Reports Server (NTRS)

    Escher, W. J. D.

    1964-01-01

    Melt-out method of constructing strong, pressure-tight fluid coolant channels for chambers is accomplished by cementing pins to the surface and by depositing a melt-out material on the surface followed by two layers of epoxy-resin impregnated glass fibers. The structure is heated to melt out the low-melting alloy.

  8. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    EPA Science Inventory

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  9. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    EPA Science Inventory

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  10. PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR TO EMBEDMENT IN CONCRETE. HIGHER PIPE IS INLET; THE OTHER, THE OUTLET LOOP. INLET PIPE WILL CONNECT TO TOP SECTION OF REACTOR VESSEL. INL NEGATIVE NO. 1287. Unknown Photographer, 1/18/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  11. 37. Upper level, chromate tanks (formerly provided coolant to missile ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    37. Upper level, chromate tanks (formerly provided coolant to missile guidance section, retractor cables for lock pin in front of ladder at left - Ellsworth Air Force Base, Delta Flight, Launch Facility, On County Road T512, south of Exit 116 off I-90, Interior, Jackson County, SD

  12. Fuels, Lubricants, and Coolants. FOS: Fundamentals of Service.

    ERIC Educational Resources Information Center

    John Deere Co., Moline, IL.

    This manual on fuels, lubricants, and coolants is one of a series of power mechanics tests and visual aids on automotive and off-the-road agricultural and construction equipment. Materials present basic information with illustrations for use by vocational students and teachers as well as shop servicemen and laymen. Focusing on fuels, the first of…

  13. Corrosion of structural materials by lead-based reactor coolants.

    SciTech Connect

    Abraham, D. P.; Leibowitz, L.; Maroni, V. A.; McDeavitt, S. M.; Raraz, A. G.

    2000-11-16

    Advanced nuclear reactor design has, in recent years, focused increasingly on the use of heavy-liquid-metal coolants, such as lead and lead-bismuth eutectic. Similarly, programs on accelerator-based transmutation systems have also considered the use of such coolants. Russian experience with heavy-metal coolants for nuclear reactors has lent credence to the validity of this approach. Of significant concern is the compatibility of structural materials with these coolants. We have used a thermal convection-based test method to allow exposure of candidate materials to molten lead and lead-bismuth flowing under a temperature gradient. The gradient was deemed essential in evaluating the behavior of the test materials in that should preferential dissolution of components of the test material occur we would expect dissolution in the hotter regions and deposition in the colder regions, thus promoting material transport. Results from the interactions of a Si-rich mild steel alloy, AISI S5, and a ferritic-martensitic stainless steel, HT-9, with the molten lead-bismuth are presented.

  14. Accident resistant transport container

    DOEpatents

    Anderson, J.A.; Cole, K.K.

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  15. Accident resistant transport container

    DOEpatents

    Andersen, John A.; Cole, James K.

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  16. Safety Is No Accident.

    ERIC Educational Resources Information Center

    Christiansen, Monty L.

    1985-01-01

    Liability suits involving accidents in park and recreation areas are expensive and intangible costs are incalculable. Risk management practices related to park planning, personnel, and administrative practices are discussed. (MT)

  17. Experimental interaction of magma and “dirty” coolants

    NASA Astrophysics Data System (ADS)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with <~10% bentonite, heat transfer is by convection, and the melt is efficiently fragmented into blocky particles through multiple thermal granulation events which produce associated acoustic signals. For all coolants with >~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate

  18. Accident management information needs

    SciTech Connect

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. )

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs.

  19. Mechanism of Improvement Effect of Ultrasonically Activated Coolant on Finished Surface Roughness in Cylindrical Grinding

    NASA Astrophysics Data System (ADS)

    Sakamoto, Haruhisa; Kajiwara, Kunio; Shimizu, Shinji; Ohmori, Shigetoshi

    The use of an ultrasonically activated coolant improves the roughness of a ground surface compared with that of an ordinary coolant. By evaluating the change in working surface conditions, it has been clarified that the ultrasonically activated coolant suppresses loading and wheel wear, and maintains a good the working surface condition. The flushing effect of the ultrasonically activated coolant, promoted by giant vibration acceleration, prevents the deposition and welding of chips on the working surface. Since the density of kinetic energy in the coolant stream increases, the activation helps the coolant to overcome the airflow due to wheel rotation and to reach the grinding point efficiently. This promotes the cooling and lubricating effects of coolant, and then protects the cutting edges from crushing, falling off and dulling. On the basis of the results, it can be said that the effects of coolant activation suppress the generation of scratches, and consequently improve the finished surface roughness.

  20. Modular Porous Plate Sublimator /MPPS/ requires only water supply for coolant

    NASA Technical Reports Server (NTRS)

    Rathbun, R. J.

    1966-01-01

    Modular porous plate sublimators, provided for each location where heat must be dissipated, conserve the battery power of a space vehicle by eliminating the coolant pump. The sublimator requires only a water supply for coolant.

  1. Directly connected heat exchanger tube section and coolant-cooled structure

    SciTech Connect

    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.

  2. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  3. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R. ); Choi, C.J. )

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

  4. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R.; Choi, C.J.

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d`Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d`Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

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

    SciTech Connect

    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 failure 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.

  6. Copper-triazole interaction and coolant inhibitor depletion

    SciTech Connect

    Bartley, L.S.; Fritz, P.O.; Pellet, R.J.; Taylor, S.A.; Van de Ven, P.

    1999-08-01

    To a large extent, the depletion of tolyltriazole (TTZ) observed in several field tests may be attributed to the formation of a protective copper-triazole layer. Laboratory aging studies, shown to correlate with field experience, reveal that copper-TTZ layer formation depletes coolant TTZ levels in a fashion analogous to changes observed in the field. XPS and TPD-MS characterization of the complex formed indicates a strong chemical bond between copper and the adsorbed TTZ which can be desorbed thermally only at elevated temperatures. Electrochemical polarization experiments indicate that the layer provides good copper protection even when TTZ is absent from the coolant phase. Examination of copper cooling system components obtained after extensive field use reveals the presence of a similar protective layer.

  7. Health physics aspects of processing EBR-I coolant

    SciTech Connect

    Burke, L.L.; Thalgott, J.O.; Poston, J.W. Jr.

    1998-12-31

    The sodium-potassium reactor coolant removed from the Experimental Breeder Reactor Number One after a partial reactor core meltdown had been stored at the Idaho National Engineering and Environmental Laboratory for 40 years. The State of Idaho considered this waste the most hazardous waste stored in the state and required its processing. The reactor coolant was processed in three phases. The first phase converted the alkali metal into a liquid sodium-potassium hydroxide. The second phase converted this caustic to a liquid sodium-potassium carbonate. The third phase solidified the sodium-potassium carbonate into a form acceptable for land disposal. Health physics aspects and dose received during each phase of the processing are discussed.

  8. Hybrid method for numerical modelling of LWR coolant chemistry

    NASA Astrophysics Data System (ADS)

    Swiatla-Wojcik, Dorota

    2016-10-01

    A comprehensive approach is proposed to model radiation chemistry of the cooling water under exposure to neutron and gamma radiation at 300 °C. It covers diffusion-kinetic processes in radiation tracks and secondary reactions in the bulk coolant. Steady-state concentrations of the radiolytic products have been assessed based on the simulated time dependent concentration profiles. The principal reactions contributing to the formation of H2, O2 and H2O2 were indicated. Simulation was carried out depending on the amount of extra hydrogen dissolved in the coolant to reduce concentration of corrosive agents. High sensitivity to the rate of reaction H+H2O=OH+H2 is shown and discussed.

  9. Injuries are not accidents

    PubMed Central

    Gutiérrez, María Isabel

    2014-01-01

    Injuries are the result of an acute exposure to exhort of energy or a consequence of a deficiency in a vital element that exceeds physiological thresholds resulting threatens life. They are classified as intentional or unintentional. Injuries are considered a global health issue because they cause more than 5 million deaths per year worldwide and they are an important contributor to the burden of disease, especially affecting people of low socioeconomic status in low- and middle-income countries. A common misconception exists where injuries are thought to be the same as accidents; however, accidents are largely used as chance events, without taken in consideration that all these are preventable. This review discusses injuries and accidents in the context of road traffic and emphasizes injuries as preventable events. An understanding of the essence of injuries enables the standardization of terminology in public use and facilitates the development of a culture of prevention among all of us. PMID:25386040

  10. Accident prevention manual

    SciTech Connect

    Not Available

    1998-05-01

    Among the many common needs and goals are the safety and well-being of families, ourselves, fellow employees, and the continuing success of this organization. To these ends--minimizing human suffering and economic waste--the Bonneville Power Administration (BPA) Accident Prevention Program and this Accident Prevention Manual (APM) are dedicated. The BPA Accident Prevention Program is revised as necessary to ensure compliance with relevant Federal safety and health standards. The mandatory rules herein express minimum requirements for dealing with the principal hazards inherent in daily work activities. These and other written requirements, which neither can nor should provide complete coverage of all work situations, must be continually reinforced through the sound and mature safety judgments of all workers on each assigned task. In the event of conflicting judgments, the more conservative interpretation shall prevail pending review and resolution by management.

  11. 92. View of transmitter building no. 102 first floor coolant ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    92. View of transmitter building no. 102 first floor coolant process water tanks (sodium bisulfate solution), stainless steel, for electronic systems cooling in transmitter and MIP rooms. RCA Services Company 29 September, 1960, official photograph BMEWS Project by unknown photograph, Photographic Services, Riverton, NJ, BMEWS, clear as negative no. A-1226 - Clear Air Force Station, Ballistic Missile Early Warning System Site II, One mile west of mile marker 293.5 on Parks Highway, 5 miles southwest of Anderson, Anderson, Denali Borough, AK

  12. Expert system for online surveillance of nuclear reactor coolant pumps

    DOEpatents

    Gross, Kenny C.; Singer, Ralph M.; Humenik, Keith E.

    1993-01-01

    An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  13. Evaluation of organic moderator/coolants for fusion breeder blankets

    SciTech Connect

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process.

  14. The Fukushima radiation accident: consequences for radiation accident medical management.

    PubMed

    Meineke, Viktor; Dörr, Harald

    2012-08-01

    The March 2011 radiation accident in Fukushima, Japan, is a textbook example of a radiation accident of global significance. In view of the global dimensions of the accident, it is important to consider the lessons learned. In this context, emphasis must be placed on consequences for planning appropriate medical management for radiation accidents including, for example, estimates of necessary human and material resources. The specific characteristics of the radiation accident in Fukushima are thematically divided into five groups: the exceptional environmental influences on the Fukushima radiation accident, particular circumstances of the accident, differences in risk perception, changed psychosocial factors in the age of the Internet and globalization, and the ignorance of the effects of ionizing radiation both among the general public and health care professionals. Conclusions like the need for reviewing international communication, interfacing, and interface definitions will be drawn from the Fukushima radiation accident. PMID:22951483

  15. The Fukushima radiation accident: consequences for radiation accident medical management.

    PubMed

    Meineke, Viktor; Dörr, Harald

    2012-08-01

    The March 2011 radiation accident in Fukushima, Japan, is a textbook example of a radiation accident of global significance. In view of the global dimensions of the accident, it is important to consider the lessons learned. In this context, emphasis must be placed on consequences for planning appropriate medical management for radiation accidents including, for example, estimates of necessary human and material resources. The specific characteristics of the radiation accident in Fukushima are thematically divided into five groups: the exceptional environmental influences on the Fukushima radiation accident, particular circumstances of the accident, differences in risk perception, changed psychosocial factors in the age of the Internet and globalization, and the ignorance of the effects of ionizing radiation both among the general public and health care professionals. Conclusions like the need for reviewing international communication, interfacing, and interface definitions will be drawn from the Fukushima radiation accident.

  16. Acceptance criteria for reactor coolant pumps and valves

    SciTech Connect

    Gupta, N.K.; Miller, R.F.; Sindelar, R.L.

    1993-05-01

    Each of the six primary coolant loop systems of the Savannah River Site (SRS) production reactors contains one reactor coolant pump, one PUMP suction side motor operated valve, and other smaller valves. The pumps me double suction, double volute, and radially split type pumps. The valves are different size shutoff and control valves rated from ANSI B16.5 construction class 150 to class 300. The reactor coolant system components, also known as the process water system (PWS), are classified as nuclear Safety Class I components. These components were constructed in the 1950`s in accordance with the then prevailing industry practices. No uniform construction codes were used for design and analysis of these components. However, no pressure boundary failures or bolting failures have ever been recorded throughout their operating history. Over the years, the in-service inspection (ISI) was limited to visual inspection of the pressure boundaries, and surface and volumetric examination of the pressure retaining bolts. Efforts are now underway to implement ISI requirements similar to the ASME Section XI requirements for pumps and valves. This report discusses the new ISI requirements which also call for volumetric examination of the pump casing and valve body welds.

  17. Coolant Design System for Liquid Propellant Aerospike Engines

    NASA Astrophysics Data System (ADS)

    McConnell, Miranda; Branam, Richard

    2015-11-01

    Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

  18. Crack stability analysis of low alloy steel primary coolant pipe

    SciTech Connect

    Tanaka, T.; Kameyama, M.; Urabe, Y.

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

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

    SciTech Connect

    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 by 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.

  20. Occupational accidents aboard merchant ships

    PubMed Central

    Hansen, H; Nielsen, D; Frydenberg, M

    2002-01-01

    Objectives: To investigate the frequency, circumstances, and causes of occupational accidents aboard merchant ships in international trade, and to identify risk factors for the occurrence of occupational accidents as well as dangerous working situations where possible preventive measures may be initiated. Methods: The study is a historical follow up on occupational accidents among crew aboard Danish merchant ships in the period 1993–7. Data were extracted from the Danish Maritime Authority and insurance data. Exact data on time at risk were available. Results: A total of 1993 accidents were identified during a total of 31 140 years at sea. Among these, 209 accidents resulted in permanent disability of 5% or more, and 27 were fatal. The mean risk of having an occupational accident was 6.4/100 years at sea and the risk of an accident causing a permanent disability of 5% or more was 0.67/100 years aboard. Relative risks for notified accidents and accidents causing permanent disability of 5% or more were calculated in a multivariate analysis including ship type, occupation, age, time on board, change of ship since last employment period, and nationality. Foreigners had a considerably lower recorded rate of accidents than Danish citizens. Age was a major risk factor for accidents causing permanent disability. Change of ship and the first period aboard a particular ship were identified as risk factors. Walking from one place to another aboard the ship caused serious accidents. The most serious accidents happened on deck. Conclusions: It was possible to clearly identify work situations and specific risk factors for accidents aboard merchant ships. Most accidents happened while performing daily routine duties. Preventive measures should focus on workplace instructions for all important functions aboard and also on the prevention of accidents caused by walking around aboard the ship. PMID:11850550

  1. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect

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

    2012-03-18

    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 accident 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.

  2. Iodine chemical forms in LWR severe accidents

    SciTech Connect

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1991-01-01

    Calculated data from seven severe accident sequences in light water reactor plants were used to assess the chemical forms of iodine in containment. In most of the calculations for the seven sequences, iodine entering containment from the reactor coolant system was almost entirely in the form of CsI with very small contributions of I or HI. The largest fraction of iodine in forms other than CsI was a total of 3.2% as I plus HI. Within the containment, the CsI will deposit onto walls and other surfaces, as well as in water pools, largely in the form of iodide (I{sup {minus}}). The radiation-induced conversion of I{sup {minus}} in water pools into I{sub 2} is strongly dependent on pH. In systems where the pH was controlled above 7, little additional elemental iodine would be produced in the containment atmosphere. When the pH falls below 7, it may be assumed that it is not being controlled and large fractions of iodine as I{sub 2} within the containment atmosphere may be produced. 17 refs., 5 tabs.

  3. Iodine chemical forms in LWR severe accidents

    SciTech Connect

    Beahm, E.C.; Weber, C.F.; Kress, T.S.; Parker, G.W.

    1991-01-01

    Calculated data from seven severe accident sequences in light-water reactor plants were used to assess the chemical forms of iodine in containment. In most of the calculations for the seven sequences, iodine entering containment from the reactor coolant system was almost entirely in the form of CsI with very small contributions of I or HI. The largest fraction of iodine in forms other than CsI was a total of 3.2% as I plus HI. Within the containment, the CsI will deposit onto walls and other surfaces, as well as in water pools, largely in the form of iodide (I{sup {minus}}). The radiation induced conversion of I{sup {minus}} in water pools into I{sub 2} is strongly dependent on pH. In systems where the pH was controlled above 7, little additional elemental iodine would be produced in the containment atmosphere. When the pH falls below 7, it may be assumed that it is not being controlled, and large fractions of iodine as I{sub 2} within the containment atmosphere may be produced. 16 refs.

  4. Effect of Coolant Temperature and Mass Flow on Film Cooling of Turbine Blades

    NASA Technical Reports Server (NTRS)

    Garg, Vijay K.; Gaugler, Raymond E.

    1997-01-01

    A three-dimensional Navier Stokes code has been used to study the effect of coolant temperature, and coolant to mainstream mass flow ratio on the adiabatic effectiveness of a film-cooled turbine blade. The blade chosen is the VKI rotor with six rows of cooling holes including three rows on the shower head. The mainstream is akin to that under real engine conditions with stagnation temperature = 1900 K and stagnation pressure = 3 MPa. Generally, the adiabatic effectiveness is lower for a higher coolant temperature due to nonlinear effects via the compressibility of air. However, over the suction side of shower-head holes, the effectiveness is higher for a higher coolant temperature than that for a lower coolant temperature when the coolant to mainstream mass flow ratio is 5% or more. For a fixed coolant temperature, the effectiveness passes through a minima on the suction side of shower-head holes as the coolant to mainstream mass flow, ratio increases, while on the pressure side of shower-head holes, the effectiveness decreases with increase in coolant mass flow due to coolant jet lift-off. In all cases, the adiabatic effectiveness is highly three-dimensional.

  5. Heat transfer characteristics for some coolant additives used for water cooled engines

    SciTech Connect

    Abou-Ziyan, H.Z.; Helali, A.H.B.

    1996-12-31

    Engine coolants contain certain additives to prevent engine overheating or coolant freezing in cold environments. Coolants, also, contain corrosion and rust inhibitors, among other additives. As most engines are using engine cooling solutions, it is of interest to evaluate the effect of engine coolants on the boiling heat transfer coefficient. This has its direct impact on radiator size and environment. This paper describes the apparatus and the measurement techniques. Also, it presents the obtained boiling heat transfer results at different parameters. Three types of engine coolants and their mixtures in distilled water are evaluated, under sub-cooled and saturated boiling conditions. A profound effect of the presence of additives in the coolant, on heat transfer, was clear since changes of heat transfer for different coolants were likely to occur. The results showed that up to 180% improvement of boiling heat transfer coefficient is experienced with some types of coolants. However, at certain concentrations other coolants provide deterioration or not enhancement in the boiling heat transfer characteristics. This investigation proved that there are limitations, which are to be taken into consideration, for the composition of engine coolants in different environments. In warm climates, ethylene glycol should be kept at the minimum concentration required for dissolving other components, whereas borax is beneficial to the enhancement of the heat transfer characteristics.

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

    SciTech Connect

    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 heat 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.

  7. Radionuclide mass balance for the TMI-2 accident: data through 1979 and preliminary assessment of uncertainties

    SciTech Connect

    Davis, R J; Tonkay, D W; Vissing, E A; Nguyen, T D; Shawn, L W; Goldman, M I

    1984-11-01

    A systematic data base of available information needed to calculate mass balances of key radionuclides arising from the Three-Mile Island Unit 2 (TMI-2) accident as a function of time has been assembled. The sample and analysis data represent the likely major sinks except for the solids remaining in the primary system. Surfaces in the primary system are represented only by preliminary data pertaining to cesium deposition on plenum surfaces. TMI-2 component description data are included for the reactor coolant, makeup and purification, and liquid waste systems and the reactor building. The chronology of liquid transfers through the end of 1979 is included. A mass transfer model has been developed. It is concluded that tritium and cesium released into the reactor coolant traveled with the reactor coolant without losses to other phases during transit and storage. The data also suggest that tritium and cesium were not leached from primary solids and surfaces after the accident, although strontium has gradually leached from the primary solids and surfaces over a long period. Much of the iodine transferred to the reactor building sump/basement is suspected of having transferred to surfaces or solids from the sump/basement water and was therefore not found in basement water samples.

  8. Applying STAMP in Accident Analysis

    NASA Technical Reports Server (NTRS)

    Leveson, Nancy; Daouk, Mirna; Dulac, Nicolas; Marais, Karen

    2003-01-01

    Accident models play a critical role in accident investigation and analysis. Most traditional models are based on an underlying chain of events. These models, however, have serious limitations when used for complex, socio-technical systems. Previously, Leveson proposed a new accident model (STAMP) based on system theory. In STAMP, the basic concept is not an event but a constraint. This paper shows how STAMP can be applied to accident analysis using three different views or models of the accident process and proposes a notation for describing this process.

  9. AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

    SciTech Connect

    Knight, T.D.

    1996-07-01

    The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA.

  10. European Pressurized water Reactor (EPR) SAR ATWS Accident Analyses by using 3D Code Internal Coupling Method

    SciTech Connect

    Gagner, Renata; Lafitte, Helene; Dormeau, Pascal; Stoudt, Roger H.

    2004-07-01

    Anticipated Transients Without Scram (ATWS) accident analyses make part of the Safety Analysis Report of the European Pressurized water Reactor (EPR), covering Risk Reduction Category A (Core Melt Prevention) events. This paper deals with three of the most penalizing RRC-A sequences of ATWS caused by mechanical blockage of the control/shutdown rods, regarding their consequences on the Reactor Coolant System (RCS) and core integrity. A new 3D code internal coupling calculation method has been introduced. (authors)

  11. Tractor accidents in Swedish traffic.

    PubMed

    Pinzke, Stefan; Nilsson, Kerstin; Lundqvist, Peter

    2012-01-01

    The objective of this study is to reach a better understanding of accidents on Swedish roads involving tractors and to suggest ways of preventing them. In an earlier study we analyzed police-reported fatal accidents and accidents that led to physical injuries from 1992 to 2005. During each year of this period, tractors were involved in 128 traffic accidents on average, an average of 7 people were killed, 44 sustained serious injuries, and 143 sustained slight injuries. The number of fatalities in these tractor accidents was about 1.3% of all deaths in traffic accidents in Sweden. Cars were most often involved in the tractor accidents (58%) and 15% were single vehicle accidents. The mean age of the tractor driver involved was 39.8 years and young drivers (15-24 years) were overrepresented (30%). We are now increasing the data collected with the years 2006-2010 in order to study the changes in the number of accidents. Special attention will be given to the younger drivers and to single vehicle accidents. Based on the results we aim to develop suggestions for reducing road accidents, e.g. including measures for making farm vehicles more visible and improvement of the training provided at driving schools. PMID:22317543

  12. System Study: High-Pressure Coolant Injection 1998-2014

    SciTech Connect

    Schroeder, John Alton

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. 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 or decreasing trends were identified in the HPCI results.

  13. Vegetable oils: liquid coolants for solar heating and cooling applications

    SciTech Connect

    Ingley, H A

    1980-02-01

    It has been proposed that vegetable oils, renewable byproducts of agriculture processes, be investigated for possible use as liquid coolants. The major thrust of the project was to investigate several thermophysical properties of the four vegetable oils selected. Vapor pressures, specific heat, viscosity, density, and thermal conductivity were determined over a range of temperatures for corn, soybean, peanut, and cottonseed oil. ASTM standard methods were used for these determinations. In addition, chemical analyses were performed on samples of each oil. The samples were collected before and after each experiment so that any changes in composition could be noted. The tests included iodine number, fatty acid, and moisture content determination. (MHR)

  14. System Study: High-Pressure Coolant Injection 1998–2013

    SciTech Connect

    Schroeder, John Alton

    2015-01-31

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 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 or decreasing trends were identified in the HPCI results.

  15. System Study: High-Pressure Coolant Injection 1998-2012

    SciTech Connect

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). 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 or decreasing trends were identified in the HPCI results.

  16. Cryogenic-coolant He4-superconductor dynamic and static interactions

    NASA Technical Reports Server (NTRS)

    Caspi, S.; Chuang, C.; Kim, Y. I.; Allen, R. J.; Frederking, T. H. E.

    1980-01-01

    A composite superconducting material (NbTi-Cu) was evaluated with emphasis on post quench solid cooling interaction regimes. The quasi-steady runs confirm the existence of a thermodynamic limiting thickness for insulating coatings. Two distinctly different post quench regimes of coated composites are shown to relate to the limiting thickness. Only one regime,, from quench onset to the peak value, revealed favorable coolant states, in particular in He2. Transient recovery shows favorable recovery times from this post quench regime (not drastically different from bare conductors) for both single coated specimens and a coated conductor bundle.

  17. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor (U)

    SciTech Connect

    O'Kula, K.R.; Olson, R.L.; Hamby, D.M. )

    1992-03-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/PBq(2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This paper suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium releases to the atmosphere than from comparable tritium source terms to water pathways.

  18. Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

    SciTech Connect

    Harrington, R.M.; Ball, S.J.

    1985-01-01

    One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600/sup 0/C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth.

  19. [Skateboard and rollerskate accidents].

    PubMed

    Lohmann, M; Petersen, A O; Pedersen, O D

    1990-05-28

    The increasing popularity of skateboards and rollerskates has resulted in an increased number of contacts with the casualty department in Denmark after accidents. As part of the Danish share in the EHLASS project (European Home and Leisure Surveillance System), 120,000 consecutive contacts with the casualty departments were reviewed. Out of these 516 were due to accidents with skateboards and rollerskates (181/335). A total of 194 of these injuries (38%) were fractures and 80% of these were in the upper limbs. Twenty fractures required reposition under general anaesthesia and two required osteosynthesis. Nine patients were admitted for observation for concussion. One patient had sustained rupture of the spleen and splenectomy was necessary. A total of 44 patients were admitted. None of the 516 patients had employed protective equipment on the injured region. Considerable reduction in the number of injuries could probably be produced by employment of suitable protective equipment.

  20. [Drowning accidents in childhood].

    PubMed

    Krandick, G; Mantel, K

    1990-09-30

    This is a report on five boys aged between 1 and 5 years who, after prolonged submersion in cold water, were treated at our department. On being taken out of the water, all the patients were clinically dead. After 1- to 3-hour successful cardiopulmonary resuscitation, with a rectal temperature of about 27 degrees C, they were rewarmed at a rate of 1 degree/hour. Two patients died within a few hours after the accident. One patient survived with an apallic syndrome, 2 children survived with no sequelae. In the event of a water-related accident associated with hypothermia, we consider suitable resuscitation to have preference over rewarming measures. The most important treatment guidelines and prognostic factors are discussed.

  1. Farm accidents in children.

    PubMed

    Cogbill, T H; Busch, H M; Stiers, G R

    1985-10-01

    During a 6 1/2 year period, 105 children were admitted to the hospital as the result of trauma that occurred on farms. The mechanism of injury was animal related in 42 (40%), tractor or wagon accident in 28 (26%), farm machinery in 21 (20%), fall from farm building in six (6%), and miscellaneous in eight (8%). Injury Severity Score was calculated for each patient. An Injury Severity Score of greater than or equal to 25 was determined for 11 children (11%). Life-threatening injuries, therefore, are frequently the result of childhood activities that take place in agricultural environments. The most common injuries were orthopedic, neurologic, thoracoabdominal, and maxillofacial. There was one death in the series, and only one survivor sustained major long-term disability. Such injuries are managed with optimal outcome in a regional trauma center. Educational programs with an emphasis on prevention and safety measures may reduce the incidence of farm accidents. PMID:4047799

  2. Load following capability of CANDLE reactor by adjusting coolant operation condition

    SciTech Connect

    Sekimoto, Hiroshi; Nakayama, Sinsuke

    2012-06-06

    The load following capability of CANDLE reactor is investigated in the condition that the control rods are unavailable. Both sodium cooled metallic fuel fast reactor (SFR) and {sup 208}Pb cooled metallic fuel fast reactor (LFR) are investigated for their performance in power rate changing by changing its coolant operation condition; either coolant flow rate or coolant inlet temperature. The change by coolant flow rate is difficult especially for SFR because the maximum temperature criteria on cladding material may be violated. The power rate can be changed for its full range easily by changing the coolant temperature at the core inlet. LFR can reduce the same amount of power rate by smaller change of temperature than SFR. However, the coolant output temperature is generally decreased for this method and the thermal efficiency becomes worse.

  3. Load following capability of CANDLE reactor by adjusting coolant operation condition

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi; Nakayama, Sinsuke

    2012-06-01

    The load following capability of CANDLE reactor is investigated in the condition that the control rods are unavailable. Both sodium cooled metallic fuel fast reactor (SFR) and 208Pb cooled metallic fuel fast reactor (LFR) are investigated for their performance in power rate changing by changing its coolant operation condition; either coolant flow rate or coolant inlet temperature. The change by coolant flow rate is difficult especially for SFR because the maximum temperature criteria on cladding material may be violated. The power rate can be changed for its full range easily by changing the coolant temperature at the core inlet. LFR can reduce the same amount of power rate by smaller change of temperature than SFR. However, the coolant output temperature is generally decreased for this method and the thermal efficiency becomes worse.

  4. Aerodynamic effect of coolant ejection in the rear part of transonic rotor blades

    NASA Astrophysics Data System (ADS)

    Kost, F. H.; Holmes, A. T.

    1985-09-01

    An investigation of transonic turbine blades designed by Rolls-Royce/Bristol concerning the aerodynamic penalties of coolant flow for two alternative cooling configurations is discussed. Rolls-Royce designed a blade with a thick trailing edge where the coolant is ejected through slots in the trailing edge and a second blade with a thin trailing edge where coolant is ejected through a row of holes on the pressure side and a row of holes on the suction side. Tests were performed in a plane cascade wind tunnel. The results indicate the sensitivity of the blade performance to cooling configuration and coolant flow rate. By combining measured data from blade surface and wake traverses it was possible to separate the various loss mechanisms. Therefore, the separate losses due to the momentum of the coolant, change of base pressure, and change of blade friction could be determined quantitatively as a function of coolant flow rate.

  5. Accident Flying Squad

    PubMed Central

    Snook, Roger

    1972-01-01

    This paper describes the organization, evaluation, and costing of an independently financed and operated accident flying squad. 132 accidents involving 302 casualties were attended, six deaths were prevented, medical treatment contributed to the survival of a further four, and the condition or comfort of many other casualties was improved. The calls in which survival was influenced were evenly distributed throughout the three-and-a-half-year survey and seven of the 10 so aided were over 16 and under 30 years of age, all 10 being in the working age group. The time taken to provide the service was not excessive and the expense when compared with the overall saving was very small. The scheme was seen to be equally suitable for basing on hospital or general practice or both, and working as an integrated team with the ambulance service. The use of specialized transport was found to be unnecessary. Other benefits of the scheme included use of the experience of attending accidents to ensure relevant and realistic training for emergency service personnel, and an appreciation of the effect of ambulance design on the patient. ImagesFIG. 1FIG. 4 PMID:5069642

  6. Optimization of the water chemistry of the primary coolant at nuclear power plants with VVER

    SciTech Connect

    Barmin, L. F.; Kruglova, T. K.; Sinitsyn, V. P.

    2005-01-15

    Results of the use of automatic hydrogen-content meter for controlling the parameter of 'hydrogen' in the primary coolant circuit of the Kola nuclear power plant are presented. It is shown that the correlation between the 'hydrogen' parameter in the coolant and the 'hydrazine' parameter in the makeup water can be used for controlling the water chemistry of the primary coolant system, which should make it possible to optimize the water chemistry at different power levels.

  7. ACHIEVING THE REQUIRED COOLANT FLOW DISTRIBUTION FOR THE ACCELERATOR PRODUCTION OF TRITIUM (APT) TUNGSTEN NEUTRON SOURCE

    SciTech Connect

    D. SIEBE; K. PASAMEHMETOGLU

    2000-11-01

    The Accelerator Production of Tritium neutron source consists of clad tungsten targets, which are concentric cylinders with a center rod. These targets are arranged in a matrix of tubes, producing a large number of parallel coolant paths. The coolant flow required to meet thermal-hydraulic design criteria varies with location. This paper describes the work performed to ensure an adequate coolant flow for each target for normal operation and residual heat-removal conditions.

  8. Viscosity of alumina nanoparticles dispersed in car engine coolant

    SciTech Connect

    Kole, Madhusree; Dey, T.K.

    2010-09-15

    The present paper, describes our experimental results on the viscosity of the nanofluid prepared by dispersing alumina nanoparticles (<50 nm) in commercial car coolant. The nanofluid prepared with calculated amount of oleic acid (surfactant) was tested to be stable for more than 80 days. The viscosity of the nanofluids is measured both as a function of alumina volume fraction and temperature between 10 and 50 C. While the pure base fluid display Newtonian behavior over the measured temperature, it transforms to a non-Newtonian fluid with addition of a small amount of alumina nanoparticles. Our results show that viscosity of the nanofluid increases with increasing nanoparticle concentration and decreases with increase in temperature. Most of the frequently used classical models severely under predict the measured viscosity. Volume fraction dependence of the nanofluid viscosity, however, is predicted fairly well on the basis of a recently reported theoretical model for nanofluids that takes into account the effect of Brownian motion of nanoparticles in the nanofluid. The temperature dependence of the viscosity of engine coolant based alumina nanofluids obeys the empirical correlation of the type: log ({mu}{sub nf}) = A exp(BT), proposed earlier by Namburu et al. (author)

  9. Development of mobile, on-site engine coolant recycling utilizing reverse-osmosis technology

    SciTech Connect

    Kughn, W.; Eaton, E.R.

    1999-08-01

    This paper presents the history of the development of self-contained, mobile, high-volume, engine coolant recycling by reverse osmosis (R/O). It explains the motivations, created by government regulatory agencies, to minimize the liability of waste generators who produce waste engine coolant by providing an engine coolant recycling service at the customer`s location. Recycling the used engine coolant at the point of origin minimizes the generators` exposure to documentation requirements, liability, and financial burdens by greatly reducing the volume of used coolant that must be hauled from the generator`s property. It describes the inherent difficulties of recycling such a highly contaminated, inconsistent input stream, such as used engine coolant, by reverse osmosis. The paper reports how the difficulties were addressed, and documents the state of the art in mobile R/O technology. Reverse osmosis provides a purified intermediate fluid that is reinhibited for use in automotive cooling systems. The paper offers a review of experiences in various automotive applications, including light-duty, medium-duty and heavy-duty vehicles operating on many types of fuel. The authors conclude that mobile embodiments of R/O coolant recycling technology provide finished coolants that perform equivalently to new coolants as demonstrated by their ability to protect vehicles from freezing, corrosion damage, and other cooling system related problems.

  10. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  11. Reactivity initiated accident test series Test RIA 1-4

    SciTech Connect

    Martinson, Z.R.; El-Genk, M.S.; Fukuda, S.K.; LaPointe, R.E.; Osetek, D.J.

    1980-05-01

    The Reactivity Initiated Accident (RIA) Test RIA 1-4, the first 9-rod fuel rod bundle RIA Test to be performed at BWR hot startup conditions, was completed on April 16, 1980. The test was performed in the Power Burst Facility (PBF). Objective for Test RIA 1-4 was to provide information regarding loss-of-coolable fuel rod geometry following a RIA event for a peak fuel enthalpy equivalent to the present licensing criteria of 280 cal/g. The most severe RIA is the postulated Boiling Water Reactor (BWR) control rod drop during reactor startup. Therefore the test was conducted at BWR hot startup coolant conditions (538 K, 6.45 MPa, 0.8 1/sec). The test sequence began with steady power operation to condition the fuel, establish a short-lived fission product inventory, and calibrate the calorimetric measurements and core power chambers, neutron flux and gamma flux detectors. The test train was removed from the in-pile tube (IPT) to replace one of the fuel rods with a nominally identical irradiated rod and twelve flux wire monitors. A 2.8 ms period power burst was then performed. Coolant flow measurements were made before and after the power burst to characterize the flow blockage that occurred as a result of fuel rod failure.

  12. [Chernobyl nuclear power plant accident and Tokaimura criticality accident].

    PubMed

    Takada, Jun

    2012-03-01

    It is clear from inspection of historical incidents that the scale of disasters in a nuclear power plant accident is quite low level overwhelmingly compared with a nuclear explosion in nuclear war. Two cities of Hiroshima and Nagasaki were destroyed by nuclear blast with about 20 kt TNT equivalent and then approximately 100,000 people have died respectively. On the other hand, the number of acute death is 30 in the Chernobyl nuclear reactor accident. In this chapter, we review health hazards and doses in two historical nuclear incidents of Chernobyl and Tokaimura criticality accident and then understand the feature of the radiation accident in peaceful utilization of nuclear power.

  13. Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

    SciTech Connect

    Roald A. Wigeland; James E. Cahalan

    2011-04-01

    Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

  14. Aircraft accident survivors as witnesses.

    PubMed

    Dodge, R E

    1983-02-01

    This is a study of the reliability of aircrash survivors as witnesses. Some of their statements are compared to known facts at the time of the crash, including the time of the accident and the weather conditions. Other facts are compared between the survivors, such as the mood of the passengers immediately post-crash. The KLM-Pan Am accident in the Canary Islands is used as the study accident. A suggestion for future use of survivors' statements is tendered.

  15. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    SciTech Connect

    Hagrman, D.T.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  16. Effect of coolant flow ejection on aerodynamic performance of low-aspect-ratio vanes. 2: Performance with coolant flow ejection at temperature ratios up to 2

    NASA Technical Reports Server (NTRS)

    Hass, J. E.; Kofskey, M. G.

    1977-01-01

    The aerodynamic performance of a 0.5 aspect ratio turbine vane configuration with coolant flow ejection was experimentally determined in a full annular cascade. The vanes were tested at a nominal mean section ideal critical velocity ratio of 0.890 over a range of primary to coolant total temperature ratio from 1.0 to 2.08 and a range of coolant to primary total pressure ratio from 1.0 to 1.4 which corresponded to coolant flows from 3.0 to 10.7 percent of the primary flow. The variations in primary and thermodynamic efficiency and exit flow conditions with circumferential and radial position were obtained.

  17. 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.

  18. 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. PMID:26568603

  19. 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

  20. Rear-end accident victims. Importance of understanding the accident.

    PubMed Central

    Sehmer, J. M.

    1993-01-01

    Family physicians regularly treat victims of rear-end vehicle accidents. This article describes how taking a detailed history of the accident and understanding the significance of the physical events is helpful in understanding and anticipating patients' morbidity and clinical course. Eight questions to ask patients are suggested to help physicians understand the severity of injury. PMID:8495140

  1. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

  2. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    SciTech Connect

    DeMuth, J. A.; Meier, W. R.; Jolodosky, A.; Frantoni, M.; Reyes, S.

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  3. Actively controlling coolant-cooled cold plate configuration

    DOEpatents

    Chainer, Timothy J.; Parida, Pritish R.

    2016-04-26

    Cooling apparatuses are provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The cooling apparatus includes the cold plate and a controller. The cold plate couples to one or more electronic components to be cooled, and includes an adjustable physical configuration. The controller dynamically varies the adjustable physical configuration of the cold plate based on a monitored variable associated with the cold plate or the electronic component(s) being cooled by the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the electronic component(s), and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the cold plate, the positioning of which may be adjusted based on the monitored variable.

  4. Radiation accident grips Goiania

    SciTech Connect

    Roberts, L.

    1987-11-20

    On 13 September two young scavengers in Goiania, Brazil, removed a stainless steel cylinder from a cancer therapy machine in an abandoned clinic, touching off a radiation accident second only to Chernobyl in its severity. On 18 September they sold the cylinder, the size of a 1-gallon paint can, to a scrap dealer for $25. At the junk yard an employee dismantled the cylinder and pried open the platinum capsule inside to reveal a glowing blue salt-like substance - 1400 curies of cesium-137. Fascinated by the luminescent powder, several people took it home with them. Some children reportedly rubbed in on their bodies like carnival glitter - an eerie image of how wrong things can go when vigilance over radioactive materials lapses. In all, 244 people in Goiania, a city of 1 million in central Brazil, were contaminated. The eventual toll, in terms of cancer or genetic defects, cannot yet be estimated. Parts of the city are cordoned off as radiation teams continue washing down buildings and scooping up radioactive soil. The government is also grappling with the political fallout from the accident.

  5. A method for measuring cooling air flow in base coolant passages of rotating turbine blades

    NASA Technical Reports Server (NTRS)

    Liebert, C. H.; Pollack, F. G.

    1975-01-01

    Method accurately determines actual coolant mass flow rate in cooling passages of rotating turbine blades. Total and static pressures are measured in blade base coolant passages. Mass flow rates are calculated from these measurements of pressure, measured temperature and known area.

  6. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2012-01-01 2012-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  7. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  8. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2014-01-01 2014-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  9. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2013-01-01 2013-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  10. Nonflammable coolants for space vehicle environmental control systems Compatibility of component materials with selected dielectric fluids.

    NASA Technical Reports Server (NTRS)

    Howard, R. T.; Korpolinski, T. S.; Mace, E. W.

    1971-01-01

    This paper summarizes a 4-year effort to evaluate and implement a nonflammable substitute coolant for application in the Saturn instrument unit (IU) environmental control system (ECS). Discussed are candidate material evaluations, detailed investigations of the properties of the coolant selected, and a summary of the implementation into a flight vehicle.

  11. MTR, TRA603. FOUNDATION PLAN, SECTION C THROUGH COOLANT WATER EXIT ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    MTR, TRA-603. FOUNDATION PLAN, SECTION C THROUGH COOLANT WATER EXIT TUNNEL ALONG NORTH SIDE AS IT RETURNS TO MAIN COOLANT TUNNEL LEAVING BUILDING TO THE NORTH. BLAW-KNOX 3150-803-35, 5/1950. INL INDEX NO. 531-0603-62-098-100591, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  12. Using automatic particle counting to monitor aluminum cold mill coolant{copyright}

    SciTech Connect

    Adkins, D.L.

    1995-08-01

    A comprehensive program of testing and evaluation of aluminum cold rolling coolant conditions has been conducted using an automatic particle counting instrument. The project had three objectives. First, there was a need to know at what level of coolant particle contamination is surface cleanliness of an aluminum sheet affected during the rolling process. Secondly, is application of particle counting technology a reliable tool for troubleshooting coolant filtration systems and finally, what are the advantages of analyzing rolling coolants for contamination levels? A testing program was designed and performed over a two-year period. The test results revealed that mineral seal and synthetic-type coolants can begin to affect aluminum sheet surface cleanliness levels when particle sizes greater than five microns are in excess of 10,000 particles power 100 milliliters of rolling coolant. After performing over 3,000 separate tests, it was very clean that particle count levels are direct indicators of how well a filtration facility is performing. Through the application of particle counting, a number of conditions in coolant filtration facilities can be readily detected. Such items as defective filter valving, torn or fractured filter cloth, damaged filter parts, improper equipment operation and many other factors will directly impact the operation of aluminum cold rolling coolant filters. 11 figs.

  13. Some peculiarities of checking the Topaz-2 system coolant filling quality

    SciTech Connect

    Ogloblin, B.G.; Svishchev, A.M.; Shalaev, A.I.

    1996-03-01

    This paper contains the analysis of validity of methods used for checking the Topaz-2 system coolant filling quality by a metering tank according to the mathematical model developed. A number of criteria is proposed for detecting occluded gas in the coolant loop. {copyright} {ital 1996 American Institute of Physics.}

  14. Blanket Module Boil-Off Times during a Loss-of-Coolant Accident - Case 0: with Beam Shutdown only

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document LBLOCA analyses for the Accelerator Production of Tritium primary blanket Heat Removal system. This report documents the analysis results of a LBLOCA where the accelerator beam is shut off without primary pump trips and neither the RHR nor the cavity flood systems operation.

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

    SciTech Connect

    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, 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. 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.

  16. Investigations of ice formation in the Space Shuttle Main Engine 0209 main injector coolant cavity

    NASA Technical Reports Server (NTRS)

    Richards, D. R.; Charklwick, D. M.

    1991-01-01

    Severe main combustion chamber wall and main injector baffle element deterioration occurred during tests of Space Shuttle Main Engine 0209. One of the possible causes considered is ice formation and blockage of coolant to these components, resulting from the mixing of leaking hot turbine exhaust gas (hydrogen rich steam) and hydrogen coolant in the injector coolant cavity. The plausibility of ice blockage is investigated through simple mixing calculations for hot gas and hydrogen, investigation of condensation and water droplet formation, calculation of the freezing times for droplets, and the prediction of ice layer thicknesses. It is concluded that condensation and droplet formation can occur, and small water droplets that form can freeze very quickly when in contact with the cold coolant cavity surfaces. Copnservative analysis predicts, however, that the maximum thickness of the ice layers formed is too small to result in significant blockage of the coolant flow.

  17. Engine coolant compatibility with the nonmetals found in automotive cooling systems

    SciTech Connect

    Greaney, J.P.; Smith, R.A.

    1999-08-01

    High temperature, short term immersion testing was used to determine the impact of propylene and ethylene glycol base coolants on the physical properties of a variety of elastomeric and thermoplastic materials found in automotive cooling systems. The materials tested are typically used in cooling system hoses, radiator end tanks, and water pump seals. Traditional phosphate or borate-buffered silicated coolants as well as extended-life organic acid formulations were included. A modified ASTM protocol was used to carry out the testing both in the laboratory and at an independent testing facility. Post-test fluid chemistry including an analysis of any solids which may have formed is also reported. Coolant impact on elastomer integrity as well as elastomer-induced changes in fluid chemistry were found to be independent of the coolant`s glycol base.

  18. Assessment of light water reactor fuel damage during a reactivity initiated accident

    SciTech Connect

    MacDonald, P.E.; Seiffert, S.L.; Martinson, Z.R.; McCardell, R.K.; Owen, D.E.; Fukuda, S.K.

    1980-01-01

    This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalpy of approximately 140 cal/g UO/sub 2/. Volume expansion of previously irradiated fuel upon melting may cause deformation and rupture of the cladding, and coolant channel blockage at higher peak enthalpies.

  19. German aircraft accident statistics, 1930

    NASA Technical Reports Server (NTRS)

    Weitzmann, Ludwig

    1932-01-01

    The investigation of all serious accidents, involving technical defects in the airplane or engine, is undertaken by the D.V.L. in conjunction with the imperial traffic minister and other interested parties. All accidents not clearly explained in the reports are subsequently cleared up.

  20. First Responders and Criticality Accidents

    SciTech Connect

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

    Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

  1. 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.

  2. 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.

  3. Characterization of uranium surfaces machined with aqueous propylene glycol-borax or perchloroethylene-mineral oil coolants

    SciTech Connect

    Cristy, S.S.; Bennett, R.K. Jr.; Dillon, J.J.; Richards, H.L.; Seals, R.D.; Byrd, V.R.

    1986-12-31

    The use of perchloroethylene (perc) as an ingredient in coolants for machining enriched uranium at the Oak Ridge Y-12 Plant has been discontinued because of environmental concerns. A new coolant was substituted in December 1985, which consists of an aqueous solution of propylene glycol with borax (sodium tetraborate) added as a nuclear poison and with a nitrite added as a corrosion inhibitor. Uranium surfaces machined using the two coolants were compared with respects to residual contamination, corrosion or corrosion potential, and with the aqueous propylene glycol-borax coolant was found to be better than that of enriched uranium machined with the perc-mineral oil coolant. The boron residues on the final-finished parts machined with the borax-containing coolant were not sufficient to cause problems in further processing. All evidence indicated that the enriched uranium surfaces machined with the borax-containing coolant will be as satisfactory as those machined with the perc coolant.

  4. Transport aircraft accident dynamics

    NASA Technical Reports Server (NTRS)

    Cominsky, A.

    1982-01-01

    A study was carried out of 112 impact survivable jet transport aircraft accidents (world wide) of 27,700 kg (60,000 lb.) aircraft and up extending over the last 20 years. This study centered on the effect of impact and the follow-on events on aircraft structures and was confined to the approach, landing and takeoff segments of the flight. The significant characteristics, frequency of occurrence and the effect on the occupants of the above data base were studied and categorized with a view to establishing typical impact scenarios for use as a basis of verifying the effectiveness of potential safety concepts. Studies were also carried out of related subjects such as: (1) assessment of advanced materials; (2) human tolerance to impact; (3) merit functions for safety concepts; and (4) impact analysis and test methods.

  5. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    SciTech Connect

    Ball, S.J. )

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

  6. Flow in serpentine coolant passages with trip strips

    NASA Technical Reports Server (NTRS)

    Tse, D. G.-N.

    1995-01-01

    Under the subject contract, an effort is being conducted at Scientific Research Associates, Inc. (SRA) to obtain flow field measurements in the coolant passage of a rotating turbine blade with ribbed walls, both in the stationary and rotating frames. The data obtained will be used for validation of computational tools and assessment of turbine blade cooling strategies. The configuration of the turbine blade passage model is given, and the measuring plane locations are given. The model has a four-pass passage with three 180 turns. This geometry was chosen to allow analyses of the velocity measurements corresponding to the heat transfer results obtained by Wagner. Two passes of the passage have a rectangular cross-section of 1.0 in x 0.5 in. Another two passes have a square cross-section of 0.5 in x 0.5 in. Trips with a streamwise pitch to trip height (P/e) = 5 and trip height to coolant passage width (e/Z) = 0.1, were machined along the leading and trailing walls. These dimensions are typical of those used in turbine blade coolant passages. The trips on these walls are staggered by the half-pitch. The trips are skewed at +/- 45 deg, and this allows the effect of trip orientation to be examined. Experiments will be conducted with flow entering the model through the 1.0 in x 0.5 in rectangular passage (Configuration C) and the 0.5 in x 0. 5 in square passage (Configuration D) to examine the effect of passage aspect ratio. Velocity measurements were obtained with a Reynolds number (Re) of 25,000, based on the hydraulic diameter of and bulk mean velocity in the half inch square passage. The coordinate system used in presenting the results for configurations C and D, respectively, is shown. The first, second and third passes of the passage will be referred to as the first, second and third passages, respectively, in later discussion. Streamwise distance (x) from the entrance is normalized by the hydraulic diameter (D). Vertical (y) and tangential (z) distances are

  7. Preliminary analysis of graphite dust releasing behavior in accident for HTR

    SciTech Connect

    Peng, W.; Yang, X. Y.; Yu, S. Y.; Wang, J.

    2012-07-01

    The behavior of the graphite dust is important to the safety of High Temperature Gas-cooled Reactors. This study investigated the flow of graphite dust in helium mainstream. The analysis of the stresses acting on the graphite dust indicated that gas drag played the absolute leading role. Based on the understanding of the importance of gas drag, an experimental system is set up for the research of dust releasing behavior in accident. Air driven by centrifugal fan is used as the working fluid instead of helium because helium is expensive, easy to leak which make it difficult to seal. The graphite particles, with the size distribution same as in HTR, are added to the experiment loop. The graphite dust releasing behavior at the loss-of-coolant accident will be investigated by a sonic nozzle. (authors)

  8. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    SciTech Connect

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

    1997-04-01

    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 branching 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.

  9. Purification of liquid metal systems with sodium coolant from oxygen using getters

    NASA Astrophysics Data System (ADS)

    Kozlov, F. A.; Konovalov, M. A.; Sorokin, A. P.

    2016-05-01

    For increasing the safety and economic parameters of nuclear power stations (NPSs) with sodium coolant, it was decided to install all systems contacting radioactive sodium, including purification systems of circuit I, in the reactor vessel. The performance and capacity of cold traps (CTs) (conventional element of coolant purification systems) in these conditions are limited by their volume. It was proposed to use hot traps (HTs) in circuit I for coolant purification from oxygen. It was demonstrated that, at rated parameters of the installation when the temperature of the coolant streamlining the getter (gas absorber) is equal to 550°C, the hot trap can provide the required coolant purity. In shutdown modes at 250-300°C, the performance of the hot trap is reduced by four orders of magnitude. Possible HT operation regimes for shutdown modes and while reaching rated parameters were proposed and analyzed. Basic attention was paid to purification modes at power rise after commissioning and accidental contamination of the coolant when the initial oxygen concentration in it reached 25 mln-1. It was demonstrated that the efficiency of purification systems can be increased using HTs with the getter in the form of a foil or granules. The possibility of implementing the "fast purification" mode in which the coolant is purified simultaneously with passing over from the shutdown mode to the rated parameters was substantiated.

  10. Polymer electrolyte membrane fuel cell performance degradation by coolant leakage and recovery

    NASA Astrophysics Data System (ADS)

    Jung, Ju Hae; Kim, Se Hoon; Hur, Seung Hyun; Joo, Sang Hoon; Choi, Won Mook; Kim, Junbom

    2013-03-01

    Coolant leakage leads to decrease in performance during the operation of electric vehicles which make use of polymer electrolyte membrane fuel cells (PEMFC). This study examines the effects of various coolant leak conditions in 3-cell stack and single cell. The experimental results show that an irreversible reduction in performance occurs after coolant injection into the anode side of the stack. Poisoning of carbon monoxide (CO) on the platinum (Pt) catalyst is caused by electro-oxidation reaction of EG. Water cleaning is selected because CO poisoning is desorbed to reaction with water molecules. Performance is quickly reduced when the interval between coolant injections is short. Performance reduction is indicated by the experimental results for the gas diffusion layer (GDL) and the membrane electrode assembly (MEA). It shows that performance of the MEA with the GDL exposed to coolant decreased, but it is recovered after water cleaning. In contrast, results for performance of the MEA exposed to coolant for long time could not be reversed after water cleaning. Therefore, we propose that performance degradation of coolant leak on the Pt catalyst surface and GDL can be recovered by the water cleaning simply without disassembly of stack.

  11. Recycling used engine coolant using high-volume stationary, multiple technology equipment

    SciTech Connect

    Haddock, M.E.; Eaton, E.R.

    1999-08-01

    Recycling used engine coolant has become increasingly desirable due to two significant factors. First, engine coolant frequently merits designation as a hazardous waste under the Federal Clean Water Act. Federal and some state environmental protection agencies have instituted strict regulation of the disposal of used engine coolant. In some cases, the disposal of engine coolant requires imposition of waste disposal fees and surcharges. Secondly, ethylene glycol, the principal cost component of engine coolant, has experienced dramatic price fluctuations and occasional shortages in supply. Therefore, there are both environmental and economic pressures to recycle engine coolant and recover the ethylene glycol component in an efficient and cost-effective manner. This paper discusses a multistage apparatus and a process for recycling used engine coolant that employs a combination of filtration, centrifugation (hydrocyclone separation), dissolved air flotation, nanofiltration, reverse osmosis, continuous deionization, and ion exchange processes for separating ethylene glycol and water from used engine coolant. The engine coolant is prefiltered through a series of filters. The filters remove particulate contaminates. This filtered fluid is then subjected to dissolved air flotation and centrifugation to remove petroleum. Then it is heated and pressurized prior to being passed over a series of two different sets of semipermeable membranes. The membrane technologies separate the feed stream into a permeate solution of ethylene glycol and water and a concentrate waste solution. The concentrate solution is returned to a concentrate tank for continuous circulation through the apparatus. The permeate solution is subjected to final refining by continuous deionization followed by a cation and anion ion exchange polishing process. The continuous deionization reduces ionic contaminants, and the ion exchange system eliminates any ionic contaminants left by the previous purification

  12. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    NASA Technical Reports Server (NTRS)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  13. A review of criticality accidents

    SciTech Connect

    Stratton, W R; Smith, D R

    1989-03-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Forty-one accidental power transients are reviewed. In each case where available, enough detail is given to help visualize the physical situation, the cause or causes of the accident, the history and characteristics of the transient, the energy release, and the consequences, if any, to personnel and property. Excursions associated with large power reactors are not included in this study, except that some information on the major accident at the Chernobyl reactor in April 1986 is provided in the Appendix. 67 refs., 21 figs., 2 tabs.

  14. [Prevention of bicycle accidents].

    PubMed

    Zwipp, H; Barthel, P; Bönninger, J; Bürkle, H; Hagemeister, C; Hannawald, L; Huhn, R; Kühn, M; Liers, H; Maier, R; Otte, D; Prokop, G; Seeck, A; Sturm, J; Unger, T

    2015-04-01

    For a very precise analysis of all injured bicyclists in Germany it would be important to have definitions for "severely injured", "seriously injured" and "critically injured". By this, e.g., two-thirds of surgically treated bicyclists who are not registered by the police could become available for a general analysis. Elderly bicyclists (> 60 years) are a minority (10 %) but represent a majority (50 %) of all fatalities. They profit most by wearing a helmet and would be less injured by using special bicycle bags, switching on their hearing aids and following all traffic rules. E-bikes are used more and more (145 % more in 2012 vs. 2011) with 600,000 at the end of 2011 and are increasingly involved in accidents but still have a lack of legislation. So even for pedelecs 45 with 500 W and a possible speed of 45 km/h there is still no legislative demand for the use of a protecting helmet. 96 % of all injured cyclists in Germany had more than 0.5 ‰ alcohol in their blood, 86 % more than 1.1 ‰ and 59 % more than 1.7 ‰. Fatalities are seen in 24.2 % of cases without any collision partner. Therefore the ADFC calls for a limit of 1.1 ‰. Some virtual studies conclude that integrated sensors in bicycle helmets which would interact with sensors in cars could prevent collisions or reduce the severity of injury by stopping the cars automatically. Integrated sensors in cars with opening angles of 180° enable about 93 % of all bicyclists to be detected leading to a high rate of injury avoidance and/or mitigation. Hanging lamps reduce with 35 % significantly bicycle accidents for children, traffic education for children and special trainings for elderly bicyclists are also recommended as prevention tools. As long as helmet use for bicyclists in Germany rates only 9 % on average and legislative orders for using a helmet will not be in force in the near future, coming up campaigns seem to be necessary to be promoted by the Deutscher

  15. [Prevention of bicycle accidents].

    PubMed

    Zwipp, H; Barthel, P; Bönninger, J; Bürkle, H; Hagemeister, C; Hannawald, L; Huhn, R; Kühn, M; Liers, H; Maier, R; Otte, D; Prokop, G; Seeck, A; Sturm, J; Unger, T

    2015-04-01

    For a very precise analysis of all injured bicyclists in Germany it would be important to have definitions for "severely injured", "seriously injured" and "critically injured". By this, e.g., two-thirds of surgically treated bicyclists who are not registered by the police could become available for a general analysis. Elderly bicyclists (> 60 years) are a minority (10 %) but represent a majority (50 %) of all fatalities. They profit most by wearing a helmet and would be less injured by using special bicycle bags, switching on their hearing aids and following all traffic rules. E-bikes are used more and more (145 % more in 2012 vs. 2011) with 600,000 at the end of 2011 and are increasingly involved in accidents but still have a lack of legislation. So even for pedelecs 45 with 500 W and a possible speed of 45 km/h there is still no legislative demand for the use of a protecting helmet. 96 % of all injured cyclists in Germany had more than 0.5 ‰ alcohol in their blood, 86 % more than 1.1 ‰ and 59 % more than 1.7 ‰. Fatalities are seen in 24.2 % of cases without any collision partner. Therefore the ADFC calls for a limit of 1.1 ‰. Some virtual studies conclude that integrated sensors in bicycle helmets which would interact with sensors in cars could prevent collisions or reduce the severity of injury by stopping the cars automatically. Integrated sensors in cars with opening angles of 180° enable about 93 % of all bicyclists to be detected leading to a high rate of injury avoidance and/or mitigation. Hanging lamps reduce with 35 % significantly bicycle accidents for children, traffic education for children and special trainings for elderly bicyclists are also recommended as prevention tools. As long as helmet use for bicyclists in Germany rates only 9 % on average and legislative orders for using a helmet will not be in force in the near future, coming up campaigns seem to be necessary to be promoted by the Deutscher

  16. Single-beam thermal lens measurement of thermal diffusivity of engine coolants

    NASA Astrophysics Data System (ADS)

    George, Nibu A.; Thomas, Nibu B.; Chacko, Kavya; T, Neethu V.; Hussain Moidu, Haroon; Piyush, K.; David, Nitheesh M.

    2015-04-01

    Automobile engine coolant liquids are commonly used for efficient heat transfer from the engine to the surroundings. In this work we have investigated the thermal diffusivity of various commonly available engine coolants in Indian automobile market. We have used single beam laser induced thermal lens technique for the measurements. Engine coolants are generally available in concentrated solution form and are recommended to use at specified dilution. We have investigated the samples in the entire recommended concentration range for the use in radiators. While some of the brands show an enhanced thermal diffusivity compared to pure water, others show slight decrease in thermal diffusivity.

  17. Computation of Space Shuttle high-pressure cryogenic turbopump ball bearing two-phase coolant flow

    NASA Technical Reports Server (NTRS)

    Chen, Yen-Sen

    1990-01-01

    A homogeneous two-phase fluid flow model, implemented in a three-dimensional Navier-Stokes solver using computational fluid dynamics methodology is described. The application of the model to the analysis of the pump-end bearing coolant flow of the high-pressure oxygen turbopump of the Space Shuttle main engine is studied. Results indicate large boiling zones and hot spots near the ball/race contact points. The extent of the phase change of the liquid oxygen coolant flow due to the frictional and viscous heat fluxes near the contact areas has been investigated for the given inlet conditions of the coolant.

  18. Cracked shaft detection on large vertical nuclear reactor coolant pump

    NASA Technical Reports Server (NTRS)

    Jenkins, L. S.

    1985-01-01

    Due to difficulty and radiation exposure associated with examination of the internals of large commercial nuclear reactor coolant pumps, it is necessary to be able to diagnose the cause of an excessive vibration problem quickly without resorting to extensive trial and error efforts. Consequently, it is necessary to make maximum use of all available data to develop a consistent theory which locates the problem area in the machine. This type of approach was taken at Three Mile Island, Unit #1, in February 1984 to identify and locate the cause of a continuously climbing vibration level of the pump shaft. The data gathered necessitated some in-depth knowledge of the pump internals to provide proper interpretation and avoid misleading conclusions. Therefore, the raw data included more than just the vibration characteristics. Pertinent details of the data gathered is shown and is necessary and sufficient to show that the cause of the observed vibration problem could logically only be a cracked pump shaft in the shaft overhang below the pump bearing.

  19. Compatibility Issues for a High Temperature Dual Coolant Blanket

    SciTech Connect

    Pint, Bruce A

    2007-01-01

    One proposed U.S. test blanket module (TBM) for ITER uses ferritic-martensitic alloys with both eutectic Pb-Li and He coolants at {approx}475 C. In order for this blanket concept to operate at higher temperatures ({approx}750 C) for a DEMO-type reactor, several Pb-Li compatibility issues need to be addressed. A SiC/SiC composite flow channel insert is proposed to reduce the steel dissolution rate (and the magnetohydrodynamic pressure drop). Prior capsule testing examined dense, high-purity SiC in Pb-Li at 800-1200 C and found detectable levels of Si in the Pb-Li after 2,000h at 1100 C and 1,000h at 1200 C. Current capsule experiments are examining several different SiC/SiC composite materials at 1000 C. Another issue involves Pb-Li transport between the first wall and heat exchanger. Aluminide coatings on type 316 stainless steel and Al-containing alloys capable of forming an external alumina scale have been studied in capsule experiments at 700 and 800 C for 1,000h. Model aluminide coatings made by chemical vapor deposition reduced the dissolution rate for 316SS at 800 C by a factor of 50.

  20. Tables of thermodynamic properties of helium magnet coolant. Revision A

    SciTech Connect

    McAshan, M.

    1992-07-01

    The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: ``Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10{sup 8} Pa``, Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923--1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: ``Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres``, Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it.

  1. Tables of thermodynamic properties of helium magnet coolant

    SciTech Connect

    McAshan, M.

    1992-07-01

    The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10{sup 8} Pa'', Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923--1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres'', Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it.

  2. Tables of thermodynamic properties of helium magnet coolant, revision A

    NASA Astrophysics Data System (ADS)

    McAshan, M.

    1992-07-01

    The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: 'Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10(exp 8) Pa', Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923-1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: 'Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres', Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it.

  3. [Orofacial injuries in skateboard accidents].

    PubMed

    Frohberg, U; Bonsmann, M

    1992-04-01

    In a clinical study, 25 accidents involving injuries by a fall with a skateboard were investigated and classified in respect of epidemiology, accident mechanism and injury patterns in the facial region. Accident victims are predominantly boys between 7 and 9 years of age. A multiple trauma involving the teeth and the dental system in general and the soft parts of the face is defined as a characteristic orofacial injury pattern in skateboard accidents. The high proportion of damage to the front teeth poses problems of functional and aesthetic rehabilitation necessitating long-term treatment courses in children and adolescents. Effective prevention of facial injuries may be possible by evolving better facial protection systems and by creating areas of playgrounds where skateboarders can practise safely.

  4. Aircraft accidents : method of analysis

    NASA Technical Reports Server (NTRS)

    1929-01-01

    This report on a method of analysis of aircraft accidents has been prepared by a special committee on the nomenclature, subdivision, and classification of aircraft accidents organized by the National Advisory Committee for Aeronautics in response to a request dated February 18, 1928, from the Air Coordination Committee consisting of the Assistant Secretaries for Aeronautics in the Departments of War, Navy, and Commerce. The work was undertaken in recognition of the difficulty of drawing correct conclusions from efforts to analyze and compare reports of aircraft accidents prepared by different organizations using different classifications and definitions. The air coordination committee's request was made "in order that practices used may henceforth conform to a standard and be universally comparable." the purpose of the special committee therefore was to prepare a basis for the classification and comparison of aircraft accidents, both civil and military. (author)

  5. Spine Immobilizer for Accident Victims

    NASA Technical Reports Server (NTRS)

    Vykukal, H. C.; Lampson, K.

    1983-01-01

    Proposed conformal bladder filled with tiny spheres called "microballoons," enables spine of accident victim to be rapidly immobilized and restrained and permit victim to be safely removed from accident scene in extremely short time after help arrives. Microballoons expand to form rigid mass when pressure within bladder is less than ambient. Bladder strapped to victim is also strapped to rescue chair. Void between bladder and chair is filled with cloth wedges.

  6. Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

    SciTech Connect

    Wenfeng Liu; Kazimi, Mujid S.

    2006-07-01

    This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0--56 MW d/kg), peak fuel enthalpy deposit (70--190 cal/g), degree of subcooling (0--80 deg. C), and extent of oxidation (0--25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface

  7. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  8. 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.

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

    SciTech Connect

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

    1991-01-01

    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 fuel 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.

  10. The experience in handling of lead-bismuth coolant contaminated by Polonium-210

    SciTech Connect

    Pankratov, D.V.; Gromov, B.F.; Solodjankin, M.A.

    1993-12-31

    During exploitation of lead-bismuth cooled reactors a wide experience in handling of radioactive coolant containing polonium has been gained. By 1990 total time of this reactor operation has reached approximately 60 reactor years.

  11. Computer code for predicting coolant flow and heat transfer in turbomachinery

    NASA Technical Reports Server (NTRS)

    Meitner, Peter L.

    1990-01-01

    A computer code was developed to analyze any turbomachinery coolant flow path geometry that consist of a single flow passage with a unique inlet and exit. Flow can be bled off for tip-cap impingement cooling, and a flow bypass can be specified in which coolant flow is taken off at one point in the flow channel and reintroduced at a point farther downstream in the same channel. The user may either choose the coolant flow rate or let the program determine the flow rate from specified inlet and exit conditions. The computer code integrates the 1-D momentum and energy equations along a defined flow path and calculates the coolant's flow rate, temperature, pressure, and velocity and the heat transfer coefficients along the passage. The equations account for area change, mass addition or subtraction, pumping, friction, and heat transfer.

  12. Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

    NASA Astrophysics Data System (ADS)

    Alekseev, V. V.; Kovalev, Yu. P.; Kalyakin, S. G.; Kozlov, F. A.; Kumaev, V. Ya.; Kondrat'ev, A. S.; Matyukhin, V. V.; Pirogov, E. P.; Sergeev, G. P.; Sorokin, A. P.; Torbenkova, I. Yu.

    2013-05-01

    Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit's purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.

  13. Thermal transfer structures coupling electronics card(s) to coolant-cooled structure(s)

    DOEpatents

    David, Milnes P; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Parida, Pritish R; Schmidt, Roger R

    2014-12-16

    Cooling apparatuses and coolant-cooled electronic systems are provided which include thermal transfer structures configured to engage with a spring force one or more electronics cards with docking of the electronics card(s) within a respective socket(s) of the electronic system. A thermal transfer structure of the cooling apparatus includes a thermal spreader having a first thermal conduction surface, and a thermally conductive spring assembly coupled to the conduction surface of the thermal spreader and positioned and configured to reside between and physically couple a first surface of an electronics card to the first surface of the thermal spreader with docking of the electronics card within a socket of the electronic system. The thermal transfer structure is, in one embodiment, metallurgically bonded to a coolant-cooled structure and facilitates transfer of heat from the electronics card to coolant flowing through the coolant-cooled structure.

  14. Hot-Gas-Slide and Coolant-Side Heat Transfer in Liquid Rocket Engine Combustors

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Luong, Van

    1994-01-01

    The objectives of this article are to develop a multidisciplinary, computational methodology to predict the hot-gas-side and coolant-side heat transfer in film cooling assisted, regeneratively cooled liquid rocket engine combustors, and to use it in parametric studies to recommend optimized design of the coolant channels for a developmental combustor. An integrated numerical model which incorporates computational fluid dynamics (CFD) for the hot-gas thermal environment, and thermal analysis for the liner and coolant channels, was developed. This integrated CFD/thermal model was validated by comparing predicted heat fluxes with those of hot-firing test and industrial design methods for a 40-k calorimeter thrust chamber and the Space Shuttle Main Engine main combustion chamber. Parametric studies were performed for the advanced main combustion chamber to find a strategy for a proposed coolant channel design.

  15. Measurement of the Coolant Channel Temperatures and Pressures of a Cooled Radial-Inflow Turbine

    NASA Technical Reports Server (NTRS)

    Dicicco, L. Danielle; Nowlin, Brent C.; Tirres, Lizet

    1994-01-01

    Instrumentation has been installed on the surface of a cooled radial-inflow turbine. Thermocouples and miniature integrated sensor pressure transducers were installed to measure steady state coolant temperatures, blade wall temperatures, and coolant pressures. These measurements will eventually be used to determine the heat transfer characteristics of the rotor. This paper will describe the procedures used to install and calibrate the instrumentation and the testing methods followed. A limited amount of data will compare the measured values to the predicted values.

  16. Influence of coolant tube curvature on film cooling effectiveness as detected by infrared imagery

    NASA Technical Reports Server (NTRS)

    Papell, S. S.; Graham, R. W.; Cageao, R. P.

    1979-01-01

    Thermal film cooling footprints observed by infrared imagery from straight, curved, and looped coolant tube geometries are compared. It was hypothesized that the differences in secondary flow and in the turbulence structure of flow through these three tubes should influence the mixing properties between the coolant and the main stream. A flow visualization tunnel, an infrared camera and detector, and a Hilsch tube were employed to test the hypothesis.

  17. Accident Tolerant Fuel Analysis

    SciTech Connect

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and

  18. [Accidents of fulguration].

    PubMed

    Virenque, C; Laguerre, J

    1976-01-01

    Fulguration, first electric accident in which the man was a victim, is to day better known. A clap of thunder is decomposed in two elements: lightning, and thunder. Lightning is caused by an electrical discharge, either within a cloud, or between two clouds, or, above all, between a cloud and the surface of the ground. Experimental equipments owned by the French Electricity Company and by the Atomic Energy Commission, have allowed to photograph lightnings and to measure certain physical characteristics (Intensity variable between 25 to 100 kA, voltage variable between 20 to 1 000 kV). The frequency of storms was learned: the isokeraunic level, in France, is about 20, meaning that thunder is heard twenty days during one year. Man may be stricken by thunder by direct hit, by sudden bursting, by earth current, or through various conductors. The electric charge which reached him may go to the earth directly by contact with the ground or may dissipate in the air through a bony promontory (elbow). The total number of victims, "wounded" or deceased, is not now known by statistics. Death comes by insulation breakdown of one of several anatomic cephalic formations: skull, meninx, brain. Many various lesions may happen in survivors: loss of consciousness, more or less long, sensorial or motion deficiencies. All these signs are momentary and generally reversible. Besides one may observe much more intense lesions on the skin: burns and, over all, characteristic aborescence (skin effect by high frequency current). The heart is protected, contrarily to what happens with industrial electrocution. The curative treatment is merely symptomatic : reanimation, surgery for burns or associated traumatic lesions. A prevention is researched to help the lonely man, in the country or in the mountains in the houses (lightning conductor, Faraday cage), in vehicles (aircraft, cars, ships). The mysterious and unforseeable character of lightning still stays, leaving a door opened for numerous

  19. Accident tolerant fuel analysis

    SciTech Connect

    Smith, Curtis; Chichester, Heather; Johns, Jesse; Teague, Melissa; Tonks, Michael Idaho National Laboratory; Youngblood, Robert

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and

  20. Elevated-pressure mixed-coolants Joule Thomson cryocooling

    NASA Astrophysics Data System (ADS)

    Maytal, B.-Z.; Nellis, G. F.; Klein, S. A.; Pfotenhauer, J. M.

    2006-01-01

    This paper explores the potential of mixed coolants at elevated pressures for Joule-Thomson cryocooling. A numerical model of a Joule-Thomson cryocooler is developed that is capable of simulating operation with mixtures of up to 9 components consisting of hydrocarbons, non-flammable halogenated refrigerants, and inert gases. The numerical model is integrated with a genetic optimization algorithm, which has a high capability for convergence in an environment of discontinuities, constraints and local optima. The genetic optimization algorithm is used to select the optimal mixture compositions that separately maximizes following two objective functions at each elevated pressure for 80, 90 and 95 K cryocooling: the molar specific cooling capacity (the highest attainable is 3200 J/mol) and the produced cooling capacity per thermal conductance which is a measure of the compactness of the recuperator. The optimized cooling capacity for a non-flammable halogenated refrigerant mixture is smaller than for a hydrocarbon mixture; however, the cooling capacity of the two types of mixtures approach one another as pressure becomes higher. The coefficient of performance, the required heat transfer area and the effect of the number of components in the mixture is investigated as a function of the pressure. It is shown that mixtures with more components provide a higher cooling capacity but require larger recuperative heat exchangers. Optimized mixtures for 90 K cryocooling have similar cooling capacity as those for 80 K. Optimized compactness for 80 K is about 50% higher than can be achieved by pure nitrogen. For 90 K, no mixture provides a more compact recuperator than can be achieved using pure argon. The results are discussed in the context of potential applications for closed and open cycle cryocoolers.

  1. Correct numerical simulation of a two-phase coolant

    NASA Astrophysics Data System (ADS)

    Kroshilin, A. E.; Kroshilin, V. E.

    2016-02-01

    Different models used in calculating flows of a two-phase coolant are analyzed. A system of differential equations describing the flow is presented; the hyperbolicity and stability of stationary solutions of the system is studied. The correctness of the Cauchy problem is considered. The models' ability to describe the following flows is analyzed: stable bubble and gas-droplet flows; stable flow with a level such that the bubble and gas-droplet flows are observed under and above it, respectively; and propagation of a perturbation of the phase concentration for the bubble and gas-droplet media. The solution of the problem about the breakdown of an arbitrary discontinuity has been constructed. Characteristic times of the development of an instability at different parameters of the flow are presented. Conditions at which the instability does not make it possible to perform the calculation are determined. The Riemann invariants for the nonlinear problem under consideration have been constructed. Numerical calculations have been performed for different conditions. The influence of viscosity on the structure of the discontinuity front is studied. Advantages of divergent equations are demonstrated. It is proven that a model used in almost all known investigating thermohydraulic programs, both in Russia and abroad, has significant disadvantages; in particular, it can lead to unstable solutions, which makes it necessary to introduce smoothing mechanisms and a very small step for describing regimes with a level. This does not allow one to use efficient numerical schemes for calculating the flow of two-phase currents. A possible model free from the abovementioned disadvantages is proposed.

  2. Reactor coolant pump testing using motor current signatures analysis

    SciTech Connect

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  3. Coolant-Flow Calibrations of Three Simulated Porous Gas-Turbine Blades

    NASA Technical Reports Server (NTRS)

    Esger, Jack B.; Lea, Alfred L.

    1951-01-01

    An investigation was conducted at the NACA Lewis laboratory to determine whether simulated porous gas-turbine blades fabricated by the Eaton Manufacturing Company of Cleveland, Ohio would be satisfactory with respect to coolant flow for application in gas-turbine engines. These blades simulated porous turbine blades by forcing the cooling air onto the blade surface through a large number of chordwise openings or slits between laminations of sheet metal or wire. This type of surface has a finite number of openings, whereas a porous surface has an almost infinite number of smaller openings for the coolant flow. The investigation showed that a blade made of sheet-metal laminations stacked on a support member that passed up through the coolant passage was completely unsatisfactory because of extremely poor coolant flow distribution over the blade surface. The flow distribution for two wire-wound blades was more uniform, but the pressure drop between the coolant supply pressure and the local pressure on the outside of the blades was too low by a factor ranging from 3 to 3.5 for the required coolant flow rates. The pressure drop could be increased by forcing the wires closer together during blade fabrication.

  4. Analysis of a water-coolant leak into a very high-temperature vitrification chamber.

    SciTech Connect

    Felicione, F. S.

    1998-06-11

    A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed.

  5. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

    NASA Technical Reports Server (NTRS)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

  6. Impact of the propylene glycol-water-borax coolant on material recovery operations

    SciTech Connect

    Duerksen, W.K.; Taylor, P.A.

    1983-05-01

    The reaction of the propylene glycol-water-borax coolant with nitric acid has now been studied in some detail. This document is intended to provide a summary of the results. Findings are summarized under nine headings. Tests have also been conducted to determine if the new coolant would have any adverse effects on the uranium recycle systems. Experiments were scientifically designed after observation of the production operations so that accurate response to the immediate production concerns could be provided. Conclusions from these studies are: formation of glycol nitrates is very improbable; the reaction of concentrated (70%) nitric acid with pure propylene glycol is very violent and hazardous; dilution of the nitric acid-glycol mixture causes a drastic decrease in the rate and intensity of the reaction; the mechanism of the nitric acid propylene glycol reaction is autocatalytic in nitrous acid; no reaction is observed between coolant and 30% nitric acid unless the solution is heated; the coolant reacts fairly vigorously with 55% nitric acid after a concentration-dependent induction time; experiments showed that the dissolution of uranium chips that had been soaked in coolant proceeded at about the same rate as if the chips had not previously contacted glycol; thermodynamic calculations show that the enthalpy change (heat liberated) by the reaction of nitric acid (30%) with propylene glycol is smaller than if the same amount of nitric acid reacted with uranium. Each of these conclusions is briefly discussed. The effect of new coolant on uranium recycle operations is then briefly discussed.

  7. 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

  8. Validation and verification of RELAP5 for Advanced Neutron Source accident analysis: Part I, comparisons to ANSDM and PRSDYN codes

    SciTech Connect

    Chen, N.C.J.; Ibn-Khayat, M.; March-Leuba, J.A.; Wendel, M.W.

    1993-12-01

    As part of verification and validation, the Advanced Neutron Source reactor RELAP5 system model was benchmarked by the Advanced Neutron Source dynamic model (ANSDM) and PRSDYN models. RELAP5 is a one-dimensional, two-phase transient code, developed by the Idaho National Engineering Laboratory for reactor safety analysis. Both the ANSDM and PRSDYN models use a simplified single-phase equation set to predict transient thermal-hydraulic performance. Brief descriptions of each of the codes, models, and model limitations were included. Even though comparisons were limited to single-phase conditions, a broad spectrum of accidents was benchmarked: a small loss-of-coolant-accident (LOCA), a large LOCA, a station blackout, and a reactivity insertion accident. The overall conclusion is that the three models yield similar results if the input parameters are the same. However, ANSDM does not capture pressure wave propagation through the coolant system. This difference is significant in very rapid pipe break events. Recommendations are provided for further model improvements.

  9. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    SciTech Connect

    O`Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-12-31

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/Pbq (2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river.

  10. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    SciTech Connect

    O'Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-01-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/Pbq (2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river.

  11. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect

    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 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 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

  12. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect

    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 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 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

  13. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    SciTech Connect

    Gauntt, Randall O.; Goldmann, Andrew S.; Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  14. Lower head creep rupture failure analysis associated with alternative accident sequences of the Three Mile Island Unit 2

    SciTech Connect

    Sang Lung, Chan

    2004-07-01

    The objective of this lower head creep rupture analysis is to assess the current version of MELCOR 1.8.5-RG against SCDAP/RELAP5 MOD 3.3kz. The purpose of this assessment is to investigate the current MELCOR in-vessel core damage progression phenomena including the model for the formation of a molten pool. The model for stratified molten pool natural heat transfer will be included in the next MELCOR release. Presently, MELCOR excludes the gap heat-transfer model for the cooling associated with the narrow gap between the debris and the lower head vessel wall. All these phenomenological models are already treated in SCDAP/RELAP5 using the COUPLE code to model the heat transfer of the relocated debris with the lower head based on a two-dimensional finite-element-method. The assessment should determine if current MELCOR capabilities adequately cover core degradation phenomena appropriate for the consolidated MELCOR code. Inclusion of these features should bring MELCOR much closer to a state of parity with SCDAP/RELAP5 and is a currently underway element in the MELCOR code consolidation effort. This assessment deals with the following analysis of the Three Mile Island Unit 2 (TMI-2) alternative accident sequences. The TMI-2 alternative accident sequence-1 includes the continuation of the base case of the TMI-2 accident with the Reactor Coolant Pumps (RCP) tripped, and the High Pressure Injection System (HPIS) throttled after approximately 6000 s accident time, while in the TMI-2 alternative accident sequence-2, the reactor coolant pumps is tripped after 6000 s and the HPIS is activated after 12,012 s. The lower head temperature distributions calculated with SCDAP/RELAP5 are visualized and animated with open source visualization freeware 'OpenDX'. (author)

  15. Identification of severe accident uncertainties

    SciTech Connect

    Rivard, J.B.; Behr, V.L.; Easterling, R.G.; Griesmeyer, J.M.; Haskin, F.E.; Hatch, S.W.; Kolaczkowski, A.M.; Lipinski, R.J.; Sherman, M.P.; Taig, A.R.

    1984-09-01

    Understanding of severe accidents in light-water reactors is currently beset with uncertainty. Because the uncertainties that are present limit the capability to analyze the progression and possible consequences of such accidents, they restrict the technical basis for regulatory actions by the US Nuclear Regulatory Commission (NRC). It is thus necessary to attempt to identify the sources and quantify the influence of these uncertainties. As a part of ongoing NRC severe-accident programs at Sandia National Laboratories, a working group was formed to pool relevant knowledge and experience in assessing the uncertainties attending present (1983) knowledge of severe accidents. This initial report of the Severe Accident Uncertainty Analysis (SAUNA) working group has as its main goal the identification of a consolidated list of uncertainties that affect in-plant processes and systems. Many uncertainties have been identified. A set of key uncertainties summarizes many of the identified uncertainties. Quantification of the influence of these uncertainties, a necessary second step, is not attempted in the present report, although attempts are made qualitatively to demonstrate the relevance of the identified uncertainties.

  16. Severe accident simulation at Olkiuoto

    SciTech Connect

    Tirkkonen, H.; Saarenpaeae, T.; Cliff Po, L.C.

    1995-09-01

    A personal computer-based simulator was developed for the Olkiluoto nuclear plant in Finland for training in severe accident management. The generic software PCTRAN was expanded to model the plant-specific features of the ABB Atom designed BWR including its containment over-pressure protection and filtered vent systems. Scenarios including core heat-up, hydrogen generation, core melt and vessel penetration were developed in this work. Radiation leakage paths and dose rate distribution are presented graphically for operator use in diagnosis and mitigation of accidents. Operating on an graphically for operator use in diagnosis and mitigation of accidents. Operating on an 486 DX2-66, PCTRAN-TVO achieves a speed about 15 times faster than real-time. A convenient and user-friendly graphic interface allows full interactive control. In this paper a review of the component models and verification runs are presented.

  17. The child accident repeater: a review.

    PubMed

    Jones, J G

    1980-04-01

    The child accident repeater is defined as one who has at least three accidents that come to medical attention within a year. The accident situation has features in common with those of the child who has a single accident through simple "bad luck", but other factors predispose him to repeated injury. In the child who has a susceptible personality, a tendency for accident repetition may be due to a breakdown in adjustment to a stressful environment. Prevention of repeat accidents should involve the usual measures considered appropriate for all children as well as an attempt to provide treatment of significant maladjustment and modification of a stressful environment.

  18. Thermal criteria to compare fast reactors coolants for the intermediate loop

    SciTech Connect

    Saez, Manuel; Rodriguez, Gilles

    2007-07-01

    Fast Breeder Reactors (FBR) are typically using a liquid metal as the primary coolant. Up to now, sodium is the referenced coolant for all large-scale FBR, but lead and sodium-potassium alloy have both also been used successfully for smaller rigs. The French Atomic Energy Commission (CEA) has an extensive experience and significant expertise in Sodium cooled Fast Reactors (SFR) over the past 40 years of R and D and feedback experiments. Some improvements are needed on the SFR to meet the Generation IV goals, and in particular the safety and the reliability through the intermediate loop coolant. As sodium reacts exo-thermically with air and water and to eliminate the drawback of the water-sodium interaction when a steam generator tube is ruptured, CEA is involved in a substantial effort in order to investigate the interest to use an alternative coolant than sodium in the intermediate loop. This paper presents the main thermal criteria to compare Fast Reactors coolants for the intermediate loop under natural and forced convection. Neutronics considerations are not taken into account for the intermediate loop coolant. Transport, transfer and energetic criteria are analysed in the field of turbulent flows. Criteria are applied to the following potential coolant candidates: sodium, lithium, tin, bismuth, lead, lead-bismuth alloy, lead-lithium alloy, gallium, indium, potassium and sodium-potassium alloy. According to this thermal analysis, the gallium as heat transfer agent for the intermediate loop is considered as a promising candidate. For the discussion of the applicability of the gallium as heat transfer agent for the intermediate loop, a limited thermal hydraulic pre-sizing of a steam generator is undertaken using simple engineering methods implemented in COPERNIC code, a CEA tool dedicated to reactor systems pre-sizing. (authors)

  19. Investigation of the Challenger Accident

    NASA Technical Reports Server (NTRS)

    1986-01-01

    The work of the Presidential Commission on the Space Shuttle Challenger Accident (hereafter referred to as the Rogers Commission) and the work of the National Aeronautics and Space Administration in investigating the causes of the accident were reviewed. In addition to reviewing the five volumes of the Rogers Commission, the entire direct on-line Rogers Commission data base, which included full-text and document retrieval capability was also reviewed. The findings and recommendations contained also include materials submitted for the record, staff investigations, interviews, and trips.

  20. [Diving accidents. Emergency treatment of serious diving accidents].

    PubMed

    Schröder, S; Lier, H; Wiese, S

    2004-11-01

    Decompression injuries are potentially life-threatening incidents mainly due to a rapid decline in ambient pressure. Decompression illness (DCI) results from the presence of gas bubbles in the blood and tissue. DCI may be classified as decompression sickness (DCS) generated from the liberation of gas bubbles following an oversaturation of tissues with inert gas and arterial gas embolism (AGE) mainly due to pulmonary barotrauma. People working under hyperbaric pressure, e.g. in a caisson for general construction under water, and scuba divers are exposed to certain risks. Diving accidents can be fatal and are often characterized by organ dysfunction, especially neurological deficits. They have become comparatively rare among professional divers and workers. However, since recreational scuba diving is gaining more and more popularity there is an increasing likelihood of severe diving accidents. Thus, emergency staff working close to areas with a high scuba diving activity, e.g. lakes or rivers, may be called more frequently to a scuba diving accident. The correct and professional emergency treatment on site, especially the immediate and continuous administration of normobaric oxygen, is decisive for the outcome of the accident victim. The definitive treatment includes rapid recompression with hyperbaric oxygen. The value of adjunctive medication, however, remains controversial.

  1. 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

  2. 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

  3. Flatness Control Using Roll Coolant Based on Predicted Flatness Variation in Cold Rolling Mills

    NASA Astrophysics Data System (ADS)

    Dohmae, Yukihiro; Okamura, Yoshihide

    Flatness control for cold rolling mills is one of the important technologies for improving of product quality and productivity. In particular, poor flatness leads to strip tearing in the extreme case and, moreover, it significantly reduces productivity. Therefore, various flatness control system has been developed. The main actuators for flatness control are classified into two types; one is mechanical equipment such as roll bender, the other is roll coolant, which controls thermal expansion of roll. Flatness variation such as center buckle or edge wave is mainly controlled by mechanical actuator which has high response characteristics. On another front, flatness variation of local zone can be controlled by roll coolant although one's response is lower than the response of mechanical actuator. For accomplishing good flatness accuracy in cold rolling mills, it is important to improve the performance of coolant control moreover. In this paper, a new coolant control method based on flatness variation model is described. In proposed method, the state of coolant spray on or off is selected to minimize the flatness deviation by using predicted flatness variation. The effectiveness of developed system has been demonstrated by application in actual plant.

  4. Coolant-side heat-transfer rates for a hydrogen-oxygen rocket and a new technique for data correlation

    NASA Technical Reports Server (NTRS)

    Schacht, R. L.; Quentmeyer, R. J.

    1973-01-01

    An experimental investigation was conducted to determine the coolant-side, heat transfer coefficients for a liquid cooled, hydrogen-oxygen rocket thrust chamber. Heat transfer rates were determined from measurements of local hot gas wall temperature, local coolant temperature, and local coolant pressure. A correlation incorporating an integration technique for the transport properties needed near the pseudocritical temperature of liquid hydrogen gives a satisfactory prediction of hot gas wall temperatures.

  5. 49 CFR 230.22 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Requirements § 230.22 Accident reports. In the case of an accident due to failure, from any cause, of a steam locomotive boiler or any part or appurtenance thereof, resulting in serious injury or death to one or...

  6. Advanced accident sequence precursor analysis level 1 models

    SciTech Connect

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K.; Schroeder, J.A.; Siu, N.O.

    1996-03-01

    INEL has been involved in the development of plant-specific Accident Sequence Precursor (ASP) models for the past two years. These models were developed for use with the SAPHIRE suite of PRA computer codes. They contained event tree/linked fault tree Level 1 risk models for the following initiating events: general transient, loss-of-offsite-power, steam generator tube rupture, small loss-of-coolant-accident, and anticipated transient without scram. Early in 1995 the ASP models were revised based on review comments from the NRC and an independent peer review. These models were released as Revision 1. The Office of Nuclear Regulatory Research has sponsored several projects at the INEL this fiscal year to further enhance the capabilities of the ASP models. Revision 2 models incorporates more detailed plant information into the models concerning plant response to station blackout conditions, information on battery life, and other unique features gleaned from an Office of Nuclear Reactor Regulation quick review of the Individual Plant Examination submittals. These models are currently being delivered to the NRC as they are completed. A related project is a feasibility study and model development of low power/shutdown (LP/SD) and external event extensions to the ASP models. This project will establish criteria for selection of LP/SD and external initiator operational events for analysis within the ASP program. Prototype models for each pertinent initiating event (loss of shutdown cooling, loss of inventory control, fire, flood, seismic, etc.) will be developed. A third project concerns development of enhancements to SAPHIRE. In relation to the ASP program, a new SAPHIRE module, GEM, was developed as a specific user interface for performing ASP evaluations. This module greatly simplifies the analysis process for determining the conditional core damage probability for a given combination of initiating events and equipment failures or degradations.

  7. Delta launch vehicle accident investigation

    NASA Astrophysics Data System (ADS)

    1986-03-01

    The text of the testimony given by several witnesses during the House hearings on the Delta launch vehicle accident of May 3, 1986 is given. Pre-launch procedures, failure analysis, the possibility of sabotage, and design and testing are among the topics discussed.

  8. Numerical analysis of the hot-gas-side and coolant-side heat transfer in liquid rocket engine combustors

    NASA Technical Reports Server (NTRS)

    Wang, T. S.; Luong, V.

    1992-01-01

    The objectives of this paper are to develop computational methods to predict the hot-gas-side and coolant-side heat transfer, and to use these methods in parametric studies to recommend optimized design of the coolant channels for regeneratively cooled liquid rocket engine combustors. An integrated numerical model which incorporates computational fluid dynamics (CFD) for the hot-gas thermal environment, and thermal analysis for the coolant channels, was developed. The mode was validated by comparing predicted heat fluxes with those of hot-firing test and industrial design methods. Parametric studies were performed to find a strategy for optimized combustion chamber coolant channel design.

  9. Determinants of injuries in passenger vessel accidents.

    PubMed

    Yip, Tsz Leung; Jin, Di; Talley, Wayne K

    2015-09-01

    This paper investigates determinants of crew and passenger injuries in passenger vessel accidents. Crew and passenger injury equations are estimated for ferry, ocean cruise, and river cruise vessel accidents, utilizing detailed data of individual vessel accidents that were investigated by the U.S. Coast Guard during the time period 2001-2008. The estimation results provide empirical evidence (for the first time in the literature) that crew injuries are determinants of passenger injuries in passenger vessel accidents.

  10. Determinants of injuries in passenger vessel accidents.

    PubMed

    Yip, Tsz Leung; Jin, Di; Talley, Wayne K

    2015-09-01

    This paper investigates determinants of crew and passenger injuries in passenger vessel accidents. Crew and passenger injury equations are estimated for ferry, ocean cruise, and river cruise vessel accidents, utilizing detailed data of individual vessel accidents that were investigated by the U.S. Coast Guard during the time period 2001-2008. The estimation results provide empirical evidence (for the first time in the literature) that crew injuries are determinants of passenger injuries in passenger vessel accidents. PMID:26070017

  11. 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.

  12. 48 CFR 836.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Accident prevention. 836... prevention. The contracting officer must insert the clause at 852.236-87, Accident Prevention, in solicitations and contracts for construction that contain the clause at FAR 52.236-13, Accident Prevention....

  13. 48 CFR 636.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Accident prevention. 636... CONTRACTING CONSTRUCTION AND ARCHITECT-ENGINEER CONTRACTS Contract Clauses 636.513 Accident prevention. (a) In... contracting activities shall insert DOSAR 652.236-70, Accident Prevention, in lieu of FAR clause...

  14. 48 CFR 1836.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 6 2010-10-01 2010-10-01 true Accident prevention. 1836... 1836.513 Accident prevention. The contracting officer must insert the clause at 1852.223-70, Safety and Health, in lieu of FAR clause 52.236-13, Accident Prevention, and its Alternate I....

  15. 28 CFR 301.106 - Repetitious accidents.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Repetitious accidents. 301.106 Section 301.106 Judicial Administration FEDERAL PRISON INDUSTRIES, INC., DEPARTMENT OF JUSTICE INMATE ACCIDENT COMPENSATION General § 301.106 Repetitious accidents. If an inmate worker is involved in successive...

  16. 28 CFR 301.106 - Repetitious accidents.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 28 Judicial Administration 2 2011-07-01 2011-07-01 false Repetitious accidents. 301.106 Section 301.106 Judicial Administration FEDERAL PRISON INDUSTRIES, INC., DEPARTMENT OF JUSTICE INMATE ACCIDENT COMPENSATION General § 301.106 Repetitious accidents. If an inmate worker is involved in successive...

  17. Nanomaterials for efficiently lowering the freezing point of anti-freeze coolants.

    PubMed

    Hong, Haiping; Zheng, Yingsong; Roy, Walter

    2007-09-01

    In this paper, we report, for the first time, the effect of the lowered freezing point in a 50% water/50% anti-freeze coolant (PAC) or 50% water/50% ethylene glycol (EG) solution by the addition of carbon nanotubes and other particles. The experimental results indicated that the nano materials are much more efficient (hundreds fold) in lowering the freezing point than the regular ionic materials (e.g., NaCl). The possible explanation for this interesting phenomenon is the colligative property of fluid and relative small size of nano material. It is quite certain that the carbon nanotubes and metal oxide nano particles could be a wonderful candidate for the nano coolant application because they could not only increase the thermal conductivity, but also efficiently lower the freezing point of traditional coolants. PMID:18019146

  18. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    NASA Astrophysics Data System (ADS)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  19. New technique of mechanical pulse jet coolant delivery system for minimal quantity lubricant (MQL) operation

    NASA Astrophysics Data System (ADS)

    Fazli, Nik; Badrul, Hamdi, M.

    2011-01-01

    Minimum quantity lubrication (MQL) machining is one of the promising solutions to decrease the cutting fluid consumption requirements. This paper describes MQL machining in a range of lubricant consumption of 2.0ml/h, which is 10-100 times smaller than the normal consumption adopted in industries. MQL machining in this range is called pulse jet coolant delivery system in this paper. A specially designed system was used for concentrating small amounts of lubricant onto the cutting interface. The performance of concentrated spraying of lubricant in pulse jet coolant delivery system design was actual simulated and compared with others systems. The concentrated spraying of lubricant with a specially designed of system was quite effective in increasing tool life in the pulse jet coolant delivery system range.

  20. Effects of coolant parameters on steady state temperature distribution in phospheric-acid fuel cell electrode

    NASA Technical Reports Server (NTRS)

    Alkasab, K. A.; Abdul-Aziz, A.

    1991-01-01

    The influence of thermophysical properties and flow rate on the steady-state temperature distribution in a phosphoric-acid fuel cell electrode plate was experimentally investigated. An experimental setup that simulates the operating conditions prevailing in a phosphoric-acid fuel cell stack was used. The fuel cell cooling system utilized three types of coolants to remove excess heat generated in the cell electrode and to maintain a reasonably uniform temperature distribution in the electrode plate. The coolants used were water, engine oil, and air. These coolants were circulated at Reynolds number ranging from 1165 to 6165 for water; 3070 to 6864 for air; and 15 to 79 for oil. Experimental results are presented.

  1. System and method for determining coolant level and flow velocity in a nuclear reactor

    DOEpatents

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  2. Measurement of Coolant in a Flat Heat Pipe Using Neutron Radiography

    NASA Astrophysics Data System (ADS)

    Mizuta, Kei; Saito, Yasushi; Goshima, Takashi; Tsutsui, Toshio

    A newly developed flat heat pipe FGHPTM (Morex Kiire Co.) was experimentally investigated by using neutron radiography. The test sample of the FGHP heat spreader was 65 × 65 × 2 mm3 composed of several etched copper plates and pure water was used as the coolant. Neutron radiography was performed at the E-2 port of the Kyoto University Research Reactor (KUR). The coolant distributions in the wick area of the FGHP and its heat transfer characteristics were measured at heating conditions. Experimental results show that the coolant distributions depend slightly on its installation posture and that the liquid thickness in the wick region remains constant with increasing heat input to the FGHP. In addition, it is found that the wick surface does not dry out even in the vertical posture at present experimental conditions.

  3. Identification of flow patterns by neutron noise analysis during actual coolant boiling in thin rectangular channels

    SciTech Connect

    Kozma, R.; van Dam, H.; Hoogenboom, J.E. )

    1992-10-01

    The primary objective of this paper is to introduce results of coolant boiling experiments in a simulated materials test reactor-type fuel assembly with plate fuel in an actual reactor environment. The experiments have been performed in the Hoger Onderwijs Reactor (HOR) research reactor at the Interfaculty Reactor Institute, Delft, The Netherlands. In the analysis, noise signals of self-powered neutron detectors located in the neighborhood of the boiling region and thermocouple in the channel wall and in the coolant are used. Flow patterns in the boiling coolant have been identified by means of analysis of probability density functions and power spectral densities of neutron noise. It is shown that boiling has an oscillating character due to partial channel blockage caused by steam slugs generated periodically between the plates. The observed phenomenon can serve as a basis for a boiling detection method in reactors with plate-type fuels.

  4. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    NASA Astrophysics Data System (ADS)

    Rezania, A.; Rosendahl, L. A.

    2012-06-01

    The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The three-dimensional governing equations for the fluid flow and the heat transfer are solved using the finite-volume method for a wide range of pressure drop laminar flows along the heat sink. The temperature and the mass flow rate distribution in the heat sink are discussed. The results, which are in good agreement with previous computational studies, show that using suggested heat sink configurations reduces the coolant pumping power in the system.

  5. Uncertainty quantification for accident management using ACE surrogates

    SciTech Connect

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

    2012-07-01

    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 to 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)

  6. Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

    SciTech Connect

    Morris, E.E.; Dunn, F.E.; Simms, R.; Gruber, E.E.

    1985-01-01

    The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of fission-gas pressure on the motion of the upper fuel pin segment following disruption. For the nominal sodium-void worth, fission-gas-driven fuel compaction did not adversely affect the outcome of the transient. When uncertainties in the sodium-void worth were considered, however, it was found that if fuel compaction occurs, loss-of-flow driven transient overpower phenomenology could not be precluded.

  7. Potassium iodide for thyroid blockade in a reactor accident: administrative policies that govern its use.

    PubMed

    Becker, D V; Zanzonico, P

    1997-04-01

    structures with large steel and concrete shells and multiple redundancy of core cooling mechanisms. These successfully prevented the release of major amounts of radionuclides in the Three Mile Island partial loss-of-primary coolant accident in 1979. The Chernobyl accident, in a different type of reactor that is common in Eastern Europe, did not have effective outer shell containment and released almost 50 MCi of 131I compared to the 20 Ci of 131I released at Three Mile Island. Such accidents have precipitated extensive re-evaluation of the design and safety devices of all operating reactors. However, a major contributing factor to the accidents was human error and considerable efforts must be made to train plant operators so they have a better understanding of reactor operation and use of safety mechanisms. PMID:9133683

  8. Potassium iodide for thyroid blockade in a reactor accident: administrative policies that govern its use.

    PubMed

    Becker, D V; Zanzonico, P

    1997-04-01

    structures with large steel and concrete shells and multiple redundancy of core cooling mechanisms. These successfully prevented the release of major amounts of radionuclides in the Three Mile Island partial loss-of-primary coolant accident in 1979. The Chernobyl accident, in a different type of reactor that is common in Eastern Europe, did not have effective outer shell containment and released almost 50 MCi of 131I compared to the 20 Ci of 131I released at Three Mile Island. Such accidents have precipitated extensive re-evaluation of the design and safety devices of all operating reactors. However, a major contributing factor to the accidents was human error and considerable efforts must be made to train plant operators so they have a better understanding of reactor operation and use of safety mechanisms.

  9. A Thermoelectric Generator Using Engine Coolant for Light-Duty Internal Combustion Engine-Powered Vehicles

    NASA Astrophysics Data System (ADS)

    Kim, Shiho; Park, Soonseo; Kim, Sunkook; Rhi, Seok-Ho

    2011-05-01

    We proposed and fabricated a thermoelectric generator (TEG) using the engine water coolant of passenger vehicles. The experimental results revealed that the maximum output power from the proposed thermoelectric generator was ~75 W, the calculated thermoelectric module efficiency of the TEG was ~2.1%, and the overall efficiency of electric power generation from the waste heat of the engine coolant was ~0.3% in the driving mode at 80 km/h. The conventional radiator can thus be replaced by the proposed TEG without additional devices or redesign of the engine water cooling system of the existing radiator.

  10. Flow boiling with enhancement devices for cold plate coolant channel design

    NASA Technical Reports Server (NTRS)

    Boyd, Ronald D., Sr.

    1989-01-01

    A research program to study the effect of enhancement devices on flow boiling heat transfer in coolant channels, which are heated either from the top side or uniformly, is discussed. Freon 11 is the working fluid involved. The specific objectives are: (1) examine the variations in both the mean and local (axial and circumferential) heat transfer coefficients for a circular coolant channel with either smooth walls or with both a twisted tape and spiral finned walls, (2) examine the effect channel diameter (and the length-to-diameter aspect ratio) variations for the smooth wall channel, and (3) develop an improved data reduction analysis.

  11. Modeling the flow of liquid-metal coolant in the T-shaped mixer

    NASA Astrophysics Data System (ADS)

    Kashinsky, O. N.; Lobanov, P. D.; Kurdyumov, A. S.; Pribaturin, N. A.

    2016-05-01

    The results of experimental studies on the structure of the temperature field in the tube cross section at the flow of liquid-metal coolant in a T-shaped mixer are presented. Experiments were carried out using the Rose alloy as the working fluid. To determine temperature distribution on the test section wall, infrared thermography was used; to determine temperature distribution in the channel cross section, a mobile thermocouple was used. Considerable temperature maldistribution in the mixing zone of liquid flows with different temperatures on the tube wall and in the coolant melt is shown.

  12. Coolant and ambient temperature control for chillerless liquid cooled data centers

    DOEpatents

    Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

    2016-02-02

    Cooling control methods include measuring a temperature of air provided to a plurality of nodes by an air-to-liquid heat exchanger, measuring a temperature of at least one component of the plurality of nodes and finding a maximum component temperature across all such nodes, comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold, and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the plurality of nodes based on the comparisons.

  13. Design, manufacture, and test of coolant pump-motor assembly for Brayton power conversion system

    NASA Technical Reports Server (NTRS)

    Gabacz, L. E.

    1973-01-01

    The design, development, fabrication, and testing of seven coolant circulating pump-motor assemblies are discussed. The pump-motor assembly is driven by the nominal 44.4-volt, 400-Hz, 3-phase output of a nominal 56-volt dc input inverter. The pump-motor assembly will be used to circulate Dow Corning 200 liquid coolant for use in a Brayton cycle space power system. The pump-motor assembly develops a nominal head of 70 psi at 3.7 gpm with an over-all efficiency of 26 percent. The design description, drawings, photographs, reliability results, and developmental and acceptance test results are included.

  14. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    SciTech Connect

    Young, Michael F.

    1999-05-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

  15. Integrated Fuel-Coolant Interaction (IFCI 6.0) code. User`s manual

    SciTech Connect

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User`s Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks.

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

    SciTech Connect

    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.

  17. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    SciTech Connect

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. )

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

  18. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    SciTech Connect

    Banati, J.

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  19. Medical management principles for radiation accidents.

    PubMed

    Meineke, Viktor; van Beuningen, Dirk; Sohns, Torsten; Fliedner, Theodor M

    2003-03-01

    The medical management of radiation accidents requires intensive planning and action. This article looks at the medical management of recent radiation accidents to derive principles for structuring and organizing the treatment of patients who may have radiation-induced health impairments. Although the radiation accidents in Tokai-mura, Japan and Lilo, Georgia were small-scale accidents, they illustrate important and characteristic symptoms and clinical developments. There are lessons to be learned and conclusions to be drawn for the military medical officers concerned with problems of medical management after radiation accidents.

  20. Exploratory analysis of Spanish energetic mining accidents.

    PubMed

    Sanmiquel, Lluís; Freijo, Modesto; Rossell, Josep M

    2012-01-01

    Using data on work accidents and annual mining statistics, the paper studies work-related accidents in the Spanish energetic mining sector in 1999-2008. The following 3 parameters are considered: age, experience and size of the mine (in number of workers) where the accident took place. The main objective of this paper is to show the relationship between different accident indicators: risk index (as an expression of the incidence), average duration index for the age and size of the mine variables (as a measure of the seriousness of an accident), and the gravity index for the various sizes of mines (which measures the seriousness of an accident, too). The conclusions of this study could be useful to develop suitable prevention policies that would contribute towards a decrease in work-related accidents in the Spanish energetic mining industry. PMID:22721539

  1. The Concept of Accident Proneness: A Review

    PubMed Central

    Froggatt, Peter; Smiley, James A.

    1964-01-01

    The term accident proneness was coined by psychological research workers in 1926. Since then its concept—that certain individuals are always more likely than others to sustain accidents, even though exposed to equal risk—has been questioned but seldom seriously challenged. This article describes much of the work and theory on which this concept is based, details the difficulties encountered in obtaining valid information and the interpretative errors that can arise from the examination of imperfect data, and explains why accident proneness became so readily accepted as an explanation of the facts. A recent hypothesis of accident causation, namely that a person's accident liability may vary from time to time, is outlined, and the respective abilities of this and of accident proneness to accord with data from the more reliable literature are examined. The authors conclude that the hypothesis of individual variation in liability is more realistic and in better agreement with the data than is accident proneness. PMID:14106130

  2. Features of temperature control of fuel element cladding for pressurized water nuclear reactor ``WWER-1000'' while simulating reactor accidents

    NASA Astrophysics Data System (ADS)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-01

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones "cladding-junction" and "junction-coolant". The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ˜5% of maximum value by the final moment of the stage of linear increase in the temperature.

  3. Possible Methods to Estimate Core Location in a Beyond-Design-Basis Accident at a GE BWR with a Mark I Containment Stucture

    SciTech Connect

    Walston, S; Rowland, M; Campbell, K

    2011-07-27

    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 in 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.

  4. Accident/Mishap Investigation System

    NASA Technical Reports Server (NTRS)

    Keller, Richard; Wolfe, Shawn; Gawdiak, Yuri; Carvalho, Robert; Panontin, Tina; Williams, James; Sturken, Ian

    2007-01-01

    InvestigationOrganizer (IO) is a Web-based collaborative information system that integrates the generic functionality of a database, a document repository, a semantic hypermedia browser, and a rule-based inference system with specialized modeling and visualization functionality to support accident/mishap investigation teams. This accessible, online structure is designed to support investigators by allowing them to make explicit, shared, and meaningful links among evidence, causal models, findings, and recommendations.

  5. Endwall Vortex Effects on Turbulent Dispersion of Film Coolant in a Turbine Vane Cascade

    NASA Astrophysics Data System (ADS)

    Yapa, Sayuri D.; Elkins, Christopher J.; Eaton, John K.

    2013-11-01

    Turbine flows include strong secondary flows due to flow turning. The dominant flow feature is the passage vortex, located in the corner between the endwall and the suction surface of the airfoil. This vortex may have a strong effect on scalar transport in the turbine wake. Experiments were conducted to examine the dispersion of coolant emitted along the trailing edge of the airfoil. 3D velocity and concentration measurements were made using magnetic resonance imaging to study turbulent mixing in a realistic film-cooled nozzle vane cascade. The passage vortex has large effects on the flow features in the vane wake and on coolant mixing. A shear layer is created on the vane's suction side and interacts with the passage vortex after shedding from the trailing edge. The resulting vortex pattern forces the coolant jet into a highly distorted shape. A key question is how this distortion affects the turbulent diffusion of coolant. The 3D MRI-based velocity and concentration measurements allows for estimation of turbulent diffusivity. Control volumes are defined using a streamtube that is defined beginning just downstream of the trailing edge. The turbulent diffusivity is determined by integrating the Reynolds-averaged advection-diffusion equation over these control volumes. This work was sponsored by the Army Research Office and General Electric.

  6. MTR, TRA603. SUBBASEMENT FLOOR PLAN. INLET/OUTLET TUNNELS FOR COOLANT WATER ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    MTR, TRA-603. SUB-BASEMENT FLOOR PLAN. INLET/OUTLET TUNNELS FOR COOLANT WATER (NORTH SIDE) AND AIR (SOUTH SIDE). RABBIT CANAL AND BULKHEADS. SUMPS AND DRAINS. BLAW-KNOX 3150-3-7, 3/1950. INL INDEX NO. 531-0603-00-098-100006, REV. 4. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  7. Method and apparatus for removing iodine from a nuclear reactor coolant

    DOEpatents

    Cooper, Martin H.

    1980-01-01

    A method and apparatus for removing iodine-131 and iodine-125 from a liquid sodium reactor coolant. Non-radioactive iodine is dissolved in hot liquid sodium to increase the total iodine concentration. Subsequent precipitation of the iodine in a cold trap removes both the radioactive iodine isotopes as well as the non-radioactive iodine.

  8. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, Wayne H.; Christiansen, David W.

    1987-01-01

    A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  9. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, Wayne H.; Christiansen, David W.

    1987-05-05

    A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  10. RAPID INDUCTION OF HETEROGENEOUS ICE NUCLEATION IN A BIOLOGICALLY COMPATIBLE COOLANT

    PubMed Central

    Lampe, Joshua; Bull, Diana; Becker, Lance

    2012-01-01

    Hypothermia is gaining recognition as an important medical treatment. To treat local cases of injury such as stroke, or certain surgical procedures, there is a need to induce local hypothermia. To treat shock or cardiac arrest survivors, there is a need to rapidly induce global hypothermia. Rapid induction of hypothermia has been achieved in animal research, but it has yet to be achieved clinically using a simple, widely practicable method. The clinical need for therapeutic hypothermia represents an engineering opportunity to develop an easy to use coolant that is sterile, biologically compatible, and maximizes coolant heat capacity. Here we present an initial characterization of a prototype platform technology designed to create a sterile, biologically compatible, high heat capacity coolant that has the potential to be used in all of these clinical applications. The coolant is a specially processed micro-particulate ice saline slurry, that can be easily pumped into a patient through surgical tubing, syringes, or minimally invasive surgical instruments. The device induces heterogeneous ice nucleation in a saline stream that has been super-cooled from room temperature to a temperature below the saline freezing point. Currently, the device begins continuous production of ice slurry that contains ~30 % ice by mass within 10 minutes. The nominal ice particle diameter is smaller than 100 μm. This work represents a significant first step toward addressing clinical needs for rapid human cooling. PMID:25954121

  11. An Improved Design for Air Removal from Aerospace Fluid Loop Coolant Systems

    NASA Technical Reports Server (NTRS)

    Ritchie, Stephen M. C.; Holladay, Jon B.; Holt, J. Mike; Clark, Dallas W.

    2003-01-01

    Aerospace applications with requirements for large capacity heat removal (launch vehicles, platforms, payloads, etc.) typically utilize a liquid coolant fluid as a transport media to increase efficiency and flexibility in the vehicle design. An issue with these systems however, is susceptibility to the presence of noncondensable gas (NCG) or air. The presence of air in a coolant loop can have numerous negative consequences, including loss of centrifugal pump prime, interference with sensor readings, inhibition of heat transfer, and coolant blockage to remote systems. Hardware ground processing to remove this air is also cumbersome and time consuming which continuously drives recurring costs. Current systems for maintaining the system free of air are tailored and have demonstrated only moderate success. An obvious solution to these problems is the development and advancement of a passive gas removal device, or gas trap, that would be installed in the flight cooling system simplifying the initial coolant fill procedure and also maintaining the system during operations. The proposed device would utilize commercially available membranes thus increasing reliability and reducing cost while also addressing both current and anticipated applications. In addition, it maintains current pressure drop, water loss, and size restrictions while increasing tolerance for pressure increases due to gas build-up in the trap.

  12. Modeling Film-Coolant Flow Characteristics at the Exit of Shower-Head Holes

    NASA Technical Reports Server (NTRS)

    Garg, Vijay K.; Gaugler, R. E. (Technical Monitor)

    2000-01-01

    The coolant flow characteristics at the hole exits of a film-cooled blade are derived from an earlier analysis where the hole pipes and coolant plenum were also discretized. The blade chosen is the VKI rotor with three staggered rows of shower-head holes. The present analysis applies these flow characteristics at the shower-head hole exits. A multi-block three-dimensional Navier-Stokes code with Wilcox's k-omega model is used to compute the heat transfer coefficient on the film-cooled turbine blade. A reasonably good comparison with the experimental data as well as with the more complete earlier analysis where the hole pipes and coolant plenum were also gridded is obtained. If the 1/7th power law is assumed for the coolant flow characteristics at the hole exits, considerable differences in the heat transfer coefficient on the blade surface, specially in the leading-edge region, are observed even though the span-averaged values of h (heat transfer coefficient based on T(sub o)-T(sub w)) match well with the experimental data. This calls for span-resolved experimental data near film-cooling holes on a blade for better validation of the code.

  13. Assembly for facilitating inservice inspection of a reactor coolant pump rotor

    DOEpatents

    Veronesi, Luciano

    1990-01-01

    A reactor coolant pump has an outer casing with an internal cavity holding a coolant and a rotor rotatably mounted in the cavity within the coolant. An assembly for permitting inservice inspection of the pump rotor without first draining the coolant from the casing cavity is attached to an end of the pump. A cylindrical bore is defined through the casing in axial alignment with an end of pump rotor and opening into the internal cavity. An extension attached on the rotor end and rotatable therewith has a cylindrical coupler member extending into the bore. An outer end of the coupler member has an element configured to receive a tool for performance of inservice rotor inspection. A hollow cylindrical member is disposed in the bore and surrounds the coupler member. The cylindrical member is slidably movable relative to the coupler member along the bore between a retracted position wherein the cylindrical member is stored for normal pump operation and an extended position wherein the cylindrical member is extended for permitting inservice rotor inspection. A cover member is detachably and sealably attached to the casing across the bore for closing the bore and retaining the cylindrical member at its retracted position for normal pump operation. Upon detachment of the cover member, the cylindrical member can be extended to permit inservice rotor inspection.

  14. ETR COOLING TOWER PUMP HOUSE, TRA645. FOUR SECONDARY COOLANT PUMPS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ETR COOLING TOWER PUMP HOUSE, TRA-645. FOUR SECONDARY COOLANT PUMPS ARE ARRANGED IN A ROW. IN REAR ARE THREE SHUTDOWN EMERGENCY PUMPS. INL NEGATIVE NO. 56-4176. Jack L. Anderson, Photographer, 12/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  15. Effect of glycol-based coolants on the suppression and recovery of platinum fuel cell electrocatalysts

    NASA Astrophysics Data System (ADS)

    Garsany, Yannick; Dutta, Sreya; Swider-Lyons, Karen E.

    2012-10-01

    We use cyclic and rotating disk electrode voltammetry to study glycol-based coolant formulations to show that individual constituents have either negligible or significant poisoning effects on the nanoscale Pt/carbon catalysts used in proton exchange membrane fuel cells. The base fluid in all these coolants is glycol (1, 3 propanediol), commercially available in a BioGlycol coolant formulation with an ethoxylated nonylphenol surfactant, and azole- and polyol-based non-ionic corrosion inhibitors. Exposure of a Pt/Vulcan carbon electrode to glycol-water or glycol-water-surfactant mixtures causes the loss of Pt electrochemical surface area (ECSA), but the Pt ECSA is fully recovered in clean electrolyte. Only mixtures with the azole corrosion inhibitor cause irreversible losses to the Pt ECSA and oxygen reduction reaction (ORR) activity. The Pt ECSA and ORR activity can only be recovered to within 70% of its initial values after aggressive voltammetric cycling to 1.50 V after azole poisoning. When poisoned with a glycol mixture containing the polyol corrosion inhibitor instead, the Pt ECSA and ORR activity is completely recovered by exposure to a clean electrolyte. The results suggest that prior to incorporation in a fuel cell, voltammetric evaluation of the constituents of coolant formulations is worthwhile.

  16. Desensitization of over tip leakage in an axial turbine rotor by tip surface coolant injection

    NASA Astrophysics Data System (ADS)

    Rao, Nikhil Molahally

    Mechanical energy extraction in axial flow turbine rotors occurs through a change in angular momentum of the working fluid. The gap between the turbine rotor and the stationary casing is referred to as the tip gap. High pressure turbine blades are typically un-shrouded and pressure driven flow through the tip gap is termed as over tip leakage. Over tip leakage reduces efficiency of the turbine stage and also causes thermal distress to blade tip surfaces. The gap height typically increases over the operational life of a turbine, leading to increased efficiency drop. The thermal load on the tip surface also increases with increasing gap height and is exacerbated by the radial transport of high temperature fluid found in the core of the combustor exit flow. Thus over tip leakage not only decreases stage efficiency, but also constrains it by limiting the maximum cycle temperature. Reducing the sensitivity of turbine performance to the effects of the tip gap is termed Tip Desensitization. An experimental investigation of tip desensitization through coolant injection from a tip surface trench was conducted in a large scale, low speed, rotating research turbine facility. Five out of twenty nine rotor blades, referred to as cooled blades, are provided with coolant injection at four locations, at 61%, 71%, 81%, and 91% blade tip axial chord length. At each of the first three locations the coolant jets are directed towards the blade pressure-side, while coolant is exhausted radially at the last location. The sensitivity of total pressure defect, due to over tip leakage, to tip gap height is reduced by both coolant injection and roughening of the casing surface. The total pressure defect due to the large gap height of 1.40% blade height is reduced to levels comparable to the defect due to a gap height of 0.72% blade height. The strong total pressure gradient that characterizes the leakage vortex due to the gap height of 1.40% blade height is considerably diminished by both

  17. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    SciTech Connect

    Hsu, J.S.; Staunton, M.R.; Starke, M.R.

    2006-09-30

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from test data at

  18. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    SciTech Connect

    Staunton, Robert H; Hsu, John S; Starke, Michael R

    2006-09-01

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from test data at

  19. Temporal Statistic of Traffic Accidents in Turkey

    NASA Astrophysics Data System (ADS)

    Erdogan, S.; Yalcin, M.; Yilmaz, M.; Korkmaz Takim, A.

    2015-10-01

    Traffic accidents form clusters in terms of geographic space and over time which themselves exhibit distinct spatial and temporal patterns. There is an imperative need to understand how, where and when traffic accidents occur in order to develop appropriate accident reduction strategies. An improved understanding of the location, time and reasons for traffic accidents makes a significant contribution to preventing them. Traffic accident occurrences have been extensively studied from different spatial and temporal points of view using a variety of methodological approaches. In literature, less research has been dedicated to the temporal patterns of traffic accidents. In this paper, the numbers of traffic accidents are normalized according to the traffic volume and the distribution and fluctuation of these accidents is examined in terms of Islamic time intervals. The daily activities and worship of Muslims are arranged according to these time intervals that are spaced fairly throughout the day according to the position of the sun. The Islamic time intervals are never been used before to identify the critical hour for traffic accidents in the world. The results show that the sunrise is the critical time that acts as a threshold in the rate of traffic accidents throughout Turkey in Islamic time intervals.

  20. Polonium Issue in Fast Reactor Lead Coolants and One of the Ways of Its Solution

    SciTech Connect

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2002-07-01

    One of the main issues in using materials for nuclear facilities is to minimize the production of the most hazardous radionuclides. In the ideal case, all nuclear reactor materials, except a fuel, should be low-activation. The term 'low-activation material' means that this one loses its induced activity in a short time after removal from irradiation. Proposals for building a fusion reactor using low-activation materials are given in Ref.1, 2. For this purpose, low-activation structural materials based on V-Ti-Cr alloys are in the stage of R and D in several countries [3,4]. Another technique to avoid the hazardous activity is in using isotopically enriched materials [5-7]. Although isotopic tailoring option requires tremendous technical efforts and it is too expensive, its application can be first of all assumed for those structural and functional materials which generate very hazardous radionuclides under irradiation. In modern projects of next generation NPPs the preference is given to fast reactors (FRs) with a lead coolant [8]. As it known, the coolant circulating through a FR core is activated, and in the future we should have problems with handling a completed coolant after FR decommissioning or at realization of repair or emergency activities. There, it is desirable to have a low-activation coolant with the low contents of hazardous radionuclides. In papers [9,10] presented at the previous ICONE conferences it was proposed to use lead isotope, Pb-206, as a coolant instead of lead natural, Pb-nat. This paper is devoted to more detailed calculations of accumulating stable bismuth, Bi-209, and polonium radioisotopes, Po-209 (T{sub 1/2}=102 y) and Po-210 (T{sub 1/2}=138 d), in 1 kg of Pb-nat or Pb-206 placed in the core of the BOR-60 type FR. (authors)

  1. Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

    NASA Technical Reports Server (NTRS)

    Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

    2009-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

  2. In-vessel Zircaloy oxidation/hydrogen generation behavior during severe accidents

    SciTech Connect

    Cronenberg, A.W. )

    1990-09-01

    In-vessel Zircaloy oxidation and hydrogen generation data from various US Nuclear Regulatory Commission severe-fuel damage test programs are presented and compared, where the effects of Zircaloy melting, bundle reconfiguration, and bundle quenching by reflooding are assessed for common findings. The experiments evaluated include fuel bundles incorporating fresh and previously irradiated fuel rods, as well as control rods. Findings indicate that the extent of bundle oxidation is largely controlled by steam supply conditions and that high rates of hydrogen generation continued after melt formation and relocation. Likewise, no retardation of hydrogen generation was noted for experiments which incorporated control rods. Metallographic findings indicate extensive oxidation of once-molten Zircaloy bearing test debris. Such test results indicate no apparent limitations to Zircaloy oxidation for fuel bundles subjected to severe-accident coolant-boiloff conditions. 46 refs., 22 figs., 12 tabs.

  3. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    SciTech Connect

    Carbajo, J.J.

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  4. Cooling Characteristics of the V-1650-7 Engine. II - Effect of Coolant Conditions on Cylinder Temperatures and Heat Rejection at Several Engine Powers

    NASA Technical Reports Server (NTRS)

    Povolny, John H.; Bogdan, Louis J.; Chelko, Louis J.

    1947-01-01

    An investigation has been conducted on a V-1650-7 engine to determine the cylinder temperatures and the coolant and oil heat rejections over a range of coolant flows (50 to 200 gal/min) and oil inlet temperatures (160 to 2150 F) for two values of coolant outlet temperature (250 deg and 275 F) at each of four power conditions ranging from approximately 1100 to 2000 brake horsepower. Data were obtained for several values of block-outlet pressure at each of the two coolant outlet temperatures. A mixture of 30 percent by volume of ethylene glycol and 70-percent water was used as the coolant. The effect of varying coolant flow, coolant outlet temperature, and coolant outlet pressure over the ranges investigated on cylinder-head temperatures was small (0 deg to 25 F) whereas the effect of increasing the engine power condition from ll00 to 2000 brake horsepower was large (maximum head-temperature increase, 110 F).

  5. TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

    NASA Technical Reports Server (NTRS)

    Gaugler, R. E.

    1994-01-01

    As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

  6. MSHA releases data on CM crushing accidents

    SciTech Connect

    2007-02-15

    The US Mine Safety and Health Administration (MHSA) recently formed a committee to identify norms and trends in remote control continuous miner crushing accidents. The final report found that these types of accidents commonly happen to experienced miners during routine mining activities, with the majority occurring while moving the miner from one face to another, place changing. Another common aspect of the accidents is that many of the victims are experienced miners who are newly employed at the mine where the accident occurred. Training all employees to stay outside the turning radius of an energized remote control continuous miner, establishing this as a safe operating procedure, and consistently enforcing this practice among miners will reduce these types of accidents. This article was excerpted from the 'Remote Control Continuous Mining Machine Crushing Accident Data Study' published in May 2006. The report may be found from the website: www.msha.gov. 4 figs., 1 tab.

  7. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    SciTech Connect

    Reyes, S; Gomez del Rio, J; Sanz, J

    2000-02-23

    Previous studies of the safety and environmental (S and E) aspects of the HYLIFE-II inertial fusion energy (IFE) power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work a set of computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) has been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here the authors consider a severe lost of coolant accident (LOCA) producing simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the containment) and of the two barriers surrounding the chamber (inner shielding and containment building it self). Even though containment failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product release and transport. The results of these calculations show that the estimated off-site dose is less than 6 mSv (0.6 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  8. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    NASA Astrophysics Data System (ADS)

    Reyes, S.; Latkowski, J. F.; Gomez del Rio, J.; Sanz, J.

    2001-05-01

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  9. Cardiac damage presenting late after road accidents.

    PubMed Central

    Mackintosh, A F; Fleming, H A

    1981-01-01

    Six examples of cardiac damage secondary to non-penetrating trauma in road accidents are described. In all six cases the lesion was not recognised at the time of the accident but became clinically important two days to 17 years later. As the patients were young or had unusual lesions, the damage could be attributed to the accident. In older patients with common cardiac problems the trauma might not be recognised as the underlying cause. PMID:7330802

  10. Transportation accident scenarios for commercial spent fuel

    SciTech Connect

    Wilmot, E L

    1981-02-01

    A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

  11. Calculation notes that support accident scenario and consequence development for the subsurface leak remaining subsurface accident

    SciTech Connect

    Ryan, G.W., Westinghouse Hanford

    1996-07-12

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

  12. Industrial Safety and Accidents Prevention

    SciTech Connect

    Sajjad Akbar

    2006-07-01

    Accident Hazards, dangers, losses and risk are what we would to like to eliminate, minimize or avoid in industry. Modern industries have created many opportunities for these against which man's primitive instincts offer no protection. In today's complex industrial environment safety has become major preoccupation, especially after the realization that there is a clear economic incentive to do so. Industrial hazards may cause by human error or by physical or mechanical malfunction, it is very often possible to eliminate the worst consequences of human error by engineering modification. But the modification also needs checking very thoroughly to ensue that it has not introduced some new and unsuspected hazard. (author)

  13. HTGR severe accident sequence analysis

    SciTech Connect

    Harrington, R.M.; Ball, S.J.; Kornegay, F.C.

    1982-01-01

    Thermal-hydraulic, fission product transport, and atmospheric dispersion calculations are presented for hypothetical severe accident release paths at the Fort St. Vrain (FSV) high temperature gas cooled reactor (HTGR). Off-site radiation exposures are calculated for assumed release of 100% of the 24 hour post-shutdown core xenon and krypton inventory and 5.5% of the iodine inventory. The results show conditions under which dose avoidance measures would be desirable and demonstrate the importance of specific release characteristics such as effective release height. 7 tables.

  14. 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

  15. 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

  16. The determinants of fishing vessel accident severity.

    PubMed

    Jin, Di

    2014-05-01

    The study examines the determinants of fishing vessel accident severity in the Northeastern United States using vessel accident data from the U.S. Coast Guard for 2001-2008. Vessel damage and crew injury severity equations were estimated separately utilizing the ordered probit model. The results suggest that fishing vessel accident severity is significantly affected by several types of accidents. Vessel damage severity is positively associated with loss of stability, sinking, daytime wind speed, vessel age, and distance to shore. Vessel damage severity is negatively associated with vessel size and daytime sea level pressure. Crew injury severity is also positively related to the loss of vessel stability and sinking.

  17. Thermal hydraulic features of the TMI accident

    NASA Astrophysics Data System (ADS)

    Tolman, B.

    1985-10-01

    The Three Mile island (TMI)-2 accident resulted in extensive core damage and recent data confirms that the reactor vessel was challenged from molten core materials. A hypothesized TMI accident scenario is presented that consistently explains the TMI data and is also consistent with research findings from independent severe fuel damage experiments. The TMI data will prove useful in confirming our understanding of severe core damage accidents under realistic reactor systems conditions. This understanding will aid in addressing safety and regulatory issues related to severe core damage accidents in light water reactors.

  18. 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.

  19. [SAFETY IN THE ELDERLY: HOME ACCIDENTS].

    PubMed

    Martín-Espinosa, Noelia M; Píriz-Campos, Rosa Ma; Cordeiro, Raú; Muñoz Bermejo, Laura; Casado Verdjo, Inés; Postigo Mota, Salvador

    2016-05-01

    Home accidents are more common in the elderly and they can have serious consequences to the injured person's health. At home, chances to suffer accidents of any type are higher, because it's the place where old people spend most of their daily time. It is important to point out that a high percentage of domestic accidents could be easily avoided by taking some simple cautions. The main aim of this paper is to know how we can prevent most common domestic accidents in the aged population: falls, burnings, poisonings and fire prevention. PMID:27405149

  20. The determinants of fishing vessel accident severity.

    PubMed

    Jin, Di

    2014-05-01

    The study examines the determinants of fishing vessel accident severity in the Northeastern United States using vessel accident data from the U.S. Coast Guard for 2001-2008. Vessel damage and crew injury severity equations were estimated separately utilizing the ordered probit model. The results suggest that fishing vessel accident severity is significantly affected by several types of accidents. Vessel damage severity is positively associated with loss of stability, sinking, daytime wind speed, vessel age, and distance to shore. Vessel damage severity is negatively associated with vessel size and daytime sea level pressure. Crew injury severity is also positively related to the loss of vessel stability and sinking. PMID:24473412

  1. Study of coolant activation and dose rates with flow rate and power perturbations in pool-type research reactors

    SciTech Connect

    Mirza, N.M.; Mirza, S.M.; Ahmad, N. )

    1991-12-01

    This paper reports on a computer code using the multigroup diffusion theory based LEOPARD and ODMUG programs that has been developed to calculate the activity in the coolant leaving the core of a pool-type research reactor. Using this code, the dose rates at various locations along the coolant path with varying coolant flow rate and reactor power perturbations are determined. A flow rate decrease from 1000 to 145 m{sup 3}/h is considered. The results indicate that a flow rate decrease leads to an increase in the coolant outlet temperature, which affects the neutron group constants and hence the group fluxes. The activity in the coolant leaving the core increases with flow rate decrease. However, at the inlet of the holdup tank, the total dose rate first increases, then passes through a maximum at {approximately} 500 m{sup 3}/h, and finally decreases with flow rate decrease. The activity at the outlet of the holdup tank is mainly due to {sup 24}Na and {sup 56}Mn, and it increases by {approximately} 2% when the flow rate decreases from 1000 to 145 m{sup 3}/h. In an accidental power rise at constant flow rate, the activity in the coolant increases, and the dose rates at all the points along the coolant path show a slight nonlinear rise as the reactor power density increases.

  2. Biorhythmic Cycles and the Incidence of Industrial Accidents

    ERIC Educational Resources Information Center

    Carvey, Davis W.; Nibler, Roger G.

    1977-01-01

    The biorhythm theory of accident explanation that has been increasingly popularized in the business press was empirically examined. Municipal employees involved in work-related vehicular accidents and in on-the-job accidents provided the data. Each accident was analyzed to determine whether or not the accident occurred on a biorhythmically…

  3. 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

  4. Transient analysis for thermal margin with COASISO during a severe accident

    SciTech Connect

    Kim, Chan S.; Chu, Ho S.; Suh, Kune Y.; Park, Goon C.; Lee, Un C.; Yoon, Ho J.

    2002-07-01

    As an IVR-EVC (in-vessel retention through external vessel cooling) design concept, external cooling of the reactor vessel was suggested to protect the lower head from being overheated due to relocated material from the core during a severe accident. The COASISO (Corium Attack Syndrome Immunization Structure Outside the vessel) adopts an external vessel cooling strategy of flooding the reactor vessel inside the thermal insulator. Its advantage is the quick response time so that the initial heat removal mechanism of the EVC is nucleate boiling from the downward-facing lower head. The efficiency of the COASISO may be estimated by the thermal margin defined as the ratio of the actual heat flux from the reactor vessel to the critical heat flux (CHF). In this study the thermal margin for the large power reactor as the APR1400 (Advanced Power Reactor 1400 MWe) was determined by means of transient analysis for the local condition of the coolant and temperature distributions within the reactor vessel. The heat split fraction in the oxide pool and the metal layer focusing effect were considered during calculation of the angular thermal load at the inner wall of the lower head. The temperature distributions in the reactor vessel resulted in the actual heat flux on the outer wall. The local quality was obtained by solving the simplified transient energy equation. The unheated section of the reactor vessel decreases the thermal margin by mean of the two-dimensional conduction heat transfer. The peak temperature of the reactor vessel was estimated in the film boiling region as the thermal margin was equal to unity. Sensitivity analyses were performed for the time of corium relocation after the reactor trip, the coolant flow rate, and the initial subcooled condition of the coolant. There is no vessel failure predicted at the worst EVC condition when the stratification is not taken into account between the metal layer and the oxidic pool. The present predictive tool may be

  5. Compression waves at boiling coolant outflow and peculiarities of their interaction with a barrier

    NASA Astrophysics Data System (ADS)

    Pribaturin, N. A.; Lezhnin, S. I.; Vozhakov, I. S.; Alekseev, M. V.

    2016-10-01

    Numerical simulation of the boiling coolant outflow at a butt-break of high-pressure pipeline was carried out. Interaction between the emerging compression wave and a barrier was studied. The temporal dynamics of the axial pressure profile and pressure in the center of the target were detected. The form of the radial pressure profile on the target was investigated. It has been obtained that in the case of the two-phase outflowing coolant the calculated pressure of the wave reflected from the barrier near the nozzle is less than the theoretical prediction for the ideal gas, and that with an increase in the distance from the nozzle to the barrier the differences between the calculated and theoretical values decrease.

  6. PIV measurements of coolant flow field in a diesel engine cylinder head

    NASA Astrophysics Data System (ADS)

    Ma, Hongwei; Zhang, Zhenyang; Xue, Cheng; Huang, Yunlong

    2015-04-01

    This paper presents experimental measurements of coolant flow field in the water jacket of a diesel engine cylinder head. The test was conducted at three different flow rates using a 2-D PIV system. Appropriate tracing particles were selected and delivery device was designed and manufactured before the test. The flow parameters, such as velocity, vorticity and turbulence, were used to analyze the flow field. The effects of vortex which was located between the intake valve and the exhaust valve were discussed. The experimental results showed an asymmetric distribution of velocity in the water jacket. This led to an asymmetric thermal distribution, which would shorten the service life of the cylinder head. The structure optimization to the water jacket of cylinder head was proposed in this paper. The experimental system, especially the 2-D PIV system, is a great help to study the coolant flow structure and analyze cooling mechanism in the diesel engine cylinder head.

  7. The effect of coolants on the performance of magnetic micro-refrigerators.

    PubMed

    Silva, D J; Bordalo, B D; Pereira, A M; Ventura, J; Oliveira, J C R E; Araújo, J P

    2014-06-01

    Magnetic refrigeration is an alternative cooling technique with envisaged technological applications on micro- and opto-electronic devices. Here, we present a magnetic micro-refrigerator cooling device with embedded micro-channels and based on the magnetocaloric effect. We studied the influence of the coolant fluid in the refrigeration process by numerically simulating the heat transfer processes using the finite element method. This allowed us to calculate the cooling power of the device. Our results show that gallium is the most efficient coolant fluid and, when used with Gd5Si2Ge2, a maximum power of 11.2 W/mm3 at a working frequency of -5 kHz can be reached. However, for operation frequencies around 50 Hz, water is the most efficient fluid with a cooling power of 0.137 W/mm3.

  8. Flow boiling with enhancement devices for cold plate coolant channel design

    NASA Technical Reports Server (NTRS)

    Boyd, Ronald D.; Turknett, Jerry C.; Smith, Alvin

    1989-01-01

    The effects of enhancement devices on flow boiling heat transfer in circular coolant channels, which are heated over a fraction of their perimeters, are studied. The variations were examined in both the mean and local (axial, and circumferential) heat transfer coefficients for a circular coolant channel with either smooth walls or with both a twisted tape and spiral finned walls. Improvements were initiated in the present data reduction analysis. These efforts should lead to the development of heat transfer correlations which include effects of single side heat flux and enhancement device configuration. It is hoped that a stage will be set for the study of heat transfer and pressure drop in single sided heated systems under zero gravity conditions.

  9. Flow boiling with enhancement devices for cold plate coolant channel design

    NASA Technical Reports Server (NTRS)

    Boyd, Ronald D.

    1991-01-01

    Future space exploration and commercialization will require more efficient heat rejection systems. For the required heat transfer rates, such systems must use advanced heat transfer techniques. Forced two phase flow boiling heat transfer with enhancements falls in this category. However, moderate to high quality two phase systems tend to require higher pressure losses. This report is divided into two major parts: (1) Multidimensional wall temperature measurement and heat transfer enhancement for top heated horizontal channels with flow boiling; and (2) Improved analytical heat transfer data reduction for a single side heated coolant channel. Part 1 summarizes over forty experiments which involve both single phase convection and flow boiling in a horizontal channel heated externally from the top side. Part 2 contains parametric dimensionless curves with parameters such as the coolant channel radius ratio, the Biot number, and the circumferential coordinate.

  10. Algebraic grid generation for coolant passages of turbine blades with serpentine channels and pin fins

    NASA Technical Reports Server (NTRS)

    Shih, T. I.-P.; Roelke, R. J.; Steinthorsson, E.

    1991-01-01

    In order to study numerically details of the flow and heat transfer within coolant passages of turbine blades, a method must first be developed to generate grid systems within the very complicated geometries involved. In this study, a grid generation package was developed that is capable of generating the required grid systems. The package developed is based on an algebraic grid generation technique that permits the user considerable control over how grid points are to be distributed in a very explicit way. These controls include orthogonality of grid lines next to boundary surfaces and ability to cluster about arbitrary points, lines, and surfaces. This paper describes that grid generation package and shows how it can be used to generate grid systems within complicated-shaped coolant passages via an example.

  11. Algebraic grid generation for coolant passages of turbine blades with serpentine channels and pin fins

    NASA Astrophysics Data System (ADS)

    Shih, T. I.-P.; Roelke, R. J.; Steinthorsson, E.

    1991-06-01

    In order to study numerically details of the flow and heat transfer within coolant passages of turbine blades, a method must first be developed to generate grid systems within the very complicated geometries involved. In this study, a grid generation package was developed that is capable of generating the required grid systems. The package developed is based on an algebraic grid generation technique that permits the user considerable control over how grid points are to be distributed in a very explicit way. These controls include orthogonality of grid lines next to boundary surfaces and ability to cluster about arbitrary points, lines, and surfaces. This paper describes that grid generation package and shows how it can be used to generate grid systems within complicated-shaped coolant passages via an example.

  12. Direct Numerical Simulation of a Coolant Jet in a Periodic Crossflow

    NASA Technical Reports Server (NTRS)

    Sharma, Chirdeep; Acharya, Sumanta

    1998-01-01

    A Direct Numerical Simulation of a coolant jet injected normally into a periodic crossflow is presented. The physical situation simulated represents a periodic module in a coolant hole array with a heated crossflow. A collocated finite difference scheme is used which is fifth-order accurate spatially and second-order accurate temporally. The scheme is based on a fractional step approach and requires the solution of a pressure-Poisson equation. The simulations are obtained for a blowing ratio of 0.25 and a channel Reynolds number of 5600. The simulations reveal the dynamics of several large scale structures including the Counter-rotating Vortex Pair (CVP), the horse-shoe vortex, the shear layer vortex, the wall vortex and the wake vortex. The origins and the interactions of these vortical structures are identified and explored. Also presented are the turbulence statistics and how they relate to the flow structures.

  13. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    NASA Technical Reports Server (NTRS)

    Siamidis, John; Mason, Lee S.

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed Microsoft Excel (Microsoft Corporation, Redmond, WA) analytical model, HRS_Opt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

  14. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Siamidis, John

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H20 for the HRS pumped loop coolant working fluid. A detailed Microsoft Excel (Microsoft Corporation, Redmond, WA) analytical model, HRS_Opt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

  15. Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

    SciTech Connect

    Soli T. Khericha

    2006-09-01

    This report presents preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T&FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420oC. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation.

  16. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    SciTech Connect

    Roussel, G.

    1997-04-01

    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 snubbers 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.

  17. Influence of coolant pH on corrosion of 6061 aluminum under reactor heat transfer conditions

    SciTech Connect

    Pawel, S.J.; Felde, D.K.; Pawel, R.E.

    1995-10-01

    To support the design of the Advanced Neutron Source (ANS), an experimental program was conducted wherein aluminum alloy specimens were exposed at high heat fluxes to high-velocity aqueous coolants in a corrosion test loop. The aluminum alloys selected for exposure were candidate fuel cladding materials, and the loop system was constructed to emulate the primary coolant system for the proposed ANS reactor. One major result of this program has been the generation of an experimental database defining oxide film growth on 6061 aluminum alloy cladding. Additionally, a data correlation was developed from the database to permit the prediction of film growth for any reasonable thermal-hydraulic excursion. This capability was utilized effectively during the conceptual design stages of the reactor. During the course of this research, it became clear that the kinetics of film growth on the aluminum alloy specimens were sensitively dependent on the chemistry of the aqueous coolant and that relatively small deviations from the intended pH 5 operational level resulted in unexpectedly large changes in the corrosion behavior. Examination of the kinetic influences and the details of the film morphology suggested that a mechanism involving mass transport from other parts of the test loop was involved. Such a mechanism would also be expected to be active in the operating reactor. This report emphasizes the results of experiments that best illustrate the influence of the nonthermal-hydraulic parameters on film growth and presents data to show that comparatively small variations in pH near 5.0 invoke a sensitive response. Simply, for operation in the temperature and heat flux range appropriate for the ANS studies, coolant pH levels from 4.5 to 4.9 produced significantly less film growth than those from pH 5.1 to 6. A mechanism for this behavior based on the concept of treating the entire loop as an active corrosion system is presented.

  18. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    NASA Astrophysics Data System (ADS)

    Siamidis, John; Mason, Lee

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS_Opt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

  19. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    SciTech Connect

    Siamidis, John; Mason, Lee

    2006-01-20

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS{sub O}pt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

  20. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, W.H.; Christiansen, D.W.

    1983-11-25

    This patent discloses a method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  1. Method for controlling coolant flow in airfoil, flow control structure and airfoil incorporating the same

    SciTech Connect

    Itzel, Gary Michael; Devine, II, Robert Henry; Chopra, Sanjay; Toornman, Thomas Nelson

    2003-07-08

    A coolant flow control structure is provided to channel cooling media flow to the fillet region defined at the transition between the wall of a nozzle vane and a wall of a nozzle segment, for cooling the fillet region. In an exemplary embodiment, the flow control structure defines a gap with the fillet region to achieve the required heat transfer coefficients in this region to meet part life requirements.

  2. Main-coolant-pump shaft-seal guidelines. Volume 3. Specification guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria to aid in the generation of procurement specifications for Main Coolant Pump Shaft Seals. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies, a review of pump and shaft seal literature and discussions with pump and seal designers. This report is preliminary in nature and could be expanded and finalized subsequent to completion of further design, test and evaluation efforts.

  3. Main-coolant-pump shaft-seal guidelines. Volume 2. Operational guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria for improving main coolant pump shaft seal operational reliability. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies. Usage procedures/practices and operational environment influence on seal life and reliability from the most recent such survey are summarized. The shaft seal and its auxiliary supporting systems are discussed both from technical and operational related viewpoints.

  4. A Serious Game for Traffic Accident Investigators

    ERIC Educational Resources Information Center

    Binsubaih, Ahmed; Maddock, Steve; Romano, Daniela

    2006-01-01

    In Dubai, traffic accidents kill one person every 37 hours and injure one person every 3 hours. Novice traffic accident investigators in the Dubai police force are expected to "learn by doing" in this intense environment. Currently, they use no alternative to the real world in order to practice. This paper argues for the use of an alternative…

  5. Normal Accident at Three Mile Island.

    ERIC Educational Resources Information Center

    Perrow, Charles

    1981-01-01

    Discusses some aspects of the accident at the Three Mile Island nuclear power plant. Explains a number of factors involved including the type of accident, warnings, design and equipment failure, operator error, and negative synergy. Presents alternatives to systems with catastrophic potential. (MK)

  6. 48 CFR 36.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Accident prevention. 36.513 Section 36.513 Federal Acquisition Regulations System FEDERAL ACQUISITION REGULATION SPECIAL... prevention. (a) The contracting officer shall insert the clause at 52.236-13, Accident Prevention,...

  7. Global estimates of fatal occupational accidents.

    PubMed

    Takala, J

    1999-09-01

    Data on occupational accidents are not available from all countries in the world. Furthermore, underreporting, limited coverage by reporting and compensation schemes, and non-harmonized accident recording and notification systems undermine efforts to obtain worldwide information on occupational accidents. This paper presents a method and new estimated global figures of fatal accidents at work by region. The fatal occupational accident rates reported to the International Labour Office are extended to the total employed workforce in countries and regions. For areas not covered by the reported information, rates from other countries that have similar or comparable conditions are applied. In 1994, an average estimated fatal occupational accident rate in the whole world was 14.0 per 100,000 workers, and the total estimated number of fatal occupational accidents was 335,000. The rates are different for individual countries and regions and for separate branches of economic activity. In conclusion, fatal occupational accident figures are higher than previously estimated. The new estimates can be gradually improved by obtaining and adding data from countries where information is not yet available. Sectoral estimates for at least key economic branches in individual countries would further increase the accuracy.

  8. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration representatives. If a scheduled United States air carrier is involved the airline representatives concerned will... promptly to the nearest office of the airline concerned and to the nearest office of the Civil...

  9. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration representatives. If a scheduled United States air carrier is involved the airline representatives concerned will... promptly to the nearest office of the airline concerned and to the nearest office of the Civil...

  10. Aircraft accidents.method of analysis

    NASA Technical Reports Server (NTRS)

    1937-01-01

    This report is a revision of NACA-TR-357. It was prepared by the Committee on Aircraft Accidents. The purpose of this report is to provide a basis for the classification and comparison of aircraft accidents, both civil and military.

  11. Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

    SciTech Connect

    Finn, P.A.

    1991-03-01

    Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

  12. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    NASA Astrophysics Data System (ADS)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  13. Inverse design of a proper number, shapes, sizes, and locations of coolant flow passages

    NASA Technical Reports Server (NTRS)

    Dulikravich, George S.

    1992-01-01

    During the past several years we have developed an inverse method that allows a thermal cooling system designer to determine proper sizes, shapes, and locations of coolant passages (holes) in, say, an internally cooled turbine blade, a scram jet strut, a rocket chamber wall, etc. Using this method the designer can enforce a desired heat flux distribution on the hot outer surface of the object, while simultaneously enforcing desired temperature distributions on the same hot outer surface as well as on the cooled interior surfaces of each of the coolant passages. This constitutes an over-specified problem which is solved by allowing the number, sizes, locations and shapes of the holes to adjust iteratively until the final internally cooled configuration satisfies the over-specified surface thermal conditions and the governing equation for the steady temperature field. The problem is solved by minimizing an error function expressing the difference between the specified and the computed hot surface heat fluxes. The temperature field analysis was performed using our highly accurate boundary integral element code with linearly varying temperature along straight surface panels. Examples of the inverse design applied to internally cooled turbine blades and scram jet struts (coated and non-coated) having circular and non-circular coolant flow passages will be shown.

  14. Parametric study on maximum transportable distance and cost for thermal energy transportation using various coolants

    SciTech Connect

    Su-Jong Yoon; Piyush Sabharwall

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as district heating, desalination, hydrogen production and other process heat applications, etc. The process heat industry/facilities will be located outside the nuclear island due to safety measures. This thermal energy from the reactor has to be transported a fair distance. In this study, analytical analysis was conducted to identify the maximum distance that thermal energy could be transported using various coolants such as molten-salts, helium and water by varying the pipe diameter and mass flow rate. The cost required to transport each coolant was also analyzed. The coolants analyzed are molten salts (such as: KClMgCl2, LiF-NaF-KF (FLiNaK) and KF-ZrF4), helium and water. Fluoride salts are superior because of better heat transport characteristics but chloride salts are most economical for higher temperature transportation purposes. For lower temperature water is a possible alternative when compared with He, because low pressure He requires higher pumping power which makes the process very inefficient and economically not viable for both low and high temperature application.

  15. Chimera grids in the simulation of three-dimensional flowfields in turbine-blade-coolant passages

    NASA Astrophysics Data System (ADS)

    Stephens, M. A.; Rimlinger, M. J.; Shih, T. I.-P.; Civinskas, K. C.

    1993-06-01

    When computing flows inside geometrically complex turbine-blade coolant passages, the structure of the grid system used can affect significantly the overall time and cost required to obtain solutions. This paper addresses this issue while evaluating and developing computational tools for the design and analysis of coolant-passages, and is divided into two parts. In the first part, the various types of structured and unstructured grids are compared in relation to their ability to provide solutions in a timely and cost-effective manner. This comparison shows that the overlapping structured grids, known as Chimera grids, can rival and in some instances exceed the cost-effectiveness of unstructured grids in terms of both the man hours needed to generate grids and the amount of computer memory and CPU time needed to obtain solutions. In the second part, a computational tool utilizing Chimera grids was used to compute the flow and heat transfer in two different turbine-blade coolant passages that contain baffles and numerous pin fins. These computations showed the versatility and flexibility offered by Chimera grids.

  16. Chimera grids in the simulation of three-dimensional flowfields in turbine-blade-coolant passages

    NASA Technical Reports Server (NTRS)

    Stephens, M. A.; Rimlinger, M. J.; Shih, T. I.-P.; Civinskas, K. C.

    1993-01-01

    When computing flows inside geometrically complex turbine-blade coolant passages, the structure of the grid system used can affect significantly the overall time and cost required to obtain solutions. This paper addresses this issue while evaluating and developing computational tools for the design and analysis of coolant-passages, and is divided into two parts. In the first part, the various types of structured and unstructured grids are compared in relation to their ability to provide solutions in a timely and cost-effective manner. This comparison shows that the overlapping structured grids, known as Chimera grids, can rival and in some instances exceed the cost-effectiveness of unstructured grids in terms of both the man hours needed to generate grids and the amount of computer memory and CPU time needed to obtain solutions. In the second part, a computational tool utilizing Chimera grids was used to compute the flow and heat transfer in two different turbine-blade coolant passages that contain baffles and numerous pin fins. These computations showed the versatility and flexibility offered by Chimera grids.

  17. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    SciTech Connect

    Chopra, O.K.; Shack, W.J.

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented.

  18. Assessment of fiber optic sensors for aging monitoring of industrial liquid coolants

    NASA Astrophysics Data System (ADS)

    Riziotis, Christos; El Sachat, Alexandros; Markos, Christos; Velanas, Pantelis; Meristoudi, Anastasia; Papadopoulos, Aggelos

    2015-03-01

    Lately the demand for in situ and real time monitoring of industrial assets and processes has been dramatically increased. Although numerous sensing techniques have been proposed, only a small fraction can operate efficiently under harsh industrial environments. In this work the operational properties of a proposed photonic based chemical sensing scheme, capable to monitor the ageing process and the quality characteristics of coolants and lubricants in industrial heavy machinery for metal finishing processes is presented. The full spectroscopic characterization of different coolant liquids revealed that the ageing process is connected closely to the acidity/ pH value of coolants, despite the fact that the ageing process is quite complicated, affected by a number of environmental parameters such as the temperature, humidity and development of hazardous biological content as for example fungi. Efficient and low cost optical fiber sensors based on pH sensitive thin overlayers, are proposed and employed for the ageing monitoring. Active sol-gel based materials produced with various pH indicators like cresol red, bromophenol blue and chorophenol red in tetraethylorthosilicate (TEOS), were used for the production of those thin film sensitive layers deposited on polymer's and silica's large core and highly multimoded optical fibers. The optical characteristics, sensing performance and environmental robustness of those optical sensors are presented, extracting useful conclusions towards their use in industrial applications.

  19. Nuclear-radiation-actuated valve. [Patent application; for increasing coolant flow to blanket

    DOEpatents

    Christiansen, D.W.; Schively, D.P.

    1982-01-19

    The present invention relates to a breeder reactor blanket fuel assembly coolant system valve which increases coolant flow to the blanket fuel assembly to minimize long-term temperature increases caused by fission of fissile fuel created from fertile fuel through operation of the breeder reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

  20. The high-temperature sodium coolant technology in nuclear power installations for hydrogen power engineering

    NASA Astrophysics Data System (ADS)

    Kozlov, F. A.; Sorokin, A. P.; Alekseev, V. V.; Konovalov, M. A.

    2014-05-01

    In the case of using high-temperature sodium-cooled nuclear power installations for obtaining hydrogen and for other innovative applications (gasification and fluidization of coal, deep petroleum refining, conversion of biomass into liquid fuel, in the chemical industry, metallurgy, food industry, etc.), the sources of hydrogen that enters from the reactor plant tertiary coolant circuit into its secondary coolant circuit have intensity two or three orders of magnitude higher than that of hydrogen sources at a nuclear power plant (NPP) equipped with a BN-600 reactor. Fundamentally new process solutions are proposed for such conditions. The main prerequisite for implementing them is that the hydrogen concentration in sodium coolant is a factor of 100-1000 higher than it is in modern NPPs taken in combination with removal of hydrogen from sodium by subjecting it to vacuum through membranes made of vanadium or niobium. Numerical investigations carried out using a diffusion model showed that, by varying such parameters as fuel rod cladding material, its thickness, and time of operation in developing the fuel rods for high-temperature nuclear power installations (HT NPIs) it is possible to exclude ingress of cesium into sodium through the sealed fuel rod cladding. However, if the fuel rod cladding loses its tightness, operation of the HT NPI with cesium in the sodium will be unavoidable. Under such conditions, measures must be taken for deeply purifying sodium from cesium in order to minimize the diffusion of cesium into the structural materials.

  1. A {open_quotes}zero waste{close_quotes} coolant management strategy

    SciTech Connect

    Kennicott, M.A.

    1994-04-01

    In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant`s Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant`s liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study`s goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a {open_quotes}zero waste{close_quotes} machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs.

  2. Evaluation of the coolant reactivity coefficient influence on the dynamic response of a small LFR system

    SciTech Connect

    Lorenzi, S.; Bortot, S.; Cammi, A.; Ponciroli, R.

    2012-07-01

    An assessment of the coolant reactivity feedback influence on a small Lead-cooled Fast Reactor (LFR) dynamics has been made aimed at providing both qualitative and quantitative insights into the system transient behavior depending on the sign of the above mentioned coefficient. The need of such an investigation has been recognized since fast reactors cooled by heavy liquid metals show to be characterized by a strong coupling between primary and secondary systems. In particular, the coolant density and radial expansion coefficients have been attested to play a major role in determining the core response to any perturbed condition on the Steam Generator (SG) side. The European Lead-cooled System (ELSY)-based demonstrator (DEMO) has been assumed as the reference LFR case study. As a first step, a zero-dimensional dynamics model has been developed and implemented in MATLAB/SIMULINK{sup R} environment; then typical transient scenarios have been simulated by incorporating the actual negative lead density reactivity coefficient and its opposite. In all the examined cases results have shown that the reactor behaves in a completely different way when considering a positive coolant feedback instead of the reference one, the system free dynamics resulting moreover considerably slower due to the core and SG mutually conflicting reactions. The outcomes of the present analysis may represent a useful feedback for both the core and the control system designers. (authors)

  3. Commercial SNF Accident Release Fractions

    SciTech Connect

    J. Schulz

    2004-11-05

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the

  4. Road accidents caused by drivers falling asleep.

    PubMed

    Sagberg, F

    1999-11-01

    About 29600 Norwegian accident-involved drivers received a questionnaire about the last accident reported to their insurance company. About 9200 drivers (31%) returned the questionnaire. The questionnaire contained questions about sleep or fatigue as contributing factors to the accident. In addition, the drivers reported whether or not they had fallen asleep some time whilst driving. and what the consequences had been. Sleep or drowsiness was a contributing factor in 3.9% of all accidents, as reported by drivers who were at fault for the accident. This factor was strongly over-represented in night-time accidents (18.6%), in running-off-the-road accidents (8.3%), accidents after driving more than 150 km on one trip (8.1%), and personal injury accidents (7.3%). A logistic regression analysis showed that the following additional factors made significant and independent contributions to increasing the odds of sleep involvement in an accident: dry road, high speed limit, driving one's own car, not driving the car daily, high education, and few years of driving experience. More male than female drivers were involved in sleep-related accidents, but this seems largely to be explained by males driving relatively more than females on roads with high speed limits. A total of 10% of male drivers and 4% of females reported to have fallen asleep while driving during the last 12 months. A total of 4% of these events resulted in an accident. The most frequent consequence of falling asleep--amounting to more than 40% of the reported incidents--was crossing of the right edge-line before awaking, whereas crossing of the centreline was reported by 16%. Drivers' lack of awareness of important precursors of falling asleep--like highway hypnosis, driving without awareness, and similar phenomena--as well as a reluctance to discontinue driving despite feeling tired are pointed out as likely contributors to sleep-related accidents. More knowledge about the drivers' experiences immediately

  5. [Poisoning accidents with household chemicals among children].

    PubMed

    Johannsen, H G; Mikkelsen, J B

    1994-10-01

    A review is presented of the registration of all poisoning accidents among children aged 0-6 years treated at the University Hospital, Odense, Denmark during the period 1.1.1980-31.12.1992. There were 1751 poisoning accidents of which 482 were accidents with household chemicals. There were 69 accidents with lamp oil (Petroleum) of which 67 were in the age group 0-3 years. A peak incidence in the age group 0-3 years old is seen in 1986. In 1992 the incidence is at about the same level as in 1980. The incidence in the age group 4-6 years is at almost the same level throughout the entire period. We conclude that it is necessary to continue with campaigns to prevent accidents with household chemicals among children.

  6. [Clinical examinations for the traffic accident patients].

    PubMed

    Hitosugi, Masahito

    2008-11-30

    Traffic accident is a leading cause of unintentional death and about six-thousands annually died in Japan. As about one-million of persons suffer from traffic injuries, most of them seek medical attention. Therefore, medical staffs have to find the injuries accurately and treat immediately. Furthermore, the cause of accident should also be considered; why the accident was occurred, human error of the driver? To solve these problems, clinical examinations were needed. Medical staffs have to understand the characteristics of the traffic injuries: severe and multiple blunt injuries, popular injuries can be estimated with considering the pattern of the accident. Because some of the accidents are occurred when the driver is under the influence of alcohol and other drugs, screening of these subjects should be performed. Because the public is largely unaware of the preventable nature of traffic injuries, in addition to diagnose and treat accurately, we medical staffs have to attend on the primary prevention of the traffic injuries.

  7. The Fukushima Daiichi Accident Study Information Portal

    SciTech Connect

    Shawn St. Germain; Curtis Smith; David Schwieder; Cherie Phelan

    2012-11-01

    This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.

  8. Accident prevention: the health visitor's role.

    PubMed

    Levene, S

    1992-10-01

    The health of the nation white paper sets targets in five key areas for reductions in both mortality and morbidity: coronary heart disease and stroke, cancers, mental illness, HIV/Aids and sexual health and accidents. In a series of articles in Health visitor, experts will be considering the opportunities the white paper offers for community nurses in each of the key areas. Here Dr Sara Levene, medical consultant to the Child Accident Prevention Trust, considers accidents, a major problem which health visitors can do much to control. She reviews how accidents are presented in the white paper, what health visitors can do and what resources are available to help them. She offers particular advice on special accident prevention initiatives and discusses some of the opportunities created by the white paper.

  9. Pilot-error accidents: male vs female.

    PubMed

    Vail, G J; Ekman, L G

    1986-12-01

    In this study, general aviation accident records from the files of the National Transportation Safety Board (NTSB), have been analysed by gender to observe the number and rate of pilot-error related accidents from 1972 to 1981 inclusive. If both females and males have no difference in performance, then data would have indicated similarities of accident rates and types of injuries. Males had a higher rate of accidents than females, and a higher portion of the male accidents resulted in fatalities or serious injuries than for females. Type of certificate, age, total flight time, flight time in type of aircraft, phase of operation, category of flying, degree of injury, specific cause factors, cause factor miscellaneous acts/conditions were analysed, taking the total number of United States Active Civilian General Aviation Pilots into consideration. The data did indicate a difference in all variables.

  10. Human Factors in Cabin Accident Investigations

    NASA Technical Reports Server (NTRS)

    Chute, Rebecca D.; Rosekind, Mark R. (Technical Monitor)

    1996-01-01

    Human factors has become an integral part of the accident investigation protocol. However, much of the investigative process remains focussed on the flight deck, airframe, and power plant systems. As a consequence, little data has been collected regarding the human factors issues within and involving the cabin during an accident. Therefore, the possibility exists that contributing factors that lie within that domain may be overlooked. The FAA Office of Accident Investigation is sponsoring a two-day workshop on cabin safety accident investigation. This course, within the workshop, will be of two hours duration and will explore relevant areas of human factors research. Specifically, the three areas of discussion are: Information transfer and resource management, fatigue and other physical stressors, and the human/machine interface. Integration of these areas will be accomplished by providing a suggested checklist of specific cabin-related human factors questions for investigators to probe following an accident.

  11. Numerical analysis of the hot-gas-side and coolant-side heat transfer in liquid rocket engine combustors

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Van, Luong

    1992-01-01

    The objective of this paper are to develop a multidisciplinary computational methodology to predict the hot-gas-side and coolant-side heat transfer and to use it in parametric studies to recommend optimized design of the coolant channels for a regeneratively cooled liquid rocket engine combustor. An integrated numerical model which incorporates CFD for the hot-gas thermal environment, and thermal analysis for the liner and coolant channels, was developed. This integrated CFD/thermal model was validated by comparing predicted heat fluxes with those of hot-firing test and industrial design methods for a 40 k calorimeter thrust chamber and the Space Shuttle Main Engine Main Combustion Chamber. Parametric studies were performed for the Advanced Main Combustion Chamber to find a strategy for a proposed combustion chamber coolant channel design.

  12. [Cerebral vascular accidents in French Polynesia].

    PubMed

    Gras, C; Papouin, G; Prigent, D; Beaugendre, E; Lionet, P; Brodin, S; Legall, R; Marjou, F; Spiegel, A; Gendron, Y

    1992-01-01

    The authors report on the results of a survey on cardiovascular accidents hospitalized between 01 April 1990 and 31 January 1991 carried out in the Services of Medicine and Cardiology in the Territorial Hospital Center of Papeete. This survey was: 56 cardiovascular accidents: 1/4 (hemorrhagic and 3/4 (42) ischemic. Mean age 59 (extremes 23-86). 36 males (64%); 20 females (36%). 50 Polynesians; 6 Chinese people. Among the risk factors recorded, 38 (68%) were hypertensed patients; 17 (30%) were due to tabagism and 15 (25%) to diabetes; 3 (5%) are known to be carriers of a hypercholesterolemia. 59% of the patients had no case history; 25% the cardiovascular accidents have been observed in patients with cardiopathy; 12.5% are recurrent cardiovascular accidents. Clinically, 5 transient ischemic accidents (12%) out of 42 cardiovascular ischemic accidents. High arterial tension was recognized in 12/14 (86%) of hemorrhagic cardiovascular accidents and in 26/42 (62%) of ischemic cardiovascular accidents. In 42 ischemic cardiovascular accidents, 31 patients suffered from cardiopathy (74%) of which 15 (36%) presented an embolic cardiopathy. Interest of echography and electrocardiogram are discussed. Ultrasonic exam of carotid vessels was found abnormal in almost half of the cases when utilized (12/26). Finally, etiological diagnosis was certain in 17 cases, of presumption in 16 cases, and in 9 cases, it was not possible to precise any cardiovascular etiology. Tomodensitometric tests are discussed. 86% of the ischemic cardiovascular accident were treated with anticoagulants/thrombocyte antiagglutination. 24% of the patients died, 50% recovered incompletely and 26% completely. PMID:1602953

  13. The Fukushima accident was preventable.

    PubMed

    Synolakis, Costas; Kânoğlu, Utku

    2015-10-28

    The 11 March 2011 tsunami was probably the fourth largest in the past 100 years and killed over 15 000 people. The magnitude of the design tsunami triggering earthquake affecting this region of Japan had been grossly underestimated, and the tsunami hit the Fukushima Dai-ichi nuclear power plant (NPP), causing the third most severe accident in an NPP ever. Interestingly, while the Onagawa NPP was also hit by a tsunami of approximately the same height as Dai-ichi, it survived the event 'remarkably undamaged'. We explain what has been referred to as the cascade of engineering and regulatory failures that led to the Fukushima disaster. One, insufficient attention had been given to evidence of large tsunamis inundating the region earlier, to Japanese research suggestive that large earthquakes could occur anywhere along a subduction zone, and to new research on mega-thrusts since Boxing Day 2004. Two, there were unexplainably different design conditions for NPPs at close distances from each other. Three, the hazard analysis to calculate the maximum probable tsunami at Dai-ichi appeared to have had methodological mistakes, which almost nobody experienced in tsunami engineering would have made. Four, there were substantial inadequacies in the Japan nuclear regulatory structure. The Fukushima accident was preventable, if international best practices and standards had been followed, if there had been international reviews, and had common sense prevailed in the interpretation of pre-existing geological and hydrodynamic findings. Formal standards are needed for evaluating the tsunami vulnerability of NPPs, for specific training of engineers and scientists who perform tsunami computations for emergency preparedness or critical facilities, as well as for regulators who review safety studies.

  14. The Fukushima accident was preventable.

    PubMed

    Synolakis, Costas; Kânoğlu, Utku

    2015-10-28

    The 11 March 2011 tsunami was probably the fourth largest in the past 100 years and killed over 15 000 people. The magnitude of the design tsunami triggering earthquake affecting this region of Japan had been grossly underestimated, and the tsunami hit the Fukushima Dai-ichi nuclear power plant (NPP), causing the third most severe accident in an NPP ever. Interestingly, while the Onagawa NPP was also hit by a tsunami of approximately the same height as Dai-ichi, it survived the event 'remarkably undamaged'. We explain what has been referred to as the cascade of engineering and regulatory failures that led to the Fukushima disaster. One, insufficient attention had been given to evidence of large tsunamis inundating the region earlier, to Japanese research suggestive that large earthquakes could occur anywhere along a subduction zone, and to new research on mega-thrusts since Boxing Day 2004. Two, there were unexplainably different design conditions for NPPs at close distances from each other. Three, the hazard analysis to calculate the maximum probable tsunami at Dai-ichi appeared to have had methodological mistakes, which almost nobody experienced in tsunami engineering would have made. Four, there were substantial inadequacies in the Japan nuclear regulatory structure. The Fukushima accident was preventable, if international best practices and standards had been followed, if there had been international reviews, and had common sense prevailed in the interpretation of pre-existing geological and hydrodynamic findings. Formal standards are needed for evaluating the tsunami vulnerability of NPPs, for specific training of engineers and scientists who perform tsunami computations for emergency preparedness or critical facilities, as well as for regulators who review safety studies. PMID:26392611

  15. [A study on fall accident].

    PubMed

    Lee, H S; Kim, M J

    1997-01-01

    The study was conducted from November 1995 to May 1996 at the one general hospital in Seoul. The total subjects of this study were 412 patients who have the experience of fall accident, among them 31 was who have fallen during hospitalization and 381 was who visited emergency room and out patient clinic. The purposes of this study were to determine the characteristics, risk factors and results of fall accident and to suggest the nursing strategies for prevention of fall. Data were collected by reviewing the medical records and interviewing with the fallers and their family members. For data analysis spss/pc+ program was utilized for descriptive statistics, adjusted standardized X2-test. The results of this study were as follows: 1) Total subjects were 412 fallers, of which 245 (59.5%) were men and 167 (40.5%) were women. Age were 0-14 years 79 (19.2%), 15-44 years 125 (30.4%), 45-64 years 104 (25.2%), over 65 years 104 (25.2%). 2) There was significant association between age and the sexes (X2 = 39.17, P = 0.00). 3) There was significant association between age and history of falls (X2 = 44.41, P = .00). And history of falls in the elderly was significantly associated with falls. 4) There was significant association with age and medical diagnosis (X2 = 140.66, P = .00), chief medical diagnosis were hypertension (34), diabetes mellitus (22), arthritis (11), stroke (8), fracture (7), pulmonary tuberculosis (6), dementia (5) and cataract (5). 5) There was significant association between age and intrinsic factors: cognitive impairment, mobility impairment, insomnia, emotional problems, urinary difficulty, visual impairments, hearing impairments, use of drugs (sedatives, antihypertensive drugs, diuretics, antidepressants) (P < 0.05). But there was no significant association between age and dizziness (X2 = 2.87, P = .41). 6) 15.3% of total fallers were drunken state when they were fallen. 7) Environmental factors of fall accident were unusual posture (50.9%), slips (35

  16. Analyses of fluid-structure interaction and structural response of reactor vessels to a postulated accident

    SciTech Connect

    Wang, C.Y.

    1993-08-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the improved hybrid Lagrangian-Eulerian code ALICE-II. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water outside the reactor vessel, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

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

    SciTech Connect

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

    1991-01-01

    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 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. 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.

  18. Early results from an experimental program to determine the behavior of containment piping penetration bellows subjected to severe accident conditions

    SciTech Connect

    Lambert, L.D.; Parks, M.B.

    1994-09-01

    Containment piping penetration bellows are an integral part of the pressure boundary in steel containments in the United States (US). Their purpose is to minimize loading on the containment shell caused by differential movement between the piping and the containment. This differential movement is typically caused by thermal gradients generated during startup and shutdown of the reactor, but can be caused by earthquake, a loss-of-coolant accident (LOCA), or ``severe`` accidents. In the event of a severe accident, the bellows would be subjected to pressure, temperature, and deflection well beyond the design basis. Most bellows are installed such that they would be subjected to elevated internal pressure, elevated temperature, axial compression, and lateral deflection during a severe accident. A few bellows would be subjected to external pressure and axial elongation, as well as elevated temperature and lateral deflection. The purpose of this experimental program is to examine the potential for leakage of containment bellows during a severe accident. The test series subjects bellows to various levels and combinations of internal pressure, elevated temperature, axial compression or elongation, and lateral deformation. The experiments are being conducted in two parts. For Part 1, all bellows specimens are tested in ``like-new`` condition, without regard for the possible degrading effect of corrosion that has been observed in some containment piping bellows in the US Part I testing, which included 13 bellows tests, has been completed. The second part of the experimental program, in which bellows are subjected to simulated corrosive environments prior to testing, has just just begun. The Part I experiments have shown that bellows in ``like-new`` condition can withstand elevated temperatures and pressures along with large deformations before leaking. In most cases, the like-new bellows were fully compressed without developing any leakage.

  19. 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.

  20. An analysis of aircraft accidents involving fires

    NASA Technical Reports Server (NTRS)

    Lucha, G. V.; Robertson, M. A.; Schooley, F. A.

    1975-01-01

    All U. S. Air Carrier accidents between 1963 and 1974 were studied to assess the extent of total personnel and aircraft damage which occurred in accidents and in accidents involving fire. Published accident reports and NTSB investigators' factual backup files were the primary sources of data. Although it was frequently not possible to assess the relative extent of fire-caused damage versus impact damage using the available data, the study established upper and lower bounds for deaths and damage due specifically to fire. In 12 years there were 122 accidents which involved airframe fires. Eighty-seven percent of the fires occurred after impact, and fuel leakage from ruptured tanks or severed lines was the most frequently cited cause. A cost analysis was performed for 300 serious accidents, including 92 serious accidents which involved fire. Personal injury costs were outside the scope of the cost analysis, but data on personnel injury judgements as well as settlements received from the CAB are included for reference.

  1. Accidents in Canada: mortality and hospitalization.

    PubMed

    Riley, R; Paddon, P

    1989-01-01

    For Canadians under 45, accidents are the leading cause of both death and hospitalization. For the Canadian population as a whole, accidents rank fourth as a cause of death, after cardiovascular disease (CVD), cancer and respiratory disease. This article analyzes accident mortality and hospitalization in Canada using age-specific rates, age-standardized mortality rates (ASMR), and potential years of life lost (PYLL). The six major causes of accidental death for men are motor vehicle traffic accidents (MVTA), falls, drowning, fires, suffocation and poisoning. For women, the order is slightly different: MVTA, falls, fires, suffocation, poisoning and drowning. From 1971 to 1986, age-standardized mortality rates (ASMR) for accidents decreased by 44% for men and 39% for women. The largest decrease occurred in the under 15 age group. Accidents accounted for 11.5% of total hospital days in 1985, and 8% of hospital discharges. Because young people have the highest rates of accidental death, potential years of life lost (PYLL) are almost as high for accidents as for cardiovascular disease, although CVD deaths outnumbered accidental deaths by almost five to one in 1985. PMID:2491351

  2. Road accidents and business cycles in Spain.

    PubMed

    Rodríguez-López, Jesús; Marrero, Gustavo A; González, Rosa Marina; Leal-Linares, Teresa

    2016-11-01

    This paper explores the causes behind the downturn in road accidents in Spain across the last decade. Possible causes are grouped into three categories: Institutional factors (a Penalty Point System, PPS, dating from 2006), technological factors (active safety and passive safety of vehicles), and macroeconomic factors (the Great recession starting in 2008, and an increase in fuel prices during the spring of 2008). The PPS has been blessed by incumbent authorities as responsible for the decline of road fatalities in Spain. Using cointegration techniques, the GDP growth rate, the fuel price, the PPS, and technological items embedded in motor vehicles appear to be statistically significantly related with accidents. Importantly, PPS is found to be significant in reducing fatal accidents. However, PPS is not significant for non-fatal accidents. In view of these results, we conclude that road accidents in Spain are very sensitive to the business cycle, and that the PPS influenced the severity (fatality) rather than the quantity of accidents in Spain. Importantly, technological items help explain a sizable fraction in accidents downturn, their effects dating back from the end of the nineties.

  3. Childhood accidents: epidemiology, trends, and prevention.

    PubMed Central

    Kemp, A; Sibert, J

    1997-01-01

    Accidents are the most common cause of death in children over one year of age. Prevention remains a high priority. We have reviewed the current epidemiology of childhood accidents and their prevention, and made recommendations for the future. In 1992, 559 children died in United Kingdom as a result of an accidents--240 from road traffic accidents and 100 from burns and scalds. Every year 50 children drown. Accidents cause significant disability to children. Many children, up to one in four of the population in urban areas, attend accident and emergency departments, and 5-10% of these are admitted to hospital. Accident risk factors include low social class, psychosocial stress, an unsafe environment, and child developmental disorders. Research has shown that prevention is best achieved by making the child's environment safer, often through legislation. Insufficient resources have been put into both research into childhood injuries and preventive work in communities. Collaboration between health authorities, NHS trusts, local authorities and community networks is vital if success is to be achieved. A national safety agenda for children would focus the attention that this problem deserves. PMID:9315935

  4. Accidents in Canada: mortality and hospitalization.

    PubMed

    Riley, R; Paddon, P

    1989-01-01

    For Canadians under 45, accidents are the leading cause of both death and hospitalization. For the Canadian population as a whole, accidents rank fourth as a cause of death, after cardiovascular disease (CVD), cancer and respiratory disease. This article analyzes accident mortality and hospitalization in Canada using age-specific rates, age-standardized mortality rates (ASMR), and potential years of life lost (PYLL). The six major causes of accidental death for men are motor vehicle traffic accidents (MVTA), falls, drowning, fires, suffocation and poisoning. For women, the order is slightly different: MVTA, falls, fires, suffocation, poisoning and drowning. From 1971 to 1986, age-standardized mortality rates (ASMR) for accidents decreased by 44% for men and 39% for women. The largest decrease occurred in the under 15 age group. Accidents accounted for 11.5% of total hospital days in 1985, and 8% of hospital discharges. Because young people have the highest rates of accidental death, potential years of life lost (PYLL) are almost as high for accidents as for cardiovascular disease, although CVD deaths outnumbered accidental deaths by almost five to one in 1985.

  5. Bayes classifiers for imbalanced traffic accidents datasets.

    PubMed

    Mujalli, Randa Oqab; López, Griselda; Garach, Laura

    2016-03-01

    Traffic accidents data sets are usually imbalanced, where the number of instances classified under the killed or severe injuries class (minority) is much lower than those classified under the slight injuries class (majority). This, however, supposes a challenging problem for classification algorithms and may cause obtaining a model that well cover the slight injuries instances whereas the killed or severe injuries instances are misclassified frequently. Based on traffic accidents data collected on urban and suburban roads in Jordan for three years (2009-2011); three different data balancing techniques were used: under-sampling which removes some instances of the majority class, oversampling which creates new instances of the minority class and a mix technique that combines both. In addition, different Bayes classifiers were compared for the different imbalanced and balanced data sets: Averaged One-Dependence Estimators, Weightily Average One-Dependence Estimators, and Bayesian networks in order to identify factors that affect the severity of an accident. The results indicated that using the balanced data sets, especially those created using oversampling techniques, with Bayesian networks improved classifying a traffic accident according to its severity and reduced the misclassification of killed and severe injuries instances. On the other hand, the following variables were found to contribute to the occurrence of a killed causality or a severe injury in a traffic accident: number of vehicles involved, accident pattern, number of directions, accident type, lighting, surface condition, and speed limit. This work, to the knowledge of the authors, is the first that aims at analyzing historical data records for traffic accidents occurring in Jordan and the first to apply balancing techniques to analyze injury severity of traffic accidents.

  6. Scoping assessments of ATF impact on late-stage accident progression including molten core-concrete interaction

    NASA Astrophysics Data System (ADS)

    Farmer, M. T.; Leibowitz, L.; Terrani, K. A.; Robb, K. R.

    2014-05-01

    Simple scoping models that can be used to evaluate ATF performance under severe accident conditions have been developed. The methodology provides a fundamental technical basis (a.k.a. metric) based on the thermodynamic boundary for evaluating performance relative to that of traditional Zr-based claddings. The initial focus in this study was on UO2 fuel with the advanced claddings 310 SS, D9, FeCrAl, and SiC. The evaluation considered only energy release with concurrent combustible gas production from fuel-cladding-coolant interactions and, separately, molten core-concrete interactions at high temperatures. Other important phenomenological effects that can influence the rate and extent of cladding decomposition (e.g., eutectic interactions, degradation of other core constituents) were not addressed. For the cladding types addressed, potential combustible gas production under both in-vessel and ex-vessel conditions was similar to that for Zr. However, exothermic energy release from cladding oxidation was substantially less for iron-based alloys (by at least a factor of 4), and modestly less (by ∼20%) for SiC. Data on SiC-clad UO2 fuel performance under severe accident conditions are sparse in the literature; thus, assumptions on the nature of the cladding decomposition process were made in order to perform this initial screening evaluation. Experimental data for this system under severe accident conditions is needed for a proper evaluation and comparison to iron-based claddings.

  7. Computational Assessment of the GT-MHR Graphite Core Support Structural Integrity in Air-Ingress Accident Condition

    SciTech Connect

    Jong B. Lim; Eung S. Kim; Chang H. Oh; Richard R. Schultz; David A. Petti

    2008-10-01

    The objective of this project was to perform stress analysis for graphite support structures of the General Atomics’ 600 MWth GT-MHR prismatic core design using ABAQUS ® (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR was analyzed based on the change of temperature, burn-off and corrosion depth during the accident period predicted by GAMMA, a multi-dimensional gas multi-component mixture analysis code developed in the Republic of Korea (ROK)/United States (US) International –Nuclear Engineering Research Initiative (I-NERI) project. Both the loading and thermal stresses were analyzed, but the thermal stress was not significant, leaving the loading stress to be the major factor. The mechanical strengths are exceeded between 11 to 11.5 days after loss-of-coolant-accident (LOCA), corresponding to 5.5 to 6 days after the start of natural convection.

  8. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis: User's guide

    SciTech Connect

    Rettig, W.H.; Wade, N.L. )

    1992-06-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODI version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MOD1 version produces results consistent with previous versions. Assessment calculations using the two TRAC-BFI versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  9. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis, Model description

    SciTech Connect

    Borkowski, J.A.; Wade, N.L.; Giles, M.M.; Rouhani, S.Z.; Shumway, R.W.; Singer, G.L.; Taylor, D.D.; Weaver, W.L. )

    1992-08-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODl version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MODI version produces results consistent with previous versions. Assessment calculations using the two TRAC-BF1 versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  10. Chemical considerations in severe accident analysis

    SciTech Connect

    Malinauskas, A.P.; Kress, T.S.

    1988-01-01

    The Reactor Safety Study presented the first systematic attempt to include fission product physicochemical effects in the determination of expected consequences of hypothetical nuclear reactor power plant accidents. At the time, however, the data base was sparse, and the treatment of fission product behavior was not entirely consistent or accurate. Considerable research has since been performed to identify and understand chemical phenomena that can occur in the course of a nuclear reactor accident, and how these phenomena affect fission product behavior. In this report, the current status of our understanding of the chemistry of fission products in severe core damage accidents is summarized and contrasted with that of the Reactor Safety Study.

  11. Impact of nuclear accidents on marine biota.

    PubMed

    Vives i Batlle, Jordi

    2011-07-01

    The accident at the Fukushima Daiichi nuclear power plant, precipitated by the earthquake and subsequent tsunami that struck the northeastern coast of Japan in March 2011, has raised concerns about the potential impact to marine biota posed by the release of radioactive water and radionuclide particles into the environment. The Fukushima accident is the only major nuclear accident that has resulted in the direct discharge of radioactive materials into a coastal environment. This article briefly summarizes what is currently understood about the effects of radioactive wastewaters and radionuclides to marine life.

  12. An Application of CICCT Accident Categories to Aviation Accidents in 1988-2004

    NASA Technical Reports Server (NTRS)

    Evans, Joni K.

    2007-01-01

    Interventions or technologies developed to improve aviation safety often focus on specific causes or accident categories. Evaluation of the potential effectiveness of those interventions is dependent upon mapping the historical aviation accidents into those same accident categories. To that end, the United States civil aviation accidents occurring between 1988 and 2004 (n=26,117) were assigned accident categories based upon the taxonomy developed by the CAST/ICAO Common Taxonomy Team (CICTT). Results are presented separately for four main categories of flight rules: Part 121 (large commercial air carriers), Scheduled Part 135 (commuter airlines), Non-Scheduled Part 135 (on-demand air taxi) and Part 91 (general aviation). Injuries and aircraft damage are summarized by year and by accident category.

  13. Review of models applicable to accident aerosols

    SciTech Connect

    Glissmeyer, J.A.

    1983-07-01

    Estimations of potential airborne-particle releases are essential in safety assessments of nuclear-fuel facilities. This report is a review of aerosol behavior models that have potential applications for predicting aerosol characteristics in compartments containing accident-generated aerosol sources. Such characterization of the accident-generated aerosols is a necessary step toward estimating their eventual release in any accident scenario. Existing aerosol models can predict the size distribution, concentration, and composition of aerosols as they are acted on by ventilation, diffusion, gravity, coagulation, and other phenomena. Models developed in the fields of fluid mechanics, indoor air pollution, and nuclear-reactor accidents are reviewed with this nuclear fuel facility application in mind. The various capabilities of modeling aerosol behavior are tabulated and discussed, and recommendations are made for applying the models to problems of differing complexity.

  14. Safety analysis of surface haulage accidents

    SciTech Connect

    Randolph, R.F.; Boldt, C.M.K.

    1996-12-31

    Research on improving haulage truck safety, started by the U.S. Bureau of Mines, is being continued by its successors. This paper reports the orientation of the renewed research efforts, beginning with an update on accident data analysis, the role of multiple causes in these accidents, and the search for practical methods for addressing the most important causes. Fatal haulage accidents most often involve loss of control or collisions caused by a variety of factors. Lost-time injuries most often involve sprains or strains to the back or multiple body areas, which can often be attributed to rough roads and the shocks of loading and unloading. Research to reduce these accidents includes improved warning systems, shock isolation for drivers, encouraging seatbelt usage, and general improvements to system and task design.

  15. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

    Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

    1999-11-01

    The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

  16. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

    Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

    1999-01-01

    The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

  17. 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.

  18. Numerical Simulation of Non-Rotating and Rotating Coolant Channel Flow Fields. Part 1

    NASA Technical Reports Server (NTRS)

    Rigby, David L.

    2000-01-01

    Future generations of ultra high bypass-ratio jet engines will require far higher pressure ratios and operating temperatures than those of current engines. For the foreseeable future, engine materials will not be able to withstand the high temperatures without some form of cooling. In particular the turbine blades, which are under high thermal as well as mechanical loads, must be cooled. Cooling of turbine blades is achieved by bleeding air from the compressor stage of the engine through complicated internal passages in the turbine blades (internal cooling, including jet-impingement cooling) and by bleeding small amounts of air into the boundary layer of the external flow through small discrete holes on the surface of the blade (film cooling and transpiration cooling). The cooling must be done using a minimum amount of air or any increases in efficiency gained through higher operating temperature will be lost due to added load on the compressor stage. Turbine cooling schemes have traditionally been based on extensive empirical data bases, quasi-one-dimensional computational fluid dynamics (CFD) analysis, and trial and error. With improved capabilities of CFD, these traditional methods can be augmented by full three-dimensional simulations of the coolant flow to predict in detail the heat transfer and metal temperatures. Several aspects of turbine coolant flows make such application of CFD difficult, thus a highly effective CFD methodology must be used. First, high resolution of the flow field is required to attain the needed accuracy for heat transfer predictions, making highly efficient flow solvers essential for such computations. Second, the geometries of the flow passages are complicated but must be modeled accurately in order to capture all important details of the flow. This makes grid generation and grid quality important issues. Finally, since coolant flows are turbulent and separated the effects of turbulence must be modeled with a low Reynolds number

  19. Film-cooling effectiveness with developing coolant flow through straight and curved tubular passages

    NASA Technical Reports Server (NTRS)

    Papell, S. S.; Wang, C. R.; Graham, R. W.

    1982-01-01

    The data were obtained with an apparatus designed to determine the influence of tubular coolant passage curvature on film-cooling performance while simulating the developing flow entrance conditions more representative of cooled turbine blade. Data comparisons were made between straight and curved single tubular passages embedded in the wall and discharging at 30 deg angle in line with the tunnel flow. The results showed an influence of curvature on film-cooling effectiveness that was inversely proportional to the blowing rate. At the lowest blowing rate of 0.18, curvature increased the effectiveness of film cooling by 35 percent; but at a blowing rate of 0.76, the improvement was only 10 percent. In addition, the increase in film-cooling area coverage ranged from 100 percent down to 25 percent over the same blowing rates. A data trend reversal at a blowing rate of 1.5 showed the straight tubular passage's film-cooling effectiveness to be 20 percent greater than that of the curved passage with about 80 percent more area coverage. An analysis of turbulence intensity detain the mixing layer in terms of the position of the mixing interface relative to the wall supported the concept that passage curvature tends to reduce the diffusion of the coolant jet into the main stream at blowing rates below about. Explanations for the film-cooling performance of both test sections were made in terms differences in turbulences structure and in secondary flow patterns within the coolant jets as influenced by flow passage geometry.

  20. 50 CFR 25.72 - Reporting of accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ..., but in no event later than 24 hours after the accident, by the persons involved, to the refuge manager... 50 Wildlife and Fisheries 8 2011-10-01 2011-10-01 false Reporting of accidents. 25.72 Section 25... Reporting of accidents. Accidents involving damage to property, injury to the public or injury to...

  1. 50 CFR 25.72 - Reporting of accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ..., but in no event later than 24 hours after the accident, by the persons involved, to the refuge manager... 50 Wildlife and Fisheries 6 2010-10-01 2010-10-01 false Reporting of accidents. 25.72 Section 25... Reporting of accidents. Accidents involving damage to property, injury to the public or injury to...

  2. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 16 2012-07-01 2012-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases...

  3. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 16 2014-07-01 2014-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases...

  4. Oranges and Peaches: Understanding Communication Accidents in the Reference Interview.

    ERIC Educational Resources Information Center

    Dewdney, Patricia; Michell, Gillian

    1996-01-01

    Librarians often have communication "accidents" with reference questions as initially presented. This article presents linguistic analysis of query categories, including: simple failures of hearing, accidents involving pronunciation or homophones, accidents where users repeat earlier misinterpretations to librarians, and accidents where users…

  5. 48 CFR 852.236-87 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Accident prevention. 852... Accident prevention. As prescribed in 836.513, insert the following clause: Accident Prevention (SEP 1993....236-13, Accident Prevention. However, only the Contracting Officer may issue an order to stop all...

  6. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases...

  7. 46 CFR 97.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Accidents to machinery. 97.30-5 Section 97.30-5 Shipping... Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use...

  8. 46 CFR 196.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Accidents to machinery. 196.30-5 Section 196.30-5... Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use...

  9. 46 CFR 78.33-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 3 2010-10-01 2010-10-01 false Accidents to machinery. 78.33-5 Section 78.33-5 Shipping... Accidents, Repairs, and Unsafe Equipment § 78.33-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use of the...

  10. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 4 2012-01-01 2012-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  11. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 4 2010-01-01 2010-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  12. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 4 2013-01-01 2013-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  13. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 4 2014-01-01 2014-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  14. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 4 2011-01-01 2011-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  15. Learning lessons from Natech accidents - the eNATECH accident database

    NASA Astrophysics Data System (ADS)

    Krausmann, Elisabeth; Girgin, Serkan

    2016-04-01

    When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of

  16. The development of Sn-Li coolant/breeding material for APEX/ALPS applications.

    SciTech Connect

    Sze, D.-K.

    1999-07-08

    A Sn-Li alloy has been identified to be a coolant/breeding material for D-T fusion applications. The key feature of this material is its very low vapor pressure, which will be very useful for free surface concepts employed in APEX, ALPS and inertial confinement fission. The vapor is dominated by lithium, which has very low Z. Initial assessment of the material indicates acceptable tritium breeding capability, high thermal conductivity, expected low tritium volubility, and expected low chemical reactivities with water and air. Some key concerns are the high activation and material compatibility issues. The initial assessment of this material, for fission applications, is presented in this paper.

  17. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    SciTech Connect

    Zapata, Luis E.; Hackel, Lloyd

    1999-01-01

    A lamp system with a very soft high-intensity output is provided over a large area by water cooling a long-arc lamp inside a diffuse reflector of polytetrafluorethylene (PTFE) and titanium dioxide (TiO.sub.2) white pigment. The water is kept clean and pure by a one micron particulate filter and an activated charcoal/ultraviolet irradiation system that circulates and de-ionizes and biologically sterilizes the coolant water at all times, even when the long-arc lamp is off.

  18. New Hydrophilic, Composite Membranes for Air Removal from Water Coolant Systems

    NASA Technical Reports Server (NTRS)

    Ritchie, Stephen M. C.; Luo, Qiang; Curtis, Salina S.; Holladay, Jon B.; Clark, Dallas W.

    2004-01-01

    Liquid coolants are commonly used as thermal transport media to increase efficiency and flexibility in aerospace vehicle design. The introduction of gas bubbles into the coolant can have negative consequences, including: loss of centrifugal pump prime, irregular sensor readings, and blockage of coolant flow to remote systems. One solution to mitigate these problems is the development of a passive gas removal device, or gas trap, installed in the flight cooling system. In this study, a new hydrophilic, composite membrane has been developed for passage of the coolant fluid and retention of gas bubbles. The trapped bubbles are subsequently vented from the system by a thin, hydrophobic, microporous membrane. The original design for this work employed a homogeneous membrane that was susceptible to fouling and pore plugging. Spare gas traps of this variety have degraded during storage, and recreation of the membranes has been complicated due to problems with polymer duplication and property variations in the final membranes. In this work, replacements have been developed based on deposition of a hydrophilic polymer on the bore-side of a porous polyethylene (PE) tube. The tube provides excellent chemical and mechanical stability, and the hydrophilic layer provides retention of gas bubbles. Preliminary results have shown that intimate contact is required between the deposited layer and the substrate to overcome material differences. This has been accomplished by presoaking the membrane tube in the solvent to raise its surface energy. Polymer solutions of various concentrations have been used to promote penetration of the polymer layer into the porous substrate and to control separation layer thickness. The resulting composite membranes have shown repeatable decrease in nitrogen permeability, which is indicative of a decrease in membrane pore size. Studies with water permeation have yielded similar results. We have observed some swelling of the added polymer layer, which

  19. Photoelectrochemical protection of stainless alloys from the stress-corrosion cracking in BWR primary coolant environment

    SciTech Connect

    Akashi, Masatsune; Iso-o, Hiroyuki; Kubota, Nobuhiko; Fukuda, Takanori; Ayabe, Muneo; Hirano, Kenji

    1995-12-31

    The feasibility of counteracting or preventing the stress-corrosion cracking in the BWR core internals by the photoelectrochemical method has been examined. For the purpose TiO{sub 2} semiconductor is noted for its capability of photo electrochemically inducing the water-oxidizing anodic reaction in low enough potential domain if supplied with a light of a wavelength shorter than 410 nm. This paper offers an empirical proof by showing that Type 304 stainless steel and Alloy 600 stainless alloy that have been plasma-spray coated with TiO{sub 2} film will do quite well in environments of BWR primary coolant.

  20. Proceedings of the CSNI specialists meeting on fuel-coolant interactions

    SciTech Connect

    1994-03-01

    A specialists meeting on fuel-coolant interactions was held in Santa Barbara, CA from January 5-7, 1993. The meeting was sponsored by the United States Nuclear Regulatory Commission in collaboration with the Committee on the Safety of Nuclear Installation (CSNI) of the OECD Nuclear Energy Agency (NEA) and the University of California at Santa Barbara. The objectives of the meeting are to cross-fertilize on-going work, provide opportunities for mutual check points, seek to focus the technical issues on matters of practical significance and re-evaluate both the objectives as well as path of future research. Individual papers have been cataloged separately.