Science.gov

Sample records for coolant accident conditions

  1. 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)

  2. Determination of transient radial-azimuthal temperature distributions in fuel bundles under loss-of-coolant-accident conditions

    SciTech Connect

    Saltos, N.T.; Christensen, R.N.; Aldemir, T.

    1988-10-01

    A methodology is presented to determine the transient temperature distributions in fuel bundles under loss-of-coolant accident (LOCA) conditions using a recently developed variational technique for the solution of radial-azimuthal heat conduction in the fuel rods and the modified view factor concept proposed by Uchida and Nakamure to model the radiative heat transfer between the rods. The variational technique is based on the Lebron-Labermont restricted variational principle and represents the temperature distribution in the rods at a given time during the LOCA via parabolic and circular trial functions in the radial and azimuthal directions, respectively. The methodology is implemented to a 4 x 4 boiling water reactor fuel bundle under typical LOCA conditions to investigate the effects of changes in rod heat transfer characteristics and simplifying modeling assumptions on predicted rod temperature distributions. The results show that these effects depend on the rod location in the assembly and LOCA phase under consideration and indicate that same degree of modelling detail may not be necessary for all the rods in the bundle at all times during the LOCA.

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

    NASA Astrophysics Data System (ADS)

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

    2016-12-01

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

  4. 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.

  5. Review of cladding-coolant interactions during LWR accident transients

    SciTech Connect

    Hobson, D.O.

    1980-01-01

    Some of the coolant-cladding interactions that can take place during the design basis loss-of-coolant accident and the Three Mile Island loss-of-coolant accident are analyzed. The physical manifestations of the interactions are quite similar, but the time sequences involved can cause very different end results. These results are described and a listing is given of the main research programs that are involved in coolant-cladding interaction research.

  6. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of class 1E electrical cables

    SciTech Connect

    Jacobus, M.J. )

    1992-11-01

    This report describes the results of aging, condition monitoring, and accident testing of ethylene propylene rubber (EPR) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal ([approx equal]100[degrees]C) and radiation ([approx equal]0.10 kGy/hr) conditions. A sequential accident consisting of high dose rate irradiation ([approx equal]6 kGy/hr) and high temperature steam followed the aging. Also exposed to the accident conditions was a fourth set of cables, which were unaged. The test results indicate that most properly installed EPR cables should be able to survive an accident after 60 years for total aging doses of at least 150--200 kGy and for moderate ambient temperatures on the order of 45--55[degrees]C (potentially higher or lower, depending on material specific activation energies and total radiation doses). Mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation.

  7. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of Class 1E electrical cables: Summary of results

    SciTech Connect

    Jacobus, M.J.

    1991-12-01

    This paper summarizes the results of aging, condition monitoring, and accident testing of Class 1E cables used in nuclear power generating stations. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx_equal} 100{degrees}C) and radiation ({approx_equal}0.10 kGy/hr) conditions. After the aging, the cables were exposed to a simulated accident consisting of high dose rate irradiation ({approx_equal}6 kGy/hr) followed by a high temperature steam exposure. A fourth set of cables, which were unaged, were also exposed to the accident conditions. The cables that were aged for 3 months and then accident tested were subsequently exposed to a high temperature steam fragility test (up to 400{degrees}C), while the cables that were aged for 6 months and then accident tested were subsequently exposed to a 1000-hour submergence test in a chemical solution. The results of the tests indicate that the feasibility of life extension of many popular nuclear power plant cable products is promising and that mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation. In the high temperature steam test, ethylene propylene rubber (EPR) cable materials generally survived to higher temperatures than crosslinked polyolefin (XLPO) cable materials. In dielectric testing after the submergence testing, the XLPO materials performed better than the EPR materials. This paper presents some recent experimental data that are not yet available elsewhere and a summary of findings from the entire experimental program.

  8. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of Class 1E electrical cables: Summary of results

    SciTech Connect

    Jacobus, M.J.

    1991-01-01

    This paper summarizes the results of aging, condition monitoring, and accident testing of Class 1E cables used in nuclear power generating stations. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx equal} 100{degrees}C) and radiation ({approx equal}0.10 kGy/hr) conditions. After the aging, the cables were exposed to a simulated accident consisting of high dose rate irradiation ({approx equal}6 kGy/hr) followed by a high temperature steam exposure. A fourth set of cables, which were unaged, were also exposed to the accident conditions. The cables that were aged for 3 months and then accident tested were subsequently exposed to a high temperature steam fragility test (up to 400{degrees}C), while the cables that were aged for 6 months and then accident tested were subsequently exposed to a 1000-hour submergence test in a chemical solution. The results of the tests indicate that the feasibility of life extension of many popular nuclear power plant cable products is promising and that mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation. In the high temperature steam test, ethylene propylene rubber (EPR) cable materials generally survived to higher temperatures than crosslinked polyolefin (XLPO) cable materials. In dielectric testing after the submergence testing, the XLPO materials performed better than the EPR materials. This paper presents some recent experimental data that are not yet available elsewhere and a summary of findings from the entire experimental program.

  9. 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.

  10. 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.

  11. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

  12. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of class 1E electrical cables. Ethylene propylene rubber cables, Volume 2

    SciTech Connect

    Jacobus, M.J.

    1992-11-01

    This report describes the results of aging, condition monitoring, and accident testing of ethylene propylene rubber (EPR) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx_equal}100{degrees}C) and radiation ({approx_equal}0.10 kGy/hr) conditions. A sequential accident consisting of high dose rate irradiation ({approx_equal}6 kGy/hr) and high temperature steam followed the aging. Also exposed to the accident conditions was a fourth set of cables, which were unaged. The test results indicate that most properly installed EPR cables should be able to survive an accident after 60 years for total aging doses of at least 150--200 kGy and for moderate ambient temperatures on the order of 45--55{degrees}C (potentially higher or lower, depending on material specific activation energies and total radiation doses). Mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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)

  19. Analysis of Kuosheng Large-Break Loss-of-Coolant Accident with MELCOR 1.8.4

    SciTech Connect

    Wang, T.-C.; Wang, S.-J.; Chien, C.-S

    2000-09-15

    The MELCOR code, developed by Sandia National Laboratories, is capable of simulating the severe accident phenomena of light water reactor nuclear power plants (NPPs). A specific large-break loss-of-coolant accident (LOCA) for Kuosheng NPP is simulated with the use of the MELCOR 1.8.4 code. This accident is induced by a double-ended guillotine break of one of the recirculation pipes concurrent with complete failure of the emergency core cooling system. The MELCOR input deck for the Kuosheng NPP is established based on the design data of the Kuosheng NPP and the MELCOR users' guides. The initial steady-state conditions are generated with a developed self-initialization algorithm. The effect of the MELCOR 1.8.4-provided initialization process is demonstrated. The main severe accident phenomena and the corresponding fission product released fractions associated with the large-break LOCA sequences are simulated. The MELCOR 1.8.4 predicts a longer time interval between the core collapse and vessel failure and a higher source term. This MELCOR 1.8.4 input deck will be applied to the probabilistic risk assessment, the severe accident analysis, and the severe accident management study of the Kuosheng NPP in the near future.

  20. 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.

  1. 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.

  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. Safety Analysis of Small Break Loss of Coolant Accident for 1200 MWe Simplified Boiling Water Reactor (SBWR-1200 BDLB)

    SciTech Connect

    Xu, Y.; Revankar, S.T.; Ishii, M.

    2002-07-01

    The objective of this research is to assess the performance of the safety systems during small break loss of coolant accident (SBLOCA) transient in the full size SBWR. RELAP5/MOD3 was used to simulate the blow-down and long-term cooling responses of the various safety systems during the accident transient. An integral test for long-term cooling under low pressure was conducted in a scaled facility with the initial conditions given by the code simulation. The code applicability and the facility scalability were evaluated by the comparison between the test data and the code simulations. The scaling analysis has been done by the comparison of the prototype code predictions and the scaled-up test data with the proper scaling multiplications and time shifting. The good agreement between the major safety parameters has shown the applicability of the RELAP5/MOD3 code and the scalability of the facility for SBWR-1200 safety analysis applications. (authors)

  4. An analytical study of a small-break loss-of-coolant accident with upper head injection

    SciTech Connect

    Leonard, M.T.

    1983-07-01

    Upper head injection (UHI) is an emergency core coolant (ECC) system design that injects subcooled water into the upper head of the reactor vessel in a pressurized water reactor. An analysis has been performed that investigates the effects of UHI on small-break transient behavior. The analysis consists of several RELAP5/MOD1 computer code calculations, which have been compared to experimental data from a series of small-break loss-of-coolant accident simulations, performed in the Semiscale Mod-2A system. Small-break transient phenomena were calculated not to be significantly affected by the introduction of subcooled liquid into the vessel upper head. Nonequilibrium effects were minimal and limited to the period of ECC injection. The analysis covered a range of small-break sizes, and the severity of the transients (in terms of minimum core coolant level) was calculated to be a maximum (with or without UHI) for a cold leg break size of about 5.0% of the cold flow area. For all break sizes, UHI was calculated to increase the margin against core uncovery. The calculated hydraulic phenomena and specific fluid conditions were generally in good agreement with data. The calculated relative magnitudes of important phenomena were preserved over the break size spectrum.

  5. Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

    SciTech Connect

    Gamble, Kyle A.; Hales, Jason D.

    2016-12-01

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

  6. 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)

  7. 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.

  8. 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.

  9. Analysis of ex-core neutron detector response during a loss-of-coolant accident

    SciTech Connect

    Baratta, A.J.; Jester, W.A. ); Gundy, L.M. ); Imel, G.R. )

    1991-06-01

    In this paper the experimental response of ex-core neutron detectors during both actual and simulated loss-of-coolant accidents (LOCAs) at a pressurized water reactor are analyzed to determine their cause. Various analytical techniques are used to reproduce the ex-core detector response during large-break LOCAs. These techniques include both discrete ordinates transport and point kernel calculations. The experiments analyzed include large-break LOCA experiments at the Loss of Fluid Test Facility and from the Three Mile Island accident. The results show that an adiabatic method is sufficiently accurate to reproduce the detector response. This response can be explained in terms of the combined effects of changes in shielding and multiplication that occur in a core during a LOCA.

  10. Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Nakamura, M.; Tobita, K.; Someya, Y.; Utoh, H.; Sakamoto, Y.; Gulden, W.

    2015-11-01

    Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. As for the in-VV LOCA, we analysed the multiple double-ended break of the first wall cooling pipes around the outboard toroidal circumference. As for the ex-VV LOCA, we analysed the double-ended break of the primary cooling pipe. The thermohydraulic analysis results suggest that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. Mitigations of the loads to the confinement barriers are also discussed.

  11. 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.

  12. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    SciTech Connect

    Szilard, Ronaldo Henriques

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

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

    NASA Astrophysics Data System (ADS)

    Gamble, K. A.; Barani, T.; Pizzocri, D.; Hales, J. D.; Terrani, K. A.; Pastore, G.

    2017-08-01

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

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

    DOE PAGES

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

    2017-04-30

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

  15. Estimating Loss-of-Coolant Accident Frequencies for the Standardized Plant Analysis Risk Models

    SciTech Connect

    S. A. Eide; D. M. Rasmuson; C. L. Atwood

    2008-09-01

    The U.S. Nuclear Regulatory Commission maintains a set of risk models covering the U.S. commercial nuclear power plants. These standardized plant analysis risk (SPAR) models include several loss-of-coolant accident (LOCA) initiating events such as small (SLOCA), medium (MLOCA), and large (LLOCA). All of these events involve a loss of coolant inventory from the reactor coolant system. In order to maintain a level of consistency across these models, initiating event frequencies generally are based on plant-type average performance, where the plant types are boiling water reactors and pressurized water reactors. For certain risk analyses, these plant-type initiating event frequencies may be replaced by plant-specific estimates. Frequencies for SPAR LOCA initiating events previously were based on results presented in NUREG/CR-5750, but the newest models use results documented in NUREG/CR-6928. The estimates in NUREG/CR-6928 are based on historical data from the initiating events database for pressurized water reactor SLOCA or an interpretation of results presented in the draft version of NUREG-1829. The information in NUREG-1829 can be used several ways, resulting in different estimates for the various LOCA frequencies. Various ways NUREG-1829 information can be used to estimate LOCA frequencies were investigated and this paper presents two methods for the SPAR model standard inputs, which differ from the method used in NUREG/CR-6928. In addition, results obtained from NUREG-1829 are compared with actual operating experience as contained in the initiating events database.

  16. 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.

  17. 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.

  18. Effects of the reactor-coolant pumps during a small-break loss-of-coolant accident

    SciTech Connect

    Elliott, J.L.

    1983-01-01

    TRAC-PD2 calculations indicate that more coolant mass remains in the system when the reactor coolant pumps are left in operation following a small cold-leg break. The analyses were performed for a Westinghouse plant (Zion-1) to help determine whether to trip the pumps at high-pressure injection initiation (the present operator directive), to trip the pumps at some later time in the transient, or to leave the pumps running indefinitely. The loop seals' behavior and the system refill characteristics primarily determined the results. 9 figures.

  19. 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.

  20. Evaluation of single failure effects during loss of coolant accidents for a VVER-440 reactor

    SciTech Connect

    Shier, W.; Kohut, P.; Horak, W.

    1995-05-01

    This paper describes the results of an analysis of loss of coolant accidents (LOCA`s) for the Soviet designed, light water cooled and moderated reactors referred to as VVERS. The VVER unit selected for this analysis is designated as VVER-440 Model 213. This plant generates 440 MWe and is of current interest since fifteen are now operating and additional units are in various stages of construction within Eastern Europe and the New Independent States. In addition to the analysis of two base case LOCA`S, this paper also presents the results of several sensitivity studies related to single failures in various plant systems that have been included in the design to mitigate the effects of LOCA`s on plant and fuel system performance. Examples of the safety systems selected for these sensitivity studies include the scram system, the accumulators, and the high pressure injection system.

  1. 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.

  2. 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.

  3. 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)

  4. Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors

    SciTech Connect

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

  5. Analysis of the Loss of Forced Reactor Coolant Flow Accident in SMART using RETRAN-03/INT

    SciTech Connect

    Kim, Tae-Wan; Suh, Kune-Yull; Lee, Un-Chul; Park, Goon-Cherl; Kim, Jae-Hak

    2002-07-01

    Small and medium integral type nuclear reactors are getting much attention for the peaceful use of nuclear energy in non-electric area such as district heating, seawater desalination and ship propulsion. An integral type nuclear co-generation reactor, SMART(System-integrated Modular Advanced ReacTor, 330 MWt), has been developed by KAERI (Korea Atomic Energy Research Institute) since 1996. In this study, the safety analysis for SMART using modified RETRAN-03 code whose name is RETRAN-03/INT is performed to examine the applicability of RETRAN-03/INT code. For the safety analysis of integral reactor with helical-coiled steam generators, RETRAN-03 code has been modified and verified using experimental results. New heat transfer coefficients are added for helical-coiled steam generator. And, the heat transfer model for steam generator is modified due to the different primary and secondary side heat flow from U-tube type steam generator. The loss of forced reactor coolant flow accident is selected for safety analysis in this study. Also it is considered as a single failure that one of three trains of passive residual heat removal system is failed. The results from MARS/SMR code and RETRAN-03/INT code are compared. (authors)

  6. 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.

  7. 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.

  8. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2

    SciTech Connect

    Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E.; Marshall, R. K.; Mohr, C. L.

    1981-09-01

    A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500°F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050°F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500°F), on deformed rods.

  9. Evaluation of engine coolants under flow boiling conditions

    SciTech Connect

    McAssey, E.V. Jr.; Stinson, C.; Gollin, M.

    1995-12-31

    An experimental program has been conducted to evaluate the heat transfer performance of two engine coolant mixtures, propylene-glycol/water and ethylene-glycol/water. In each mixture, the concentration was 50-50 by volume. Performance in this situation is defined as the ability to maintain a lower surface temperature for a given flux. The heat transfer regimes considered covered the range from single phase forced convection through saturated flow boiling. Results show that both coolants perform satisfactorily. However, in single phase convection, ethylene-glycol/water is slightly more effective. Conversely, for sub-cooled nucleate boiling and saturated boiling, propylene-glycol/water results in slightly lower metal temperatures.

  10. 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

  11. 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.

  12. 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.

  13. 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.

  14. 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.

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

    SciTech Connect

    Pastore, G.; Novascone, S. R.; Williamson, R. L.; Hales, J. D.; Spencer, B. W.; Stafford, D. S.

    2015-09-01

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

  16. 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.

  17. 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.

  18. 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.

  19. LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor: Appendix A-4

    SciTech Connect

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510/sup 0/C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs.

  20. 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.

  1. 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.

  2. Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

    SciTech Connect

    Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A.; Voeroess, L.; Techy, Z.; Katona, T.

    1998-07-01

    A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

  3. 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).

  4. The response of ex-core neutron detectors to large- and small-break loss-of-coolant accidents in pressurized water reactors

    SciTech Connect

    Okyere, E.W. ); Baratta, A.J.; Jester, W.A. . Dept. of Nuclear Engineering)

    1991-12-01

    This paper reports on a variety of water level measurement systems that are proposed to resolve the problem of reactor vessel level measurement. Two such systems, the heated thermocouple and the multiple differential pressure cell system, are used commercially. A third system based on ex-core neutron detectors was tested at the Pennsylvania State University Breazeale nuclear reactor facility and at the Idaho National Engineering Laboratory Loss-of-Fluid Test Facility. Results of these tests show that such a system is sensitive to both large- and small-break loss-of-coolant accidents and to voiding in the upper plenum of the vessel.

  5. 10 CFR 71.73 - Hypothetical accident conditions.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Package, Special Form, and LSA-III Tests 2 § 71.73 Hypothetical accident conditions. (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the...

  6. 10 CFR 71.73 - Hypothetical accident conditions.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Package, Special Form, and LSA-III Tests 2 § 71.73 Hypothetical accident conditions. (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the...

  7. 10 CFR 71.73 - Hypothetical accident conditions.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Package, Special Form, and LSA-III Tests 2 § 71.73 Hypothetical accident conditions. (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the...

  8. 10 CFR 71.73 - Hypothetical accident conditions.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Package, Special Form, and LSA-III Tests 2 § 71.73 Hypothetical accident conditions. (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the...

  9. 10 CFR 71.73 - Hypothetical accident conditions.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Package, Special Form, and LSA-III Tests 2 § 71.73 Hypothetical accident conditions. (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the...

  10. 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.

  11. 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)

  12. 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.

    1991-01-01

    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.

  13. 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.

  14. Corrosion fatigue behavior of low alloy steels under simulated BWR coolant conditions

    NASA Astrophysics Data System (ADS)

    Huang, J. Y.; Young, M. C.; Jeng, S. L.; Yeh, J. J.; Huang, J. S.; Kuo, R. C.

    2010-10-01

    The corrosion fatigue crack growth behavior of A533 and A508 low alloy steels under simulated boiling water reactor (BWR) coolant conditions was studied. Corrosion fatigue crack growth rates of A533B3 and A508 cl. 3 steels were significantly affected by the steel sulfur content, loading frequency and dissolved oxygen content of water environments. The data points outside the bound of Eason's model could be attributed to the low frequency, higher steel sulfur content and high dissolved oxygen in water environments. The sulfur dissolved in the water environment from the higher-sulfur steels was sufficiently concentrated to acidify the crack tip chemistry even in the hydrogen water chemistry (HWC). Therefore, nitrogenated or HWC water showed little or no beneficiary effect on the high-sulfur steels. For the steel specimens of the same sulfur level, their corrosion fatigue crack growth rates were comparable in different orientations, which could be related to the exposure of fresh sulfides to the water environment. The percentages of sulfides per unit area, by quantitative metallography, were comparable for the steel specimens of both orientations. When the steel sulfur content was decreased to a critical sulfur content 0.005 wt.%, the crack growth rates decreased remarkably.

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

    NASA Astrophysics Data System (ADS)

    Ha, Taesung

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

  16. 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.

  17. Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs

    SciTech Connect

    Sienicki, J.J.; Chu, C.C.; Spencer, B.W.; Frid, W.; Loewenhielm, G.

    1993-01-01

    In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL analyses.

  18. 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

  19. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  20. Evaluation of current regulations and real accident conditions

    SciTech Connect

    Fischer, L.E.; Kimura, C.Y.; Witte, M.C.

    1985-04-05

    In order to improve estimates of the effectiveness of current regulatory standards, a program was initiated by the US Nuclear Regulatory Commission (NRC) to have the Lawrence Livermore Laboratory (LLNL) evaluate regulatory standards against real world accident conditions. This paper presents the results of the evaluation performed for the hypothetical 30-foot drop onto an unyielding surface and real world impact conditions which might be experienced by a spent fuel cask being transported by a truck. The results of the evaluations performed for other pertinent accident conditions for truck and train transport will be documented at the conclusion of the program. 10 refs., 8 figs., 3 tabs.

  1. 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.

  2. The TOPAZ II space reactor response under accident conditions

    SciTech Connect

    Voss, S.S.

    1993-12-31

    The TOPAZ II is a single-cell thermionic space reactor power system developed by the Russians during the period of time from {approximately}1969 to 1989. The TOPAZ II has never been flight demonstrated, but the system was extensively tested on the ground. As part of the development and test program, the response of the TOPAZ II under accident conditions was analyzed and characterized. The US TOPAZ II team has been working closely with the Russian specialists to understand the TOPAZ II system, its operational characteristics, and its response under potential accident conditions. The purpose of the technical exchange is to enable a potential launch of a TOPAZ II by the US. The information is required to integrate the system with a US spacecraft and to support the safety review process. The purpose of this paper is to provide a brief overview of the system and its response under actual and postulated accident conditions.

  3. 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.

  4. Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

    SciTech Connect

    Majumdar, S.

    1996-09-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents.

  5. 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.

  6. The BLOW-3A: A theoretical model to describe transient two phase flow conditions in Liquid Metal Fast Breeder Reactor (LMFBR) coolant channels

    NASA Astrophysics Data System (ADS)

    Bottoni, M.; Struwe, D.

    The theoretical background of the BLOW-3A program is reported, including the basic equations used to determine temperature fields in the fuel, clad, coolant and structure material as well as the coolant dynamics in single and two-phase flow conditions. The two-phase flow model assumes an annular flow regime. Special aspects to calculate two-phase pressure drops for these conditions are discussed. Examples of the experimental validation of the program are given.

  7. 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

  8. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    DOEpatents

    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.

  9. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    SciTech Connect

    Zapata, L.E.; Hackel, L.

    1999-10-26

    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.

  10. 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.

  11. The TOPAZ 2 space reactor response under accident conditions

    NASA Astrophysics Data System (ADS)

    Voss, S. S.

    The TOPAZ 2 is a single-cell thermionic space reactor power system developed by the U.S.S.R. during the period of time from approximately 1969 to 1989. The TOPAZ 2 was never flight demonstrated, but the system was extensively tested on the ground. As part of the development and test program, the response of the TOPAZ 2 under accident conditions was analyzed and characterized. The U.S. TOPAZ 2 team has been working closely with the Russian specialists to understand the TOPAZ 2 system, its operational characteristics, and its response under potential accident conditions. The purpose of the technical exchange is to enable a potential launch of a TOPAZ 2 by the U.S. The information is required to integrate the system with a U.S. spacecraft and to support the safety review process. The purpose of this paper is to provide a brief overview of the system and its response under actual and postulated accident conditions.

  12. Predictions of structural integrity of steam generator tubes under normal operating, accident, an severe accident conditions

    SciTech Connect

    Majumdar, S.

    1997-02-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation was confirmed by further tests at high temperatures, as well as by finite-element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation was confirmed by finite-element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate-sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure was developed and validated by tests under various temperature and pressure loadings that can occur during postulated severe accidents.

  13. Investigation of air cleaning system response to accident conditions

    SciTech Connect

    Andrae, R.W.; Bolstad, J.W.; Foster, R.D.; Gregory, W.S.; Horak, H.L.; Idar, E.S.; Martin, R.A.; Ricketts, C.I.; Smith, P.R.; Tang, P.K.

    1980-01-01

    Air cleaning system response to the stress of accident conditions are being investigated. A program overview and hghlight recent results of our investigation are presented. The program includes both analytical and experimental investigations. Computer codes for predicting effects of tornados, explosions, fires, and material transport are described. The test facilities used to obtain supportive experimental data to define structural integrity and confinement effectiveness of ventilation system components are described. Examples of experimental results for code verification, blower response to tornado transients, and filter response to tornado and explosion transients are reported.

  14. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    SciTech Connect

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  15. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

    NASA Astrophysics Data System (ADS)

    Ott, L. J.; Robb, K. R.; Wang, D.

    2014-05-01

    Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/Zircaloy (Zr) 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. Analyses are presented that illustrate the impact of these new candidate fuel/cladding materials on the fuel performance at normal operating conditions and on the reactor system under DB and BDB accident conditions.

  16. 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.

  17. Experimental database of E110 claddings exposed to accident conditions

    NASA Astrophysics Data System (ADS)

    Perez-Feró, Erzsébet; Győri, Csaba; Matus, Lajos; Vasáros, László; Hózer, Zoltán; Windberg, Péter; Maróti, László; Horváth, Márta; Nagy, Imre; Pintér-Csordás, Anna; Novotny, Tamás

    2010-02-01

    An experimental database of E110 alloy has been developed on the basis of about 600 separate and combined effect tests of the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute. It contains the data of oxidation, ballooning, tensile and compression tests, the results of post-test investigations, photos, figures, information concerning the test conditions and the corresponding English-language publications. The aim of this database is to give adequate information on the E110 cladding behaviour (oxidation, hydrogen uptake, mechanical performance) under accident conditions and to provide valuable experimental data for model development and code validation. This database is a part of the International Fuel Performance Experimental Database. It is accessible on-line, via the internet. This paper gives an overview of the experiments, the test facilities and conditions involved in the database. It presents the most important results and consequences and introduces the directory structure of the database.

  18. Analysis of an Air Conditioning Coolant Solution for Metal Contamination Using Atomic Absorption Spectroscopy: An Undergraduate Instrumental Analysis Exercise Simulating an Industrial Assignment

    ERIC Educational Resources Information Center

    Baird, Michael J.

    2004-01-01

    A real-life analytical assignment is presented to students, who had to examine an air conditioning coolant solution for metal contamination using an atomic absorption spectroscopy (AAS). This hands-on access to a real problem exposed the undergraduate students to the mechanism of AAS, and promoted participation in a simulated industrial activity.

  19. Analysis of an Air Conditioning Coolant Solution for Metal Contamination Using Atomic Absorption Spectroscopy: An Undergraduate Instrumental Analysis Exercise Simulating an Industrial Assignment

    ERIC Educational Resources Information Center

    Baird, Michael J.

    2004-01-01

    A real-life analytical assignment is presented to students, who had to examine an air conditioning coolant solution for metal contamination using an atomic absorption spectroscopy (AAS). This hands-on access to a real problem exposed the undergraduate students to the mechanism of AAS, and promoted participation in a simulated industrial activity.

  20. Hypothetical accident conditions thermal analysis of the 5320 package

    SciTech Connect

    Hensel, S.J.; Gromada, R.J.

    1995-12-31

    An axisymmetric model of the 5320 package was created to perform hypothetical accident conditions (HAC) thermal calculations. The analyses assume the 5320 package contains 359 grams of plutonium-238 (203 Watts) in the form of an oxide powder at a minimum density of 2.4 g/cc or at a maximum density of 11.2 g/cc. The solution from a non-solar 100 F ambient steady-state analysis was used as the initial conditions for the fire transient. A 30 minute 1,475 F fire transient followed by cooling via natural convection and thermal radiation to a 100 F non-solar environment was analyzed to determine peak component temperatures and vessel pressures. The 5320 package was considered to be horizontally suspended within the fire during the entire transient.

  1. ORNL studies of fission product release under LWR accident conditions

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Collins, J.L.

    1991-01-01

    High burnup Zircaloy-clad UO{sub 2} fuel specimens have been heated to study the release of fission products in tests simulating LWR accident conditions. The dominant variable was found to be temperature, with atmosphere, time, and burnup also being significant variables. Comparison of data from tests in steam and hydrogen, at temperatures of 2000 to 2700 K, have shown that the releases of the most volatile species (Kr, Xe, I, and Cs) are relatively insensitive to atmosphere. The releases of the less-volatile species (Sr, Mo, Ru, Sb, Te, Ba, and Eu), however, may vary by orders of magnitude depending on atmosphere. In addition, the atmosphere may drastically affect the mode and extent of fuel destruction.

  2. Risk Assessment of Oil Pipeline Accidents in Special Climatic Conditions

    NASA Astrophysics Data System (ADS)

    Vtorushina, A. N.; Anishchenko, Y. V.; Nikonova, E. D.

    2017-05-01

    The present study identifies the main accidents’ factors and causes for oil pipeline located in Siberia and operated in special climatic conditions. Various types of pipeline accident scenarios were modeled. It is showed that the most dangerous scenarios are oil spills fire and oil vapor explosion due to the loss of piping integrity (rupture) of the pipeline’s section, laying on marshlands and oil spill on the water surface due to the loss of piping integrity (puncture). The most probable scenario is oil spills fire due to the loss of piping integrity (puncture) of the pipeline’s section, laying on dry lands and marshlands. To estimate the scenarios «event tree analysis» is used. Also such risk indexes as individual, societal, public and potential risks were determined.

  3. 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.

  4. Heat transfer performance of engine coolants under sub-cooled boiling conditions

    SciTech Connect

    Bhowmick, S.; Branchi, C.; McAssey, E.V. Jr.; Gollin, M.

    1996-12-31

    An experimental program has been conducted to evaluate the heat transfer performance of two engine cooling fluid mixtures, propylene-glycol/water and ethylene-glycol/water. These tests were performed under conditions closely simulating normal engine operation. For both mixtures, results were obtained over a range of heat transfer regimes from single phase convection to saturated flow boiling. Tests showed that propylene-glycol/water and ethylene-glycol/water have very similar heat transfer performances. Performance is defined as the steady state wall temperature maintained for a given surface heat flux and test section inlet velocity. For the lowest velocity tested, the test section experienced saturated boiling over approximately one-half of its heated length. The experimental results were also compared to analytical predictions based upon the Chen correlation. At higher fluxes, the analytical methods under-predicted the test section wall temperature.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    SciTech Connect

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  10. Behavior of Zr1%Nb Fuel Cladding under Accident Conditions

    SciTech Connect

    Perez-Fero, E.; Hozer, Z.; Windberg, P.; Nagy, I.; Vimi, A.; Ver, N.; Matus, L.; Kunstar, M.; Novotny, T.; Horvath, M.; Gyori, Cs.

    2007-07-01

    The behavior of the VVER fuel (E110) cladding under accident conditions has been investigated at the AEKI in order to study the role of oxidation and hydrogen uptake on the cladding embrittlement and to understand the phenomena that took place during the Paks-2 cleaning tank incident (2003). The test programme covered small scale tests and large scale tests with electrically heated 7-rod bundles in the CODEX (Core Degradation Experiment) facility. Since a hydrogen rich atmosphere could have been formed in the closed tank, the experiments were carried out in hydrogen-steam mixture. According to the results of the small scale tests, a former correlation for the ductile-brittle transitions of E110 in pure steam remained valid in hydrogen rich steam atmosphere as well. During the large scale tests the main conditions of the incident were reconstructed. The test characterized the high temperature oxidation and embrittlement of zirconium in hydrogen rich steam. The observed cladding failure phenomena and the extent of the damage of the test bundle in the quenching phase were very similar to those of the VVER assemblies in the incident. The simulation of the cleaning tank incident provided detailed information on the most probable scenario of the incident. (authors)

  11. Shipping container response to severe highway and railway accident conditions: Main report

    SciTech Connect

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    This report describes a study performed by the Lawrence Livermore National Laboratory to evaluate the level of safety provided under severe accident conditions during the shipment of spent fuel from nuclear power reactors. The evaluation is performed using data from real accident histories and using representative truck and rail cask models that likely meet 10 CFR 71 regulations. The responses of the representative casks are calculated for structural and thermal loads generated by severe highway and railway accident conditions. The cask responses are compared with those responses calculated for the 10 CFR 71 hypothetical accident conditions. By comparing the responses it is determined that most highway and railway accident conditions fall within the 10 CFR 71 hypothetical accident conditions. For those accidents that have higher responses, the probabilities anf potential radiation exposures of the accidents are compared with those identified by the assessments made in the ''Final Environmental Statement on the Transportation of Radioactive Material by Air and other Modes,'' NUREG-0170. Based on this comparison, it is concluded that the radiological risks from spent fuel under severe highway and railway accident conditions as derived in this study are less than risks previously estimated in the NUREG-0170 document.

  12. [The functional condition of patients following a cerebrovascular accident].

    PubMed

    Azevêdo da Costa, F; Araujo da Silva, D L; da Rocha, V M

    Most patients who suffer a cerebrovascular accident (CVA) belong to the adult-elderly age group. Thus, such patients are likely to display a certain degree of functional compromise as a result of the natural process of aging, regardless of whether they have suffered a CVA or not. AIM. To examine the functional condition of patients with CVA, as well as that of healthy individuals who have never suffered a stroke. An analytical cross-sectional study was conducted with a control group made up of 82 individuals -41 of whom were CVA patients (G1) and 41 were individuals with no history of CVA (G2)-. These subjects were paired according to sex, age and educational level. The sample was predominantly made up of females (60%), who were married (70%), with ischaemic CVA (70%), in the right hemisphere of the brain (56%) and with a primary education (36%). The prevalence rate of patients with severe functional impairment was 58.53%, whereas in the control group there were no subjects with this level of deficiency. A relationship was found to exist (p < 0.05) between the functional condition and the variables age and causation of CVA for G1, and also between the functional condition and the variable age for the individuals in G2, which indicates that functional compromise is greater in more elderly subjects. Findings suggest that the functional compromise of patients with CVA is higher than that of individuals with no history of CVA with the same sex, age and educational level. This clearly shows that there is an association between CVA and functional impairment in these patients.

  13. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    SciTech Connect

    Clayton, Dwight A.; Poore, III, Willis P.

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  14. The modelling of fuel volatilisation in accident conditions

    NASA Astrophysics Data System (ADS)

    Manenc, H.; Mason, P. K.; Kissane, M. P.

    2001-04-01

    For oxidising conditions, at high temperatures, the pressure of uranium vapour species at the fuel surface is predicted to be high. These vapour species can be transported away from the fuel surface, giving rise to significant amounts of volatilised fuel, as has been observed during small-scale experiments and taken into account in different models. Hence, fuel volatilisation must be taken into account in the conduct of a simulated severe accident such as the Phebus FPT-4 experiment. A large-scale in-pile test is designed to investigate the release of fission products and actinides from irradiated UO 2 fuel in a debris bed and molten pool configuration. Best estimate predictions for fuel volatilisation were performed before the test. This analysis was used to assess the maximum possible loading of filters collecting emissions and the consequences for the filter-change schedule. Following successful completion of the experiment, blind post-test analysis is being performed; boundary conditions for the calculations are based on the preliminary post-test analysis with the core degradation code ICARE2 [J.C. Crestia, G. Repetto, S. Ederli, in: Proceedings of the Fourth Technical Seminar on the PHEBUS FP Programme, Marseille, France, 20-22 March 2000]. The general modelling approach is presented here and then illustrated by the analysis of fuel volatilisation in Phebus FPT4 (for which results are not yet available). Effort was made to reduce uncertainties in the calculations by improving the understanding of controlling physical processes and by using critically assessed thermodynamic data to determine uranium vapour pressures. The analysis presented here constitutes a preliminary, blind, post-test estimate of fuel volatilised during the test.

  15. Identification of traffic accident risk-prone areas under low-light conditions

    NASA Astrophysics Data System (ADS)

    Ivan, K.; Haidu, I.; Benedek, J.; Ciobanu, S. M.

    2015-09-01

    Besides other non-behavioural factors, low-light conditions significantly influence the frequency of traffic accidents in an urban environment. This paper intends to identify the impact of low-light conditions on traffic accidents in the city of Cluj-Napoca, Romania. The dependence degree between light and the number of traffic accidents was analysed using the Pearson correlation, and the relation between the spatial distribution of traffic accidents and the light conditions was determined by the frequency ratio model. The vulnerable areas within the city were identified based on the calculation of the injury rate for the 0.5 km2 areas uniformly distributed within the study area. The results show a strong linear correlation between the low-light conditions and the number of traffic accidents in terms of three seasonal variations and a high probability of traffic accident occurrence under the above-mentioned conditions at the city entrances/exits, which represent vulnerable areas within the study area. Knowing the linear dependence and the spatial relation between the low light and the number of traffic accidents, as well as the consequences induced by their occurrence, enabled us to identify the areas of high traffic accident risk in Cluj-Napoca.

  16. Identification of traffic accident risk-prone areas under low lighting conditions

    NASA Astrophysics Data System (ADS)

    Ivan, K.; Haidu, I.; Benedek, J.; Ciobanu, S. M.

    2015-02-01

    Besides other non-behavioural factors, the low lighting conditions significantly influence the frequency of the traffic accidents in the urban environment. This paper intends to identify the impact of low lighting conditions on the traffic accidents in the city of Cluj-Napoca. The dependence degree between lighting and the number of traffic accidents was analyzed by the Pearson's correlation and the relation between the spatial distribution of traffic accidents and the lighting conditions was determined by the frequency ratio model. The vulnerable areas within the city were identified based on the calculation of the injured persons rate for the 0.5 km2 equally-sized areas uniformly distributed within the study area. The results have shown a strong linear dependence between the low lighting conditions and the number of traffic accidents in terms of three seasonal variations and a high probability of traffic accidents occurrence under the above-mentioned conditions, at the city entrances-exits, which represent also vulnerable areas within the study area. Knowing the linear dependence and the spatial relation between the low lighting and the number of traffic accidents, as well as the consequences induced by their occurrence enabled us to identify the high traffic accident risk areas in the city of Cluj-Napoca.

  17. A review of monitoring, sampling and analysis of reactor coolant, reactor containment atmosphere and airborne reactor effluents in post accident concentrations

    SciTech Connect

    Hull, A.P.; White, J.R.; Knox, W.H.

    1986-01-01

    A post-implementation review has been made in NRC Region I of the post-accident sampling systems (PASS), the gaseous effluent monitors, and the provisions for sampling effluent particulates and radioiodines which were required by the NRC subsequent to the TMI-2 accident (NUREG-0737). Prefabricated PASS systems were predominant. Problems included insufficient purge times, inadequate separation of dissolved gases, excessive dilution and the accuracy of analytical techniques in the presence of interferences. Microprocessor-controlled high-range gas monitors with integral provisions for sampling particulates and radioiodines in high concentrations were widely used. Calibration information was generally insufficient for the unambiguous conversion of monitor readings to release rates for a varying postaccident mixture of radiogases. The referenced sampling guidance (ANSI-N 13.1-1969) was inappropriate for the long sampling lines customarily used. Generic research is needed to establish the behavior of particulates and radioiodines in these lines.

  18. 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.

  19. PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 1: Code models and correlations

    SciTech Connect

    Sienicki, J.J.

    1997-06-01

    A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. Flashing from coolant release, condensation heat transfer, intercompartment transport, and engineered safety features are described using best estimate models and correlations often based upon experiment analyses. Two notable capabilities of PACER that differ from most other containment loads codes are the modeling of the rates of steam and water formation accompanying coolant release as well as the correlations for steam condensation upon structure.

  20. Chemistry of fission product iodine under nuclear reactor accident conditions

    SciTech Connect

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs.

  1. Accident Conditions versus Regulatory Test for NRC-Approved UF6 Packages

    SciTech Connect

    MILLS, G. SCOTT; AMMERMAN, DOUGLAS J.; LOPEZ, CARLOS

    2003-01-01

    The Nuclear Regulatory Commission (NRC) approves new package designs for shipping fissile quantities of UF{sub 6}. Currently there are three packages approved by the NRC for domestic shipments of fissile quantities of UF{sub 6}: NCI-21PF-1; UX-30; and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR Part 71. The primary objective of this project was to relate the conditions experienced by these packages in the tests described in 10 CFR Part 71 to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR Part 71 tests was achieved by means of computer modeling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from test and other fire scenarios. In addition, the likelihood of encountering bodies of water or sufficient rainfall to cause complete or partial immersion during transport over representative truck routes was assessed. Modeled effects, and their associated probabilities, were combined with existing event-tree data, plus accident rates and other characteristics gathered from representative routes, to derive generalized probabilities of encountering accident conditions comparable to the 10 CFR Part 71 conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents, i.e. the likelihood of UF{sub 6} being dispersed as a result of accident impact or fire is small. Moreover, given that an accident has occurred, exposure to water by fire-fighting, heavy rain or submersion in a body of water is even less probable by factors ranging from 0.5 to 8E-6.

  2. 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.

    1995-09-01

    Core flow blockage events have been determined to represent a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. 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. Hat transfer between melt particle and coolant, which affects the particle breakup characteristics, 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 the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. The results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.

  3. Review and evaluation of the RELAP5YA computer code and the Vermont Yankee LOCA (Loss-of-Coolant Accident) licensing analysis model for use in small and large break BWR (Boiling Water Reactor) LOCAS

    SciTech Connect

    Jones, J.L.

    1987-01-01

    A review has been completed of the RELAP5YA computer code to determine its acceptability for performing licensing analyses. The review was limited to Boiling Water Reactor (BWR) reactor applications. In addition, a Loss-Of-Coolant Accident (LOCA) licensing analysis method, using the RELAP5YA computer code, has been reviewed. This method is applicable to the Vermont Yankee Nuclear Power Station to perform full break spectra LOCA and fuel cycle independent analyses. The review of the RELAP5YA code consisted of an evaluation of all Yankee Atomic Electric Company (YAEC) incorporated modifications to the RELAP5/MOD1 Cycle 18 computer code from which the licensing version of the code originated. Qualifying separate and integral effects assessment calculations were reviewed to evaluate the validity and proper implementation of the various added models. The LOCA licensing method was assessed by reviewing two RELAP5YA system input models and evaluating several small and large break qualifying transient calculations. A review of the RELAP5YA code modifications and their assessments, as well as the submitted LOCA licensing method, is given and the results of the review are provided.

  4. Potential behavior of depleted uranium penetrators under shipping and bulk storage accident conditions

    SciTech Connect

    Mishima, J.; Parkhurst, M.A.; Scherpelz, R.I.

    1985-03-01

    An investigation of the potential hazard from airborne releases of depleted uranium (DU) from the Army's M829 munitions was conducted at the Pacific Northwest Laboratory. The study included: (1) assessing the characteristics of DU oxide from an April 1983 burn test, (2) postulating conditions of specific accident situations, and (3) reviewing laboratory and theoretical studies of oxidation and airborne transport of DU from accidents. Results of the experimental measurements of the DU oxides were combined with atmospheric transport models and lung and kidney exposure data to help establish reasonable exclusion boundaries to protect personnel and the public at an accident site. 121 references, 44 figures, 30 tables.

  5. Correlations for fission product release from N Reactor fuel under high-temperature accident conditions

    SciTech Connect

    Birney, K.R.; Bechtold, D.B.; McCall, T.B.

    1988-03-01

    Empirical correlations were derived for fission product release from metallic uranium alloy 601 N Reactor fuel during postulated accident conditions in which the fuel nears, reaches, or exceeds the melting temperature. The correlations were based on a sparse data base from fuel melted in an inert or steam atmosphere. The empirical correlations are presented for use in subsequent deterministic analyses of N Reactor behavior during hypothetical severe accidents beyond the design basis. 20 refs., 4 figs., 4 tabs.

  6. Accident sequences simulated at the Juragua nuclear power plant

    SciTech Connect

    Carbajo, J.J.

    1998-08-01

    Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident (LOCA) with the emergency core coolant system (ECCS) on, (2) a station blackout (SBO), (3) a small LOCA (SLOCA) concurrent with SBO, (4) a large LOCA (LLOCA) concurrent with SBO, and (5) a LLOCA concurrent with SBO and with the containment breached at time zero. Timings of important events and source term releases have been calculated for the different sequences analyzed. Under certain weather conditions, the fission products released from the severe accident sequences may travel to southern Florida.

  7. 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.

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

    SciTech Connect

    Crosswait, Kenneth Mitchell

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel 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.

  9. Relationship between work-related accidents and hot weather conditions in Tuscany (central Italy).

    PubMed

    Morabito, Marco; Cecchi, Lorenzo; Crisci, Alfonso; Modesti, Pietro Amedeo; Orlandini, Simone

    2006-07-01

    Nowadays, no studies have been published on the relationship between meteorological conditions and work-related mortality and morbidity in Italy. The aim of this study was to evaluate the relationship between hot weather conditions and hospital admissions due to work-related accidents in Tuscany (central Italy) over the period 1998-2003. Apparent temperature (AT) values were calculated to evaluate human weather discomfort due to hot conditions and then tested for work accident differences using non-parametric procedures. Present findings showed that hot weather conditions might represent a risk factor for work-related accidents in Italy during summer. In particular early warming days during June, characterized by heat discomfort, are less tolerated by workers than warming days of the following summer months. The peak of work-related accidents occurred on days characterized by high, but not extreme, thermal conditions. Workers maybe change their behaviour when heat stress increases, reducing risks by adopting preventive measures. Results suggested that days with an average daytime AT value ranged between 24.8 degrees C and 27.5 degrees C were at the highest risk of work-related accidents. In conclusion, present findings might represent the first step for the development of a watch/warning system for workers that might be used by employers for planning work activities.

  10. Steady Heat Removal Test by BWR Drywell Cooler under Accident Management Conditions

    SciTech Connect

    Yokobori, Seiichi; Tobimatsu, Toshimi; Akinaga, Makoto; Fukasawa, Masanori; Nagasaka, Hideo

    2002-07-01

    This paper deals with the heat removal performance of the BWR drywell local cooler (DWC) applied as a Japanese phase-II accident management. Separated effect tests were conducted using a single DWC unit of a typical BWR plant under severe accident (SA) condition. It was demonstrated that noncondensable gas mixture with nitrogen and helium was constantly vented from the DWC casing and the favorable steam condensation rate was maintained even under the highest assumed gas condition. The DWC was found to be promising even under wide range of SA conditions. (authors)

  11. 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.

  12. TECHNICAL BASIS FOR THE NUCLEAR CRITICALITY REPRESENTATIVE ACCIDENT & ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    SciTech Connect

    GOETZ, T.G.

    2003-06-17

    This document was developed to support the documented safety analysis (DSA) and describes the process and basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. Revision 1 incorporates ORP IRT comments to enhance the technical presentation and also makes editorial changes. This technical basis document was developed to support the documented safety analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence.

  13. 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.

  14. 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.

  15. On the effect of accident conditions on the molten core debris relocation into lower head of a PWR vessel

    NASA Astrophysics Data System (ADS)

    An, Xuegao

    From 1975 to present, it has been found that the primary risk to the public health and safety from nuclear power reactors lies in ``beyond design basis'' accidents. During such severe accidents, melting of the reactor core may lead to a loss of primary system integrity, or even containment failure, which will allow escape of significant amounts of radioactive material to the environment. It is very important to understand the mechanism of reactor core degradation during a severe accident. In this study, the damage progression of the reactor core and the slumping mechanism of molten material to the lower head of the reactor vessel were examined through simulation of severe accident scenarios that lead to large-scale core damage. The calculations were carried out using the computer code SCDAP/RELAP5. Different modeling parameters or models were used in calculations by version MOD3.2. The cladding oxidation shell ``durability'' parameter, which can control the timing of fuel clad failure, was varied. The heat flux model of steady-state natural convection of the molten pool was changed. The ultimate strength of the crust supporting the molten pool was doubled. These changes were made to examine the effects on the calculated core damage, and the molten pool expansion and its slumping. Different accident scenarios were simulated. The HPI/makeup flow rates were changed. The timing of opening and closing the PORV was considered. Reflood by restart of coolant pump 2B was also studied. Finally, the size of the PORV opening was also changed. The effects of these accident scenarios on accident progression and core damage process were studied. From the calculated results, it was concluded that the accurate modeling of core damage phenomena was very important to the prediction of the later stage of an accident. According to code MOD3.2, the molten material in a pool slumped to the lower head of the reactor vessel when the juncture of the top and side crusts failed after the

  16. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  17. Historical civilian nuclear accident based Nuclear Reactor Condition Analyzer

    NASA Astrophysics Data System (ADS)

    McCoy, Kaylyn Marie

    There are significant challenges to successfully monitoring multiple processes within a nuclear reactor facility. The evidence for this observation can be seen in the historical civilian nuclear incidents that have occurred with similar initiating conditions and sequences of events. Because there is a current lack within the nuclear industry, with regards to the monitoring of internal sensors across multiple processes for patterns of failure, this study has developed a program that is directed at accomplishing that charge through an innovation that monitors these systems simultaneously. The inclusion of digital sensor technology within the nuclear industry has appreciably increased computer systems' capabilities to manipulate sensor signals, thus making the satisfaction of these monitoring challenges possible. One such manipulation to signal data has been explored in this study. The Nuclear Reactor Condition Analyzer (NRCA) program that has been developed for this research, with the assistance of the Nuclear Regulatory Commission's Graduate Fellowship, utilizes one-norm distance and kernel weighting equations to normalize all nuclear reactor parameters under the program's analysis. This normalization allows the program to set more consistent parameter value thresholds for a more simplified approach to analyzing the condition of the nuclear reactor under its scrutiny. The product of this research provides a means for the nuclear industry to implement a safety and monitoring program that can oversee the system parameters of a nuclear power reactor facility, like that of a nuclear power plant.

  18. Hydrogen-control systems for severe LWR accident conditions - a state-of-technology report

    SciTech Connect

    Hilliard, R K; Postma, A K; Jeppson, D W

    1983-03-01

    This report reviews the current state of technology regarding hydrogen safety issues in light water reactor plants. Topics considered in this report relate to control systems and include combustion prevention, controlled combustion, minimization of combustion effects, combination of control concepts, and post-accident disposal. A companion report addresses hydrogen generation, distribution, and combustion. The objectives of the study were to identify the key safety issues related to hydrogen produced under severe accident conditions, to describe the state of technology for each issue, and to point out ongoing programs aimed at resolving the open issues.

  19. Study of light water reactor containments under important severe accident conditions

    SciTech Connect

    Hofmayer, C.H.; Pratt, W.T.; Bagchi, G.; Noonan, V.S.

    1985-01-01

    The US Nuclear Regulatory Commission has sponsored studies to develop a ''LEAKAGE-BEFORE-FAILURE'' model for use in severe accident risk assessments to provide a means of accounting for significant containment leakage prior to reaching the containment threshold pressure. Six containment types have been studied (large dry, subatmospheric, ice condenser, Mark I, II, and III). Potential leak paths through major containment penetration assemblies were investigated and upper-bound estimates of leak areas established. These leak areas may result from increasing internal pressure and degradation of nonmetallic seal materials due to severe accident conditions. This paper describes the approach and summarizes the results and conclusions of this study.

  20. The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

    SciTech Connect

    Cinbiz, Mahmut N; Brown, Nicholas R; Terrani, Kurt A; Lowden, Rick R; ERDMAN III, DONALD L

    2017-01-01

    This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of both accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.

  1. Analysis of concrete containment structures under severe accident loading conditions

    SciTech Connect

    Porter, V.L.

    1993-12-31

    One of the areas of current interest in the nuclear power industry is the response of containment buildings to internal pressures that may exceed design pressure levels. Evaluating the response of structures under these conditions requires computing beyond design load to the ultimate load of the containment. For concrete containments, this requirement means computing through severe concrete cracking and into the regime of wide-spread plastic rebar and/or tendon response. In this regime of material response, an implicit code can have trouble converging. This paper describes some of the author`s experiences with Version 5.2 of ABAQUS Standard and the ABAQUS concrete model in computing the axisymmetric response of a prestressed concrete containment to ultimate global structural failure under high internal pressures. The effects of varying the tension stiffening parameter in the concrete material model and variations of the parameters for the CONTROLS option are discussed.

  2. 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 ...

  3. 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 ...

  4. PRESSURE INTEGRITY OF 3013 CONTAINER UNDER POSTULATED ACCIDENT CONDITIONS

    SciTech Connect

    Rawls, G.

    2010-02-01

    A series of tests was carried out to determine the threshold for deflagration-to-detonation transition (DDT), structural loading, and structural response of the Department of Energy 3013 storage systems for the case of an accidental explosion of evolved gas within the storage containers. Three experimental fixtures were used to examine the various issues and three mixtures consisting of either stoichiometric hydrogen-oxygen, stoichiometric hydrogen-oxygen with added nitrogen, or stoichiometric hydrogen-oxygen with an added nitrogen-helium mixture were tested. Tests were carried out as a function of initial pressure from 1 to 3.5 bar and initial temperature from room temperature to 150 C. The elevated temperature tests resulted in a slight increase in the threshold pressure for DDT. The elevated temperature tests were performed to ensure the test results were bounding. Because the change was not significant the elevated temperature data are not presented in the paper. The explosions were initiated with either a small spark or a hot surface. Based on the results of these tests under the conditions investigated, it can be concluded that DDT of a stoichiometric hydrogen-oxygen mixture (and mixtures diluted with nitrogen and helium) within the 3013 containment system does not pose a threat to the structural integrity of the outer container.

  5. Experimental research to investigate the performance of bio coolant when turning of mild carbon steel

    NASA Astrophysics Data System (ADS)

    Agus Susanto, Tri; Nur, Rusdi

    2017-04-01

    Some literatures have been reported that the using bio coolant show better lubricating and cooling performances and reduce the occupational health risks associated with petroleum-oil-based coolant since they have lower toxicity. This paper investigates the effect the cutting conditions on the surface roughness through turning of mild carbon steel using dry, coolant and bio coolant. Measurement of surface roughness was conducted and then compared with the change of the cutting conditions. The relationship between surface roughness and cutting conditions was created in a curve for different of the cutting speed and coolant. The results indicate that the surface roughness was reduced when the speed of cutting is set to the highest level for all of coolant conditions (dry, coolant, and bio coolant) and constant of DOC and feed. The surface roughness had better performance using bio coolant than coolant conventional (mineral fluid).

  6. Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

    SciTech Connect

    Hoover, M.D.; Farrell, R.F.; Newton, G.J.

    1995-12-01

    The recent 1995 WIPP Safety Analysis Report (SAR) Update provided detailed analyses of potential radiation doses to members of the public at the site boundary during postulated accident scenarios at the U.S. Department of Energy`s Waste Isolation Pilot Plant (WIPP). The SAR Update addressed the complete spectrum of potential accidents associated with handling and emplacing transuranic waste at WIPP, including damage to waste drums from fires, punctures, drops, and other disruptions. The report focused on the adequacy of the multiple layers of safety practice ({open_quotes}defense-in-depth{close_quotes}) at WIPP, which are designed to (1) reduce the likelihood of accidents and (2) limit the consequences of those accidents. The safeguards which contribute to defense-in-depth at WIPP include a substantial array of inherent design features, engineered controls, and administrative procedures. The SAR Update confirmed that the defense-in-depth at WIPP is adequate to assure the protection of the public and environment. As a supplement to the 1995 SAR Update, we have conducted additional analyses to confirm that these controls will also provide adequate protection to workers at the WIPP. The approaches and results of the worker dose assessment are summarized here. In conformance with the guidance of DOE Standard 3009-94, we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposures under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR Update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, members of the public, and the environment.

  7. Effects of coolant volatility on simulated HCDA bubble expansions. [LMFBR

    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). In past experiments, water was used to simulate the sodium coolant in the prototype. In these experiments, 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.

  8. Estimation of maximum temperature in a package subjected to hypothetical-accident thermal-test conditions

    SciTech Connect

    Shah, V.L.

    1996-05-01

    A simple reading-the-graph procedure has been developed to provide a quick estimate of maximum temperature in a package during hypothetical-accident thermal-test conditions. The procedure, based on applying the analytical solutions of a transient thermal-diffusion equation for temperatures in a semi-infinite solid and in an infinitely long circular cylinder subjected to step surface temperature boundary conditions, is applicable to all packages, with or without an internal heat source, that are protected by rectangular or cylindrical thermal insulating overpacks. Simple to follow, the procedure provides a conservative estimate of package maximum temperatures during hypothetical-accident thermal-test conditions. To demonstrate its applicability, the procedure is applied to the Mound 9859 tritium trap package and two values of the thermal diffusivity of the insulation in the overpack.

  9. Coolant passage heat transfer with rotation

    NASA Technical Reports Server (NTRS)

    Hajek, T. J.; Higgins, A. W.

    1985-01-01

    The objective 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.

  10. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  11. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    SciTech Connect

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  12. Causal Factors and Adverse Conditions of Aviation Accidents and Incidents Related to Integrated Resilient Aircraft Control

    NASA Technical Reports Server (NTRS)

    Reveley, Mary S.; Briggs, Jeffrey L.; Evans, Joni K.; Sandifer, Carl E.; Jones, Sharon Monica

    2010-01-01

    The causal factors of accidents from the National Transportation Safety Board (NTSB) database and incidents from the Federal Aviation Administration (FAA) database associated with loss of control (LOC) were examined for four types of operations (i.e., Federal Aviation Regulation Part 121, Part 135 Scheduled, Part 135 Nonscheduled, and Part 91) for the years 1988 to 2004. In-flight LOC is a serious aviation problem. Well over half of the LOC accidents included at least one fatality (80 percent in Part 121), and roughly half of all aviation fatalities in the studied time period occurred in conjunction with LOC. An adverse events table was updated to provide focus to the technology validation strategy of the Integrated Resilient Aircraft Control (IRAC) Project. The table contains three types of adverse conditions: failure, damage, and upset. Thirteen different adverse condition subtypes were gleaned from the Aviation Safety Reporting System (ASRS), the FAA Accident and Incident database, and the NTSB database. The severity and frequency of the damage conditions, initial test conditions, and milestones references are also provided.

  13. Study of Air Ingress Across the Duct During the Accident Conditions

    SciTech Connect

    Hassan, Yassin

    2013-05-06

    The goal of this project is to study the fundamental physical phenomena associated with air ingress in very high temperature reactors (VHTRs). Air ingress may occur due to a rupture of primary piping and a subsequent breach in the primary pressure boundary in helium-cooled and graphite-moderated VHTRs. Significant air ingress is a concern because it introduces potential to expose the fuel, graphite support rods, and core to a risk of severe graphite oxidation. Two of the most probable air ingress scenarios involve rupture of a control rod or fuel access standpipe, and rupture in the main coolant pipe on the lower part of the reactor pressure vessel. Therefore, establishing a fundamental understanding of air ingress phenomena is critical in order to rationally evaluate safety of existing VHTRs and develop new designs that minimize these risks. But despite this importance, progress toward development these predictive capabilities has been slowed by the complex nature of the underlying phenomena. The combination of inter-diffusion among multiple species, molecular diffusion, natural convection, and complex geometries, as well as the multiple chemical reactions involved, impose significant roadblocks to both modeling and experiment design. The project team will employ a coordinated experimental and computational effort that will help gain a deeper understanding of multiphased air ingress phenomena. This project will enhance advanced modeling and simulation methods, enabling calculation of nuclear power plant transients and accident scenarios with a high degree of confidence. The following are the project tasks: Perform particle image velocimetry measurement of multiphase air ingresses; and, Perform computational fluid dynamics analysis of air ingress phenomena.

  14. Cooling Characteristics of the V-1650-7 Engine. 1; Coolant-Flow Distribution, Cylinder Temperatures, and Heat Rejections at Typical Operating Conditions

    NASA Technical Reports Server (NTRS)

    Povolny, John H.; Bogdan, Louis J.

    1947-01-01

    An investigation was conducted to determine the coolant-flow distribu tion, the cylinder temperatures, and the heat rejections of the V-165 0-7 engine . The tests were run a t several power levels varying from minimum fuel consumption to war emergency power and at each power l evel the coolant flows corresponded to the extremes of those likely t o be encountered in typical airplane installations, A mixture of 30-p ercent ethylene glycol and 70-percent water was used as the coolant. The temperature of each cylinder was measured between the exhaust val ves, between the intake valves, in the center of the head, on the exh aust-valve guide, at the top of the barrel on the exhaust side, and o n each exhaust spark-plug gasket. For an increase in engine power fro m 628 to approximately 1700 brake horsepower the average temperature for the cylinder heads between the exhaust valves increased from 437 deg to 517 deg F, the engine coolant heat rejection increased from 12 ,600 to 22,700 Btu. per minute, the oil heat rejection increased from 1030 to 4600 Btu per minute, and the aftercooler-coolant heat reject ion increased from 450 to 3500 Btu -per minute.

  15. 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)

  16. Shipping container response to severe highway and railway accident conditions: Appendices

    SciTech Connect

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

  17. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  18. Dose evaluation in criticality accident conditions using transient critical facilities fueled with a fissile solution.

    PubMed

    Nakamura, T; Tonoike, K; Miyoshi, Y

    2004-01-01

    Neutron dose measurement and evaluation techniques in criticality accident conditions using a thermo luminescence dosemeter (TLD) was studied at the Transient Experiment Critical Facility (TRACY) of Japan Atomic Energy Research Institute (JAERI). In the present approach, the absorbed dose is derived from the ambient dose equivalent measured with a TLD, using the appropriate conversion factor given by computation. Using this technique, the neutron dose around the SILENE reactor of the Institute for Radioprotection and Nuclear Safety (IRSN) of France was measured in the Accident Dosimetry Intercomparison Exercise (June 10-21, 2002) organized by OECD/NEA and IRSN. In this exercise, the gamma dose was also measured with a TLD. In this report, measurements and evaluation results at TRACY and SILENE are presented.

  19. a Study of the Interferences with the On-Line Radioiodine Measurement Under Nuclear Accident Conditions

    NASA Astrophysics Data System (ADS)

    Tseng, Tung-Tse

    In this research the interferences with the on -line detection of radioiodines, under nuclear accident conditions, were studied. The special tool employed for this research is the developed on-line radioiodine monitor (the Penn State Radioiodine Monitor), which is capable of detecting low levels of radioiodine on-line in air containing orders of magnitude higher levels of radioactive noble gases. Most of the data reported in this thesis were collected during a series of experiments called "Source -Term Experiment Program (STEP)." The experiments were conducted at the Argonne National Laboratory's TREAT reactor located at the Idaho National Engineering Laboratory (INEL). In these tests, fission products were released from the Light Water Reactor (LWR) test fuels as a result of simulating a reactor accident. The Penn State Monitor was then used to sample the fission products accumulated in a large container which simulated the reactor containment building. The test results proved that the Penn State Monitor was not affected significantly by the passage of large amounts of noble gases through the system. Also, it confirmed the predicted results that the operation of conventional on-line radioiodine detectors would, under nuclear accident conditions, be seriously impaired by the passage of high concentrations of radioactive noble gases through such systems. This work also demonstrated that under conditions of high noble gas concentrations and low radioiodine concentrations, the formation of noble-gas-decayed alkali metals can seriously interfere with the on-line detection of radioiodine, especially during the 24 hours immediately after the accident. The decayed alkali metal particulates were also found to be much more penetrating than the ordinary type of particulates, since a large fraction (15%) of the particulates were found to penetrate through the commonly used High Efficiency Particulate Air (HEPA) filter (rated >99.97% for 0.3 (mu)m particulate). Also, a

  20. Analysis 320 coal mine accidents using structural equation modeling with unsafe conditions of the rules and regulations as exogenous variables.

    PubMed

    Zhang, Yingyu; Shao, Wei; Zhang, Mengjia; Li, Hejun; Yin, Shijiu; Xu, Yingjun

    2016-07-01

    Mining has been historically considered as a naturally high-risk industry worldwide. Deaths caused by coal mine accidents are more than the sum of all other accidents in China. Statistics of 320 coal mine accidents in Shandong province show that all accidents contain indicators of "unsafe conditions of the rules and regulations" with a frequency of 1590, accounting for 74.3% of the total frequency of 2140. "Unsafe behaviors of the operator" is another important contributory factor, which mainly includes "operator error" and "venturing into dangerous places." A systems analysis approach was applied by using structural equation modeling (SEM) to examine the interactions between the contributory factors of coal mine accidents. The analysis of results leads to three conclusions. (i) "Unsafe conditions of the rules and regulations," affect the "unsafe behaviors of the operator," "unsafe conditions of the equipment," and "unsafe conditions of the environment." (ii) The three influencing factors of coal mine accidents (with the frequency of effect relation in descending order) are "lack of safety education and training," "rules and regulations of safety production responsibility," and "rules and regulations of supervision and inspection." (iii) The three influenced factors (with the frequency in descending order) of coal mine accidents are "venturing into dangerous places," "poor workplace environment," and "operator error."

  1. 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.

  2. Status report of advanced cladding modeling work to assess cladding performance under accident conditions

    SciTech Connect

    B.J. Merrill; Shannon M. Bragg-Sitton

    2013-09-01

    Scoping simulations performed using a severe accident code can be applied to investigate the influence of advanced materials on beyond design basis accident progression and to identify any existing code limitations. In 2012 an effort was initiated to develop a numerical capability for understanding the potential safety advantages that might be realized during severe accident conditions by replacing Zircaloy components in light water reactors (LWRs) with silicon carbide (SiC) components. To this end, a version of the MELCOR code, under development at the Sandia National Laboratories in New Mexico (SNL/NM), was modified by replacing Zircaloy for SiC in the MELCOR reactor core oxidation and material properties routines. The modified version of MELCOR was benchmarked against available experimental data to ensure that present SiC oxidation theory in air and steam were correctly implemented in the code. Additional modifications have been implemented in the code in 2013 to improve the specificity in defining components fabricated from non-standard materials. An overview of these modifications and the status of their implementation are summarized below.

  3. Polyalphaolefins: A New Improved Cost Effective Aircraft Radar Coolant

    DTIC Science & Technology

    1991-06-01

    the silicate ester dielectric coolant used in many different aircraft electronic radar systems. Weather and environmental conditions, which can be...Naval Air Test Center in Patuxenti River, Maryland. 2. LABORATORY EVALUATION AND ASSESSMENT Several critical requirements of a dielectric coolant must...Study, Conclusions/Recommendations from Final Report, March, 1988. 3. PAO Summary Report - "Extended Testing of Poly Alpha Olefin as A Dielectric / Coolant Fluid

  4. Role of Winter Weather Conditions and Slipperiness on Tourists’ Accidents in Finland

    PubMed Central

    Lépy, Élise; Rantala, Sinikka; Huusko, Antti; Nieminen, Pentti; Hippi, Marjo; Rautio, Arja

    2016-01-01

    (1) Background: In Finland, slippery snowy or icy ground surface conditions can be quite hazardous to human health during wintertime. We focused on the impacts of the variability in weather conditions on tourists’ health via documented accidents during the winter season in the Sotkamo area. We attempted to estimate the slipping hazard in a specific context of space and time focusing on the weather and other possible parameters, responsible for fluctuations in the numbers of injuries/accidents; (2) Methods: We used statistical distributions with graphical illustrations to examine the distribution of visits to Kainuu Hospital by non-local patients and their characteristics/causes; graphs to illustrate the distribution of the different characteristics of weather conditions; questionnaires and interviews conducted among health care and safety personnel in Sotkamo and Kuusamo; (3) Results: There was a clear seasonal distribution in the numbers and types of extremity injuries of non-local patients. While the risk of slipping is emphasized, other factors leading to injuries are evaluated; and (4) Conclusions: The study highlighted the clear role of wintery weather conditions as a cause of extremity injuries even though other aspects must also be considered. Future scenarios, challenges and adaptive strategies are also discussed from the viewpoint of climate change. PMID:27537899

  5. Role of Winter Weather Conditions and Slipperiness on Tourists' Accidents in Finland.

    PubMed

    Lépy, Élise; Rantala, Sinikka; Huusko, Antti; Nieminen, Pentti; Hippi, Marjo; Rautio, Arja

    2016-08-15

    (1) BACKGROUND: In Finland, slippery snowy or icy ground surface conditions can be quite hazardous to human health during wintertime. We focused on the impacts of the variability in weather conditions on tourists' health via documented accidents during the winter season in the Sotkamo area. We attempted to estimate the slipping hazard in a specific context of space and time focusing on the weather and other possible parameters, responsible for fluctuations in the numbers of injuries/accidents; (2) METHODS: We used statistical distributions with graphical illustrations to examine the distribution of visits to Kainuu Hospital by non-local patients and their characteristics/causes; graphs to illustrate the distribution of the different characteristics of weather conditions; questionnaires and interviews conducted among health care and safety personnel in Sotkamo and Kuusamo; (3) RESULTS: There was a clear seasonal distribution in the numbers and types of extremity injuries of non-local patients. While the risk of slipping is emphasized, other factors leading to injuries are evaluated; and (4) CONCLUSIONS: The study highlighted the clear role of wintery weather conditions as a cause of extremity injuries even though other aspects must also be considered. Future scenarios, challenges and adaptive strategies are also discussed from the viewpoint of climate change.

  6. A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

    SciTech Connect

    Brown, G. S.; Cashwell, J. W.; Apple, M. L.

    1991-01-01

    Shipments of radioactive material (RAM) constitute but a small fraction of the total hazardous materials shipped in the United States each year. Public perception, however, of the potential consequences of a release from a transportation package containing RAM has resulted in significant regulation of transport operations, both to ensure the integrity of a package in accident conditions and to place operational constraints on the shipper. Much of this attention has focused on shipments of spent nuclear fuel and high level wastes which, although comprising a very small number of total shipments, constitute a majority of the total curies transported on an annual basis. This report discusses the shipment of these highly radioactive materials.

  7. Temporal Change of Environmental Contamination Conditions in Five Years after the Fukushima Accident

    NASA Astrophysics Data System (ADS)

    Saito, Kimiaki

    2017-09-01

    The temporal change of environmental contamination conditions after the Fukushima accident have been clarified based on large-scale environmental monitoring data repeatedly obtained in the 80 km zone. The decreasing tendency of air dose rates was confirmed to obviously depend on land uses. In human-related diverse environments the air dose rates have decreased much faster than the physical decay of radiocesium. The horizontal movement of radiocesium in undisturbed fields were found to be generally quite small, though it has gradually penetrated into the deeper parts of the ground.

  8. Hypothetical accident condition thermal analysis and testing of a Type B drum package

    SciTech Connect

    Hensel, S.J.; Alstine, M.N. Van; Gromada, R.J.

    1995-07-01

    A thermophysical property model developed to analytically determine the thermal response of cane fiberboard when exposed to temperatures and heat fluxes associated with the 10 CFR 71 hypothetical accident condition (HAC) has been benchmarked against two Type B drum package fire test results. The model 9973 package was fire tested after a 30 ft. top down drop and puncture, and an undamaged model 9975 package containing a heater (21W) was fire tested to determine content heat source effects. Analysis results using a refined version of a previously developed HAC fiberboard model compared well against the test data from both the 9973 and 9975 packages.

  9. 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.

  10. DYNAMIC ANALYSIS OF HANFORD UNIRRADIATED FUEL PACKAGE SUBJECTED TO SEQUENTIAL LATERAL LOADS IN HYPOTHETICAL ACCIDENT CONDITIONS

    SciTech Connect

    Wu, T

    2008-04-30

    Large fuel casks present challenges when evaluating their performance in the Hypothetical Accident Conditions (HAC) specified in the Code of Federal Regulations Title 10 part 71 (10CFR71). Testing is often limited by cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing by using simplified analytical methods. This paper presents a numerical technique for evaluating the dynamic responses of large fuel casks subjected to sequential HAC loading. A nonlinear dynamic analysis was performed for a Hanford Unirradiated Fuel Package (HUFP) [1] to evaluate the cumulative damage after the hypothetical accident Conditions of a 30-foot lateral drop followed by a 40-inch lateral puncture as specified in 10CFR71. The structural integrity of the containment vessel is justified based on the analytical results in comparison with the stress criteria, specified in the ASME Code, Section III, Appendix F [2], for Level D service loads. The analyzed cumulative damages caused by the sequential loading of a 30-foot lateral drop and a 40-inch lateral puncture are compared with the package test data. The analytical results are in good agreement with the test results.

  11. Evaluation of molten lead mixing in sodium coolant by diffusion for application to PAHR. [LMFBR

    SciTech Connect

    Chawla, T.C.; Pedersen, D.R.; Leaf, G.; Minkowycz, W.J.

    1983-01-01

    In post-accident heat removal (PAHR) applications the use of a lead slab is being considered for protecting a porous bed of steel shots in ex-vessel cavity from direct impingement of molten steel or fuel upon vessel failure following a hypothetical core dissembly accident in an LMFBR. The porous bed is provided to increase coolability of the fuel debris by the sodium coolant. The objectives of the present study are (1) to determine melting rates of lead slabs of various thicknesses in contact with sodium coolant and (2) to evaluate the extent of penetration and mixing rates of molten lead into sodium coolant by molecular diffusion alone.

  12. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  13. Containment performance of prototypical reactor containments subjected to severe accident conditions

    SciTech Connect

    Klamerus, E.W.; Bohn, M.P.; Wesley, D.A.; Krishnaswamy, C.N.

    1996-12-01

    In SECY-90-016, the NTRC proposed a safety goal of a conditional containment failure probability (CCFP) of 0.1 and the alternative acceptance criteria allowed for steel containments, which specifies that the stresses should not exceed ASNE Level C allowables for severe accident pressures and temperatures. In this work, the need for an equivalent criterion for concrete containments was studied. Six surrogate containments were designed and analyzed in order to compare the margins between design pressure, pressure resulting in exceedance of Level C (or yield) stress limits, and ultimate pressure. For comparability, each containment has an identical internal volume and design pressure. Results from the analysis showed margins to yield are comparable and display a similar margin for both steel and concrete containments. In addition, the margin to failure, although slightly higher in the steel containments, were also comparable. Finally, a CCFP for code design was determined based on general membrane behavior and imposing an upper bound severe accident curve developed in the DCH studies. The resulting CCFP`s were less then 0.02 (or 2%) for all the surrogate containments studied, showing that these containment designs all achieved the NRC safety goal.

  14. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    SciTech Connect

    Audin, L.

    1990-12-01

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs.

  15. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    SciTech Connect

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  16. Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2

    SciTech Connect

    Lambert, L.D.; Parks, M.B.

    1995-10-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ``like-new`` condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ``like-new`` condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report.

  17. Severe accident testing of electrical penetration assemblies

    SciTech Connect

    Clauss, D.B. )

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.

  18. Reactor coolant pump flywheel

    DOEpatents

    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.

  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. Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

    NASA Astrophysics Data System (ADS)

    Yano, Y.; Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T.; Ukai, S.; Oono, N.; Kimura, A.; Hayashi, S.; Torimaru, T.

    2017-04-01

    Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900-1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

  1. A model for nonvolatile fission product release during reactor accident conditions

    SciTech Connect

    Lewis, B.J.; Andre, B.; Ducros, G.; Maro, D.

    1996-10-01

    An analytical model has been developed to describe the release kinetics of nonvolatile fission products (e.g., molybdenum, cerium, ruthenium, and barium) from uranium dioxide fuel under severe reactor accident conditions. This treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of antimony and tellurium in the Zircaloy cladding. This model interprets the release behavior of fission products observed in Commissariat a l`Energie Atomique experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere.

  2. Probabilistic dose assessment of normal operations and accident conditions for an assured isolation facility in Texas

    NASA Astrophysics Data System (ADS)

    Arno, Matthew Gordon

    Texas is investigating building a long-term waste storage facility, also known as an Assured Isolation Facility. This is an above-ground low-level radioactive waste storage facility that is actively maintained and from which waste may be retrieved. A preliminary, scoping-level analysis has been extended to consider more complex scenarios of radiation streaming and skyshine by using the computer code Monte Carlo N-Particle (MCNP) to model the facility in greater detail. Accidental release scenarios have been studied in more depth to better assess the potential dose to off-site individuals. Using bounding source term assumptions, the projected radiation doses and dose rates are estimated to exceed applicable limits by an order of magnitude. By altering the facility design to fill in the hollow cores of the prefabricated concrete slabs used in the roof over the "high-gamma rooms," where the waste with the highest concentration of gamma emitting radioactive material is stored, dose rates outside the facility decrease by an order of magnitude. With the modified design, the annual dose at the site fenceline is estimated at 86 mrem, below the 100 mrem annual limit for exposure of the public. Within the site perimeter, the dose rates are lowered sufficiently such that it is not necessary to categorize many workers and contractor personnel as radiation workers, saving on costs as well as being advisable under ALARA principles. A detailed analysis of bounding accidents incorporating information on the local meteorological conditions indicate that the maximum committed effective dose equivalent from the passage of a plume of material released in an accident at any of the cities near the facility is 59 :rem in the city of Eunice, NM based on the combined day and night meteorological conditions. Using the daytime meteorological conditions, the maximum dose at any city is 7 :rem, also in the city of Eunice. The maximum dose at the site boundary was determined to be 230 mrem

  3. 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

  4. 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

  5. Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

    SciTech Connect

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean

    2016-08-01

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

  6. Generation IV benchmarking of TRISO fuel performance models under accident conditions: Modeling input data

    SciTech Connect

    Collin, Blaise P.

    2014-09-01

    This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document

  7. GEN-IV BENCHMARKING OF TRISO FUEL PERFORMANCE MODELS UNDER ACCIDENT CONDITIONS MODELING INPUT DATA

    SciTech Connect

    Collin, Blaise Paul

    2016-09-01

    This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: • The modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release. • The modeling of the AGR-1 and HFR-EU1bis safety testing experiments. • The comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from “Case 5” of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. “Case 5” of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to “effects of the numerical calculation method rather than the physical model” [IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read

  8. Thermochemistry of Ruthenium Oxyhydroxide Species and Their Impact on Volatile Speciations in Severe Nuclear Accident Conditions.

    PubMed

    Miradji, Faoulat; Virot, François; Souvi, Sidi; Cantrel, Laurent; Louis, Florent; Vallet, Valérie

    2016-02-04

    Literature thermodynamic data of ruthenium oxyhydroxides reveal large uncertainties in some of the standard enthalpies of formation, motivating the use of high-level relativistic correlated quantum chemical methods to reduce the level of discrepancies. Reaction energies leading to the formation of all possible oxyhydroxide species RuOx(OH)y(H2O)z have been calculated for a series of reactions combining DFT (TPSSh-5%HF) geometries and partition functions, CCSD(T) energies extrapolated to the complete basis set limits. The highly accurate ab initio thermodynamic data were used as input data of thermodynamic equilibrium computations to derive the speciation of gaseous ruthenium species in the temperature, pressure and concentration conditions of severe nuclear accidents occurring in pressurized water reactors. At temperatures lower than 1000 K, gaseous ruthenium tetraoxide is the dominating species, between 1000 and 2000 K ruthenium trioxide becomes preponderant, whereas at higher temperatures gaseous ruthenium oxide, dioxide and even Ru in gaseous phase are formed. Although earlier studies predicted the formation of oxyhydroxides in significant quantities, the use of highly accurate ab initio thermodynamic data for ruthenium gaseous species leads to a more reliable inventory of gaseous ruthenium species in which gaseous oxyhydroxide ruthenium molecules are formed only in negligible amounts.

  9. Coolant passage heat transfer with rotation

    NASA Technical Reports Server (NTRS)

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

    1986-01-01

    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.

  10. Simulation of the transient processes of load rejection under different accident conditions in a hydroelectric generating set

    NASA Astrophysics Data System (ADS)

    Guo, W. C.; Yang, J. D.; Chen, J. P.; Peng, Z. Y.; Zhang, Y.; Chen, C. C.

    2016-11-01

    Load rejection test is one of the essential tests that carried out before the hydroelectric generating set is put into operation formally. The test aims at inspecting the rationality of the design of the water diversion and power generation system of hydropower station, reliability of the equipment of generating set and the dynamic characteristics of hydroturbine governing system. Proceeding from different accident conditions of hydroelectric generating set, this paper presents the transient processes of load rejection corresponding to different accident conditions, and elaborates the characteristics of different types of load rejection. Then the numerical simulation method of different types of load rejection is established. An engineering project is calculated to verify the validity of the method. Finally, based on the numerical simulation results, the relationship among the different types of load rejection and their functions on the design of hydropower station and the operation of load rejection test are pointed out. The results indicate that: The load rejection caused by the accident within the hydroelectric generating set is realized by emergency distributing valve, and it is the basis of the optimization for the closing law of guide vane and the calculation of regulation and guarantee. The load rejection caused by the accident outside the hydroelectric generating set is realized by the governor. It is the most efficient measure to inspect the dynamic characteristics of hydro-turbine governing system, and its closure rate of guide vane set in the governor depends on the optimization result in the former type load rejection.

  11. Transient two-phase performance of LOFT reactor coolant pumps

    SciTech Connect

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

  12. Analysis of the TMI-2 source range monitor during the TMI (Three Mile Island) accident

    SciTech Connect

    Wu, Horng-Yu; Baratta, A.J.; Hsiao, Ming-Yuan; Bandini, B.R.

    1987-06-01

    The source range monitor (SRM) data recorded during the first 4 hours of the Three Mile Island Unit No. 2 (TMI-2) accident following reactor shutdown were analyzed. An effort to simulate the actual SRM response was made by performing a series of neutron transport calculations. Primary emphasis was placed on simulating the changes in SRM response to various system events during the accident, so as to obtain useful information about core conditions at the various stages. Based on the known end-state reactor conditions, the major system events, and the acutal SRM readings, self-consistent estimates were made of core liquid level, void fraction in the coolant, and locations of core materials. This analysis expands the possible interpretation of the SRM data relative to core damage progression. The results appear to be consistent with other studies of the TMI-2 Accident Evaluation Program, and provide information useful for the developemnt and determination of the TMI-2 accident scenario.

  13. Performance of metal and oxide fuel cores during accidents in large liquid-metal-cooled reactors

    SciTech Connect

    Royl, P.H.; Kussmaul, G. ); Cahalan, J.E.; Wigeland, R.A. ); Friedel, G. ); Moreau, J. ); Perks, M. )

    1992-02-01

    This paper reports on a cooperative effort among European and U.S. analysts, which is an assessment of the comparative safety performance of metal and oxide fuels during accidents in a 3500-MW (thermal), pool-type, liquid-metal-cooled reactor (LMR) is performed. The study focuses on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower, and the unprotected loss-of-heat-sink (ULOHS). Core designs with a similar power output that have been previously analyzed in Europe under ULOF accident conditions are also included in this comparison. Emphasis is placed on identification of design features that provide passive, self-limiting responses to postulated accident conditions and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than do oxide-fueled reactors of the same design.

  14. Varenicline and risk of psychiatric conditions, suicidal behaviour, criminal offending, and transport accidents and offences: population based cohort study.

    PubMed

    Molero, Yasmina; Lichtenstein, Paul; Zetterqvist, Johan; Gumpert, Clara Hellner; Fazel, Seena

    2015-06-02

    To examine associations between varenicline and the incidence of a range of adverse outcomes. Population based cohort study using within person analyses to control for confounding by indication. Whole population of Sweden. 7,917,436 people aged 15 and over, of whom 69,757 were treated with varenicline between 2006 and 2009. Incidence of new psychiatric conditions, suicidal behaviour, suspected and convicted criminal offending, transport accidents, and suspected and convicted traffic offences. In the whole population, 337,393 new psychiatric conditions were diagnosed during follow-up. In addition, 507,823 suspected and 338,608 convicted crimes, 40,595 suicidal events, 124,445 transport accidents, and 99,895 suspected and 57,068 convicted traffic crimes were recorded. Within person analyses showed that varenicline was not associated with significant hazards of suicidal behaviour, criminal offending, transport accidents, traffic offences, or psychoses. However, varenicline was associated with a small increase in the risk of anxiety conditions (hazard ratio 1.23, 95% confidence interval 1.01 to 1.51) and mood conditions (1.31, 1.06 to 1.63), which was only seen in people with pre-existing psychiatric disorders. Concerns that varenicline is associated with an increased risk of many adverse outcomes, including suicidality and accidents, are not supported in this observational study. The small increase in risk of two psychiatric conditions in people with pre-existing psychiatric disorders needs to be confirmed using other research designs. © Molero et al 2015.

  15. Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents

    SciTech Connect

    Shafer, J.M.; Oberlander, P.L.; Skaggs, R.L.

    1984-04-01

    The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables.

  16. 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.

  17. LOCA with consequential or delayed LOOP accidents: Unique issues, plant vulnerability, and CDF contributions

    SciTech Connect

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

    1998-08-01

    A loss-of-coolant accident (LOCA) can cause a loss-of-offsite power (LOOP) wherein the LOOP is usually delayed by few seconds or longer. Such an accident is called LOCA with consequential LOOP, or LOCA with delayed LOOP (here, abbreviated as LOCA/LOOP). This paper analyzes the unique conditions that are associated with a LOCA/LOOP, presents a model, and quantifies its contribution to core damage frequency (CDF). The results show that the CDF contribution can be a dominant contributor to risk for certain plant designs, although boiling water reactors (BWRs) are less vulnerable than pressurized water reactors (PWRs).

  18. 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.

  19. Treatment of mixed waste coolant

    SciTech Connect

    Kidd, S.; Bowers, J.S.

    1995-09-01

    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 meetings 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 {open_quotes}Best Demonstrated Available Technologies,{close_quotes} 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 mutiphasic 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.

  20. 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.

  1. Assessment on Integrity of BWR Internals Against Impact Load by Water Hammer Under Conditions of Reactivity Initiated Accidents

    SciTech Connect

    Azuma, Mie; Taniguchi, Atsushi; Hotta, Akitoshi; Ohta, Takeshi

    2005-03-15

    The integrity of the reactor pressure vessel (RPV) head and reactor internals was assessed by means of fluid and fluid-structural coupled analyses to evaluate the water hammer phenomenon arising from postulated high burnup fuel failure under reactivity initiated accident (RIA) conditions. The fluid viscosity effect on the water column burst as well as the complex three-dimensional flow paths caused by a core shroud and standpipes were considered in this study. It is shown that fluid viscosity becomes an influential factor to dissipate impacting kinetic energy. Integrity of the RPV head and the shroud head was ensured with a sufficient level of margin even under these excessively conservative RIA conditions.

  2. Hungarian surveillance of germinal mutations. Lack of detectable increase in indicator conditions caused by germinal mutations following the Chernobyl accident.

    PubMed

    Czeizel, A

    1989-07-01

    The Hungarian surveillance of germinal mutations is based on three indicator conditions seen in offspring, i.e., 15 sentinel anomalies, Down syndrome and component anomaly pairs of unidentified multiple congenital anomalies. It is an "opportunistic program," because the necessary data are available from the Hungarian Congenital Malformation Registry. This system is described and the criteria of a good registry are summarized. The analysis of indicator conditions caused by germinal mutations did not reveal any measurable mutagenic effects in Hungary following the accident at the Chernobyl nuclear power plant. The pros and cons of germinal mutation surveillance are discussed.

  3. 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.

  4. 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.

  5. 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.

  6. How propeller suction is the dominant factor for ship accidents at shallow water conditions

    NASA Astrophysics Data System (ADS)

    Acar, Dursun; Alpar, Bedri; Ozeren, Sinan

    2017-04-01

    The laminar flow comes to the fore with the disappearance of the several other directions in the internal displacements in the water current. Due to the dominant speed direction during the straightforward motion of the ship, the underwater hull is associated with the continuous flow of laminar currents. The open marine environment acts as a compressible liquid medium because of the presence of many variables about water volume overflow boundaries where the ship is associated. Layers of water rising over the sea surface due to ship's body and the propeller's water push provides loss of liquid lifting force for the ship. These situations change the well-known sea-floor morphology and reliable depth limits, and lead to probable accidents. If the ship block coefficient for the front side is 0.7 or higher, the "squat" will be more on the bow, because the associated factor "displacement volume" causes to the low-pressure environment due to large and rapid turbulence. Thus, the bow sinks further, which faced with liquid's weaker lift force. The vessels Gerardus Mercator, Queen Elizabeth and Costa Concordia had accidents because of unified reasons of squat, fast water mass displacement by hull push and propeller suction interaction. In the case of water mass displacement from the bow side away, that accident occurred in 2005 by the vessel Gerardus Mercator with excessive longitudinal trim angularity in the shallow water. The vessel Costa Concordia (2012), voluminous water displaced from the rear left side was an important factor because of the sharp manoeuvre of that the captain made before the accident. Observations before the accident indicate that full-speed sharp turn provided listed position for the ship from left (port side) in the direction of travel before colliding and then strike a rock on the sloping side of the seabed. The reason why the ship drifted to the left depends mainly the water discharge occurred at the left side of the hull during left-hand rudder

  7. 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)

  8. Vertical reactor coolant pump instabilities

    NASA Technical Reports Server (NTRS)

    Jones, R. M.

    1985-01-01

    The investigation conducted at the Tennessee Valley Authority's Sequoyah Nuclear Power Plant to determine and correct increasing vibrations in the vertical reactor coolant pumps is described. Diagnostic procedures to determine the vibration causes and evaluate the corrective measures taken are also described.

  9. 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.

  10. 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.

  11. Experimental Investigation of Coolant Boiling in a Half-Heated Circular Tube - Final CRADA Report

    SciTech Connect

    Yu, Wenhua; Singh, Dileep; France, David M.

    2016-11-01

    Coolant subcooled boiling in the cylinder head regions of heavy-duty vehicle engines is unavoidable at high thermal loads due to high metal temperatures. However, theoretical, numerical, and experimental studies of coolant subcooled flow boiling under these specific application conditions are generally lacking in the engineering literature. The objective of this project was to provide such much-needed information, including the coolant subcooled flow boiling characteristics and the corresponding heat transfer coefficients, through experimental investigations.

  12. 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.

  13. Analysis of the source range monitor during the first four hours of the Three Mile Island Unit 2 accident

    SciTech Connect

    Wu, H.Y.; Bandini, B.R. ); Hsiao, M.Y.; Baratta, A.J.; Bandini, B.R. . Dept. of Nuclear Engineering); Tolman, E.L. )

    1989-02-01

    The source range monitor (SRM) data recorded during the first 4 h of the Three Mile Island Unit 2 (TMI-2) accident following reactor shutdown were analyzed. An effort to simulate the actual SRM response was made by performing a series of neutron transport calculations. Primary emphasis was placed on simulating the changes in SRM response to various system events during the accident so as to obtain useful information about core conditions at the various stages. Based on the known end-state reactor conditions, the major system events and the actual SRM readings, self-consistent estimates were made of core liquid level, void fraction in the coolant, and locations of core materials. This analysis expands the possible interpretation of the SRM data relative to core damage progression. The results appear to be consistent with other studies of the TMI-2 Accident Evaluation Program, and provide information useful for the development and determination of the TMI-2 accident scenario.

  14. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    SciTech Connect

    Weber, P.; Umminger, K.J.; Schoen, B.

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  15. Coolant monitoring apparatus for nuclear reactors. [PWR; BWR

    SciTech Connect

    Tokarz, R.D.

    1981-08-06

    A system for monitoring coolant conditions within a pressurized vessel is described. 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.

  16. 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.

  17. Formation and characterization of fission-product aerosols under postulated HTGR accident conditions

    SciTech Connect

    Tang, I.N.; Munkelwitz, H.R.

    1982-07-01

    The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release source, a vapor deposition tube, and a filter assembly for collecting particulate matter. Simulated fission products Sr and Ba as oxides are separately impregnated in H451 graphite wafers and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperature. The release and transport of simulated fission product Ag as metal are also investigated.

  18. Radiolysis of cesium iodide solutions in conditions prevailing in a pressurized water reactor severe accident

    SciTech Connect

    Lucas, M. )

    1988-08-01

    Measurements were made of I/sub 2/ formed when aqueous cesium iodide (CsI) solutions were exposed to two temperatures, 43 and 95/sup 0/C, with irradiation. Iodine partition coefficients were obtained from the experiments. The parameters varied were dose, CsI concentration, and Cs/sub 2/CO/sub 3/ concentration, in the presence of air-carbon dioxide and air-carbon dioxide-hydrogen mixtures, to provide information to calculate the form in which iodine released from fuel as CsI in a reactor accident might reach the environment. In a series of experiments, a two-compartment cell was used to trap the gaseous iodine produced. In this case, it was found that the quantity of gaseous iodine produced increased approximately linearly with the dose (at the dose rate used).

  19. 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.

  20. Severe accident sequences simulated at a VVER 440-213

    SciTech Connect

    Carbajo, J.J.

    1998-09-01

    The MELCOR severe accident code version 1.8.4QK has been employed to calculate the timing of events and source terms released into the environment for several hypothetical severe accidents as Unit 1 of the Juragua VVER 440-213 plant, located near Cienfuegos, Cuba. Different severe accidents were simulated, and results from two of them are presented here. The two accidents presented are a station blackout and a large loss-of-coolant accident (LOCA) guillotine break in one cold leg concurrent with loss of all power. For both accidents, core safety injection systems, feedwater systems, and sprays are not available. Only the four passive accumulators are operational.

  1. [Assessment of the suitability of workers based on the Swiss Law on Accident Insurance--declaration of unsuitability / declaration of conditional suitability].

    PubMed

    Jost, M; Rast, H

    2007-08-01

    In order to prevent the development or aggravation of an occupational disease or an occupational accident risk lying in an individual employee the Swiss National Accident Insurance Fund (Suva) based on the Swiss Law on Accident Insurance is entitled to prohibit certain activities that have proven to be hazardous to the individual (declaration of "unsuitability" (DOU) or rarely declaration of conditional suitability). Occupational skin and respiratory diseases are the most frequent disorders, and flour; cement, isocyanates and hairdressing substances the most important occupational factors requiring a DOU. This measure also provides financial security to the employee for up to 4 years by the accident insurer in cooperation with the unemployment insurance. For retraining and occupational rehabilitation a special state insurance, the Invalidity insurance is responsible.

  2. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    NASA Astrophysics Data System (ADS)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  3. Accident investigation

    NASA Technical Reports Server (NTRS)

    Brunstein, A. I.

    1979-01-01

    Aircraft accident investigations are discussed with emphasis on those accidents that involved weather as a contributing factor. The organization of the accident investigation board for air carrier accidents is described along with the hearings, and formal report preparation. Statistical summaries of the investigations of general aviation accidents are provided.

  4. Creep failure of a reactor pressure vessel lower head under severe accident conditions

    SciTech Connect

    Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y.; Rashid, Y.R.

    1998-08-01

    A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

  5. 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.

  6. Influence of main variables modifications on accident transient based on AP1000-like MELCOR model

    NASA Astrophysics Data System (ADS)

    Malicki, M.; Pieńkowski, L.

    2016-09-01

    Analysis of Severe Accidents (SA) is one of the most important parts of nuclear safety researches. MELCOR is a validated system code for severe accident analysis and as such it was used to obtain presented results. Analysed AP1000 model is based on publicly available data only. Sensitivity analysis was done for the main variables of primary reactor coolant system to find their influence on accident transient. This kind of analysis helps to find weak points of reactor design and the model itself. Performed analysis is a base for creation of Small Modular Reactor (SMR) generic model which will be the next step of the investigation aiming to estimate safety level of different reactors. Results clearly help to establish a range of boundary conditions for main the variables in future SMR model.

  7. Criticality safety assessment of a TRIGA reactor spent-fuel pool under accident conditions

    SciTech Connect

    Glumac, B; Ravnik, M.; Logar, M.

    1997-02-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10{sup {minus}6} per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool.

  8. YNPS main coolant system decontamination

    SciTech Connect

    Metcalf, E.T.

    1996-12-31

    The Yankee Nuclear Power Station (YNPS) located in Rowe, Massachusetts, is a four-loop pressurized water reactor that permanently ceased power operation on February 26, 1992. Decommissioning activities, including steam generator removal, reactor internals removal, and system dismantlement, have been in progress since the shutdown. One of the most significant challenges for YNPS in 1996 was the performance of the main coolant system chemical decontamination. This paper describes the objectives, challenges, and achievements involved in the planning and implementation of the chemical decontamination.

  9. 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.

  10. Thermodynamic evaluation of the solidification phase of molten core-concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

    NASA Astrophysics Data System (ADS)

    Kitagaki, Toru; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro

    2017-04-01

    The solidification phases of molten core-concrete under the estimated molten core-concrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool, FactSage 6.2, and the NUCLEA database in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core-concrete forms (U,Zr)O2 and (Zr,U)SiO4, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO4 requires a relatively long time because it involves a change in the crystal structure from fcc-(U,Zr)O2 to tet-(U,Zr)O2, followed by the formation of (Zr,U)SiO4 by reaction with SiO2. Therefore, the formation of (Zr,U)SiO4 is limited under quenching conditions. Other common phases are the oxide phases, CaAl2Si2O8, SiO2, and CaSiO3, and the metallic phases of the Fe-Si and Fe-Ni alloys. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

  11. 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.

  12. Recent condition of Fukushima-Daiichi nuclear plant accident in Japan

    NASA Astrophysics Data System (ADS)

    Ohnishi, Takeo

    2012-07-01

    Japanese government pronounced that the second step had been succeeded in the cooling down of the reactors on the middle of Dec 2011 at Fukushima-Daiichi nuclear power plant. In future, government aims to take out fuels from 4 reactors and shields their units. The nuclear power plants in Japan are gradually decreasing, because the checking for them has been performed and the permission of the re-start of them are difficult to be gained. On January 1st 2012, only 7 units are operating in Japan, though the about 54 units were set before the accident. At the end of December 2011, most radiations are emitted from cesium. The radioactivity in air and land around the plant was daily reported in newspaper. Government often gave the information about some RI-contamination in foods. They were taken off from the markets. At now stage, the most important project is the decontamination of radioactive materials from houses, schools, public facilities and industries. Government will newly classify three evacuation areas from April 2012. At the end of March, evacuees under 20 mSv/year possibly can go back their homes (evacuation-free area). The environmental doses will be depressed by decontamination under 10 mSv/year. At the range of 20-50 mSv, people will be controlled to live these area, they can go back their houses temporally (evacuation area). Over 50 mSv/year, however, people can go back house until 5 years at least (prohibited area). In new radiation limitation for a risk of human health, government made 100 mSv and 20 mSv for life span for one year, respectively. The aim of decontamination was set up to 10 mSv for 1 year and 5 mSv for next stage. A target at school is under1 mSv for children. Government accepted a new severe limitation per1 Kg at four groups; milk of baby (100 Bq) and milk (100 Bq), drinking water (10 Bq) and food (100 Bq). Tokyo electric Power Company and government should pay the sufficient compensation to evacuees. In future, they should keep health

  13. Activity ratios in soil contaminated by the source of different reactor condition in the FDNPP accident

    NASA Astrophysics Data System (ADS)

    Satou, Yukihiko; Sueki, Keisuke; Sasa, Kimikazu; Matsunaka, Tetsuya; Shibayama, Nao; Takahashi, Tsutomu; Kinoshita, Norikazu

    2014-05-01

    The Fukushima Dai-ichi Nuclear power plant (FDNPP) accident caused radioactive contamination on the surface soil at Fukushima and its adjacent prefectures. Substantial contamination has been found in the northwestern area from the FDNPP, according to the airborne monitoring and ground base survey by the Japanese government. Activity ratios would have characteristic information on emission sources because each relevant reactor had different amount of radionuclide and different activity ratio. The ratios can be used to clarify more detailed source and process in the contamination. We have addressed to consider them in Namie town, northwestern region from the FDNPP. This study focused on the gamma-ray emitting radionuclides of 134Cs, 137Cs, and 110mAg. The activities were decay-corrected as of 11th March, 2011 when all nuclear reactors scrammed. Data of activity ratios by our results and the Japanese official report classified the investigated northwestern region into 3 groups. Ratios of 0.02 for 110mAg/137Cs and 0.90 for 134Cs/137Cs were observed in the northern region of 15 km inside from the FDNPP. On the other hand, two kinds of 110mAg/137Cs ratios of 0.005 and 0.002 were distributed broadly in the region 60 km away from the plant. The 134Cs/137Cs ratio was 0.98 there. The activity ratios of 110mAg/137Cs and 134Cs/137Cs in the northern region from the FDNPP correspond to those of nuclear fuel in Unit 1 according to estimation using the ORIGEN code. The 134Cs/137Cs in the northwestern area from FDNPP agrees with that of Unit 2 and 3. The 110mAg/137Cs ratios of 0.005 and0.002 are 1/5 - 1/10 of the Unit 2 and 3. Official report has announced that discharges of the radionuclides from Unit 2 and 3 occurred on 14th March, 2011. It is known that contamination in the northwestern region from the FDNPP took place on 15th March, 2011. Plausible species for silver in reactor core, metal, and halide etc. have higher boiling point than those species for cesium. The core would

  14. Cleaning of aluminum after machining with coolants

    SciTech Connect

    Roop, B.

    1995-07-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended.

  15. 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.

  16. Accidents waiting to happen: the contribution of latent conditions to patient safety

    PubMed Central

    Lowe, C M

    2006-01-01

    Poor design of elements in a healthcare system produce the latent conditions which result in patient safety incidents. A better understanding of these elements and specific healthcare design challenges will result in improved patient safety. PMID:17142613

  17. Design and fabrication of magnetic coolant filter

    NASA Astrophysics Data System (ADS)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  18. Coolant passage heat transfer with rotation

    NASA Technical Reports Server (NTRS)

    Rohde, J. E.

    1982-01-01

    Although the effects of the coriolis and buoyancy forces due to rotation on coolant-side heat transfer are generally not included in the design methods for blades, the influence of these forces could be large. Comparisons of nonrotating heat transfer data and extrapolations of available correlation for the average heat transfer coefficients with radial outflow of cooling air showed that neglecting rotation at gas turbine engine conditions result in variations in the heat transfer coefficient by as much as 45 percent. This, in effect, results in blade metal temperatures running as much as 100 F different from predicted values. This also may explain why rotating blade metal temperatures in engine tests are often higher than expected from results obtained in nonrotating cascade tests.

  19. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 34 2013-07-01 2013-07-01 false Coolants. 1065.745 Section 1065.745 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...

  20. FASTGRASS: A mechanistic model for the prediction of Xe, I, Cs, Te, Ba, and Sr release from nuclear fuel under normal and severe-accident conditions

    SciTech Connect

    Rest, J.; Zawadzki, S.A. )

    1992-09-01

    The primary physical/chemical models that form the basis of the FASTGRASS mechanistic computer model for calculating fission-product release from nuclear fuel are described. Calculated results are compared with test data and the major mechanisms affecting the transport of fission products during steady-state and accident conditions are identified.

  1. Creation of Computational Benchmarks for LEU and MOX Fuel Assemblies Under Accident Conditions

    SciTech Connect

    Pavlovitchev, A M; Kalashnikov, A G; Kalugin, M A; Lazarenko, A P; Maiorov, L V; Sidorenko, V D

    1999-11-01

    The result of VVER-1000 computational benchmarks, calculations obtained with the use of various Russian codes (such as MCU-RFFI/A, TVS-M and WIMS-ABBN) are presented. List of benchmarks includes LEU and MOX cells with fresh and spent fuel under various conditions (for calculation of kinetic parameters, Doppler coefficient, reactivity effect of decreasing the water density). Calculations results are compared with each other and results of this comparison are discussed.

  2. [The rehabilitation under alpine conditions of the participants in the cleanup of the accident at the Chernobyl Atomic Electric Power Station who are ill with chronic bronchitis].

    PubMed

    Brimkulov, N N; Abdulina, A A; Davletalieva, N E; Bakirova, A N; Karamuratov, A; Mirrakhimov, M M

    1996-01-01

    24 patients exposed to low-dose radiation after the Chernobyl accident were examined before and after 24-day treatment of chronic bronchitis in the high-altitude rehabilitation center (3200 m above the sea level) in Tien Shan. Sanogenic alpine climate improved the patients' general condition, physical performance and lung ventilation, corrected compromised immunity. After high-altitude adaptation tracheobronchial inflammation alleviated, cytologic composition and surface activity of bronchoalveolar fluid returned to normal. Therefore, high-altitude treatment of Chernobyl accident victims with chronic bronchitis is effective and can be recommended for such patients.

  3. 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.

  4. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    SciTech Connect

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  5. Feasibility study of superconducting power cables for DC electric railway feeding systems in view of thermal condition at short circuit accident

    NASA Astrophysics Data System (ADS)

    Kumagai, Daisuke; Ohsaki, Hiroyuki; Tomita, Masaru

    2016-12-01

    A superconducting power cable has merits of a high power transmission capacity, transmission losses reduction, a compactness, etc., therefore, we have been studying the feasibility of applying superconducting power cables to DC electric railway feeding systems. However, a superconducting power cable is required to be cooled down and kept at a very low temperature, so it is important to reveal its thermal and cooling characteristics. In this study, electric circuit analysis models of the system and thermal analysis models of superconducting cables were constructed and the system behaviors were simulated. We analyzed the heat generation by a short circuit accident and transient temperature distribution of the cable to estimate the value of temperature rise and the time required from the accident. From these results, we discussed a feasibility of superconducting cables for DC electric railway feeding systems. The results showed that the short circuit accident had little impact on the thermal condition of a superconducting cable in the installed system.

  6. Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in `like-new` conditions

    SciTech Connect

    Lambert, L.D.; Parks, M.B.

    1994-09-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the `like-new` condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed.

  7. 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.

  8. SAS4A: A computer model for the analysis of hypothetical core disruptive accidents in liquid metal reactors

    SciTech Connect

    Tentner, A.M.; Birgersson, G.; Cahalan, J.E.; Dunn, F.E.; Kalimullah; Miles, K.J.

    1987-01-01

    To ensure that the public health and safety are protected under any accident conditions in a Liquid Metal Fast Breeder Reactor (LMFBR), many accidents are analyzed for their potential consequences. The SAS4A code system, described in this paper, provides such an analysis capability, including the ability to analyze low probability events such as the Hypothetical Core Disruptive Accidents (HCDAs). The SAS4A code system has been designed to simulate all the events that occur in a LMFBR core during the initiating phase of a Hypothetical Core Disruptive Accident. During such postulated accident scenarios as the Loss-of-Flow and Transient Overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling and fuel and cladding melting and relocation. During to the strong neutronic feedback present in a nuclear reactor, these events can significantly influence the reactor power. The SAS4A code system is used in the safety analysis of nuclear reactors, in order to estimate the energetic potential of very low probability accidents. The results of SAS4A simulations are also used by reactor designers in order to build safer reactors and eliminate the possibility of any accident which could endanger the public safety.

  9. Cloud conditions for low atmospheric electricity during disturbed period after the Fukushima nuclear accident

    NASA Astrophysics Data System (ADS)

    Yatagai, Akiyo; Yamauchi, Masatoshi; Ishihara, Masahito; Watanabe, Akira; Murata, Ken T.

    2016-04-01

    The vertical (downward) component of the atmospheric electric field, or potential gradient (PG) under cloud generally reflects the electric charge distribution in the cloud. The PG data at Kakioka, 150 km southwest of the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) suggested that this relation can be modified when the radioactive dust was floating in the air, and the exact relation between the weather and this modification could lead to new insight in plasma physics in the wet atmosphere. Unfortunately the detailed weather data was not available above Kakioka (only the precipitation data was available). Therefore, estimation of the cloud condition during March 2011 was strongly needed. We have developed various meteorological information links (http://www.chikyu.ac.jp/akiyo/firis/) and original radar and precipitation data will be released from the page. Here we present various radar images that we have prepared for March 2011. We prepared three-dimensional radar reflectivity of the C-band radar of JMA in every 10 minutes over all Kanto Plain centered at Tokyo and Fukushima prefecture centered at Sendai. We have released images of each altitude (1km interval) for 15th - 16thand 21th March (http://sc-web.nict.go.jp/fukushima/). The vertical structure of the rainfall is almost the same at 4km with the surface and sporadic high precipitation is observed at 6 km height for 15-16th. While, generally precipitation pattern that is similar to the surface is observed at 5km height on 21th. On the other hand, an X-band radar centered at Fukushima university is also used to know more localized raindrop patterns at zenith angle of 4 degree. We prepared 10-minutes/120m mesh precipitation patterns for March 15th, 16th, 17th, 18th, 20th, 21th, 22th and 23th. Quantitative estimate is difficult from this X-band radar, but localized structure, especially for the rain-band along Nakadori (middle valley in Fukushima prefecture), that is considered to determine the highly

  10. Flow boiling test of GDP replacement coolants

    SciTech Connect

    Park, S.H.

    1995-08-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C{sub 4}F{sub 10} mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C{sub 4}F{sub 10} weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd.

  11. 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.

  12. 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.

  13. 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.

  14. Probabilities of Ground Impact Conditions of the New Horizons Spacecraft and RTG for Near Launch Pad Accidents

    NASA Astrophysics Data System (ADS)

    McGrath, Brian E.; Frostbutter, Dave A.; Chang, Yale

    2007-01-01

    As part of the Pluto New Horizons mission's safety effort, assessment of accidental ground impacts of the spacecraft (SC) and its components, including the radioisotope thermoelectric generator (RTG), near the launch pad are of particular interest as they determine the severity of the mechanical insult to the hardware. Two configurations are studied: the SC with RTG joined to the third stage STAR™ 48B solid rocket motor [Launch Vehicle (LV) payload], and the RTG joined to the RTG mounting fixture but separated from the SC after an at-altitude destruct action. The objective of the analyses conducted is to determine the probabilities of impact orientation and average impact velocity of these configurations for a near launch pad accident These are of interest because of the possibility that the STAR 48B solid rocket motor could impact on top of the RTG, and because the RTG/RTG mounting fixture impact orientations probabilities and velocities directly affect the mechanical response of the internal GPHS modules. The probabilities of impact orientation and impact velocity of the LV payload as a function of mission elapsed time at thrust termination are determined using a six degree of freedom motion simulation computer program coupled with a Monte Carlo method. The motion simulation accounts for the LV payload aerodynamic properties, mass properties, and the initial flight conditions (αt, γ, V, q and r). Baseline conditions for position, direction, velocity and angular rates, are obtained from the mission timeline information for the Atlas V 551 launch vehicle. The results from this new and unique approach contributed information to safety assessments for the launch approval process. As the environments associated with the RTG/RTG mounting fixture impact orientations probabilities and velocities were less severe than earlier assumptions, this contributed to a reduction in the estimated risk for the Pluto mission.

  15. 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.

  16. 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.

  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. 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.

  19. Accident liability.

    PubMed Central

    Kuné, J B

    1985-01-01

    The idea of accident proneness, which originated in the early 1900s, has proved to be ineffectual as an operational concept. Discrete econometric methods may be useful to find out which factors are at work in the process that leads to accidents and whether there are individuals who are more liable to accidents than others. PMID:3986144

  20. 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.

  1. Minimizing thermal damage of aerospace components using coolant nozzle and coolant system optimization

    SciTech Connect

    Mindek, R.B. Jr.; Webster, J.A.

    1994-12-31

    Research to optimize the application of coolant in the creep feed grinding of aerospace components was conducted at the Center for Grinding Research and Development during the past year. During this research, work was performed in the areas of coolant jet nozzle and coolant system design to identify optimum jet nozzle designs, nozzle positioning and coolant system configurations. The knowledge gained from initial screening tests and grinding trials of flat surfaces was applied to final grinding trials on actual blade and vane (profiled) production components. Final grinding test results of four specific production operations showed that at least a 27% improvement in wheel life could be realized, relative to the levels previously established in production, by optimizing grinding fluid application. In addition, a set of guidelines for optimized coolant nozzle and coolant system design and manufacture have been developed from the results of this research, and are applicable to other types of grinding or machining as well.

  2. 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.

  3. 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...

  4. Glycol coolants improve heat transfer and corrosion control

    SciTech Connect

    Holfield, R.

    1995-03-01

    Various liquids from plain water to exotic fluids have been used as coolants in large stationary diesel engines that drive compressors on natural gas pipeline distribution systems. Although water is an efficient heat transfer medium, its drawbacks of freezing at {minus}32 F and boiling at 212 F seriously limit its usefulness. Special glycol-based heat transfer fluids are available and refined specifically for long-term needs of gas compressor engines. Appropriate corrosion inhibitors have been formulated for metallurgy and operating conditions encountered with these engines. Propylene glycol was developed as an alternative for use in environmentally sensitive areas. Glycol-based fluids must be specifically inhibited for industrial applications because uninhibited or improperly inhibited coolants can seriously damage reciprocating engines.

  5. Thermal stratification potential in rocket engine coolant channels

    NASA Technical Reports Server (NTRS)

    Kacynski, Kenneth J.

    1992-01-01

    The potential for rocket engine coolant channel flow stratification was computationally studied. A conjugate, 3-D, conduction/advection analysis code (SINDA/FLUINT) was used. Core fluid temperatures were predicted to vary by over 360 K across the coolant channel, at the throat section, indicating that the conventional assumption of a fully mixed fluid may be extremely inaccurate. Because of the thermal stratification of the fluid, the walls exposed to the rocket engine exhaust gases will be hotter than an assumption of full mixing would imply. In this analysis, wall temperatures were 160 K hotter in the turbulent mixing case than in the full mixing case. The discrepancy between the full mixing and turbulent mixing analyses increased with increasing heat transfer. Both analysis methods predicted identical channel resistances at the coolant inlet, but in the stratified analysis the thermal resistance was negligible. The implications are significant. Neglect of thermal stratification could lead to underpredictions in nozzle wall temperatures. Even worse, testing at subscale conditions may be inadequate for modeling conditions that would exist in a full scale engine.

  6. 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.

  7. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    SciTech Connect

    Cahalan, J.; Wigeland, R. ); Friedel, G. , Bergisch Gladbach ); Kussmaul, G.; Royl, P. ); Moreau, J. ); Perks, M.

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs.

  8. 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.

  9. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 3 2012-10-01 2012-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 must, as soon...

  10. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 3 2011-10-01 2011-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 must, as soon...

  11. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 3 2014-10-01 2014-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 must, as soon...

  12. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 shall as soon...

  13. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 3 2013-10-01 2013-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 must, as soon...

  14. 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

  15. 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

  16. 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.

  17. Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities; Yucca Mountain Site Characterization Project

    SciTech Connect

    Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J.; Laub, T.W.

    1992-06-01

    This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10{sup {minus}11}/yr to 10{sup {minus}5}/yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10{sup {minus}9}/yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution.

  18. Experience in Organization of Urgent Medical Care in Large-Scale Accident Conditions at Nuclear Power Stations

    DTIC Science & Technology

    2001-05-01

    approximately 340 Gwt) which is 17% of the worldwide electricity production level. The Chernobyl Nuclear Disaster with its long-term medico-biological...population to avoid or minimize the risk of their exposure. In October 1986, after the Chernobyl Nuclear Disaster the Federal Center of Radiation...period of their cooperation the Urgent Medical Care Department and the Rapid Response Teams visited the following sites of accidents: "* Chernobyl (a fire

  19. Bicycle accidents in childhood.

    PubMed

    Nixon, J; Clacher, R; Pearn, J; Corcoran, A

    1987-05-16

    The results of a 10 year study of bicycle fatalities and an eight year study of serious non-fatalities are reported for urban Brisbane (population 1,000,000). There were 845 serious non-fatal bicycle accidents and 46 fatalities during the study. Boys were involved in 86% of accidents. Boys have an accident rate of 134.21 per 100,000 population at risk and a fatality rate of 5.06 per 100,000 at risk. Serious bicycle accidents have increased by 50% in this decade; but considering fatal cases alone, no secular trend was evident over the 10 year period of the study. This suggests that an increase in the overall rate of bicycle accidents has been in part compensated by less serious injuries. In 70% of fatalities children had head injuries, and 87% of fatalities followed a collision between a cyclist and a motor vehicle or a train. Bicycle accidents on the roads most commonly occur to boys aged between 12 and 14 years on a straight road at "mid-block" between 3 and 5 pm in clear weather conditions and in daylight. It is concluded that injuries and fatalities after bicycle accidents can be reduced by protecting children's heads, separating child cyclists from other road traffic, or educating and training both cyclists and other road users in safe behaviour. The compulsory use of helmets and the restriction of access to the roads by child cyclists to reduce injuries are, however, still controversial in many areas.

  20. 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.

  1. 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.

  2. Detailed heat transfer coefficient distributions under an array of impinging jets with coolant extraction

    SciTech Connect

    Huang, Y.; Ekkad, S.V.; Han, J.C.

    1996-12-31

    Jet impingement cooling is a high performance technique for heat transfer enhancement. Local heat transfer distributions are presented for an array of jets impinging on a target plate with a series of coolant extraction holes. The flow enters the pressure channel, impinges on the target plate and exits toward the sides and through the coolant extraction holes. The impingement plate has four rows of 12 jet holes and the target plate has three rows of 11 coolant extraction holes. The jet holes and the coolant extraction holes have the same diameters and are staggered such that the air impinging from the jet hole does not exit directly through the extraction hole. The detailed heat transfer coefficient distributions are measured using a transient technique and liquid crystal coating. Results are presented for a range of jet Reynolds numbers between 4,000 and 20,000. The effect of crossflow is also studied by changing the exit opening of the impingement channel to provide three different spent air exit directions. Heat transfer results for the target plate with coolant extraction are compared with those without coolant extraction at the same flow conditions.

  3. 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

  4. 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

  5. Lessons on in-vessel severe accidents from experiments at KfK and the INEL

    SciTech Connect

    Hagen, S.; Hofmann, P. ); Allison, C. )

    1993-01-01

    At the time of the Three Mile Island unit 2 accident, the severe fuel damage data base was very limited. However, as a result of international severe accident research programs, that data base has been greatly expanded to include experiments ranging from single rod tests in the NIELS facility at the Kernforschungszentrum, Karlsruhe, Germany (KfK), to the coupled reactor coolant system (RCS) thermal-hydraulics-assembly melting loss-of-fluid test (LOFT) LP-FP-2 test in the LOFT reactor at the Idaho National Engineering Laboratory (INEL). In addition to these tests at the extreme ends of the scaling spectrum, other experiments in the CORA facility at the KfK and the Power Burst Facility at the INEL have also made a substantial contribution to the severe fuel damage data base. The experiments performed at the KfK and the INEL have included different modes of heating from the electrically heated experiments in Germany to the fission and decay-heat-driven experiments in Idaho. The tests have included (a) different heating rates, peak bundle temperatures, and coolant conditions, (b) fresh, trace-irradiated, and previously irradiated (36,000 MWd/tonne U) fuel, and (c) different bundle designs, including fuel-only bundles, fuel bundles with Ag-In-Cd control rods, and fuel bundles with representative B[sub 4]C control blade/channel boxes. This paper discusses several key results from these experiments that are common across all the facilities.

  6. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... your engine in use. (b) For laboratory testing of liquid-cooled engines, you may use water with or... 40 Protection of Environment 33 2014-07-01 2014-07-01 false Coolants. 1065.745 Section 1065.745 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING...

  7. 40 CFR 1065.745 - Coolants.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... your engine in use. (b) For laboratory testing of liquid-cooled engines, you may use water with or... 40 Protection of Environment 34 2012-07-01 2012-07-01 false Coolants. 1065.745 Section 1065.745 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING...

  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. Reactor coolant pump shaft seal stability during station blackout

    SciTech Connect

    Rhodes, D B; Hill, R C; Wensel, R G

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

  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. Application of NUREG-1150 methods and results to accident management

    SciTech Connect

    Dingman, S.; Sype, T.; Camp, A.; Maloney, K.

    1990-01-01

    The use of NUREG-1150 and similar Probabilistic Risk Assessments in NRC and industry risk management programs is discussed. Risk management'' is more comprehensive than the commonly used term accident management.'' Accident management includes strategies to prevent vessel breach, mitigate radionuclide releases from the reactor coolant system, and mitigate radionuclide releases to the environment. Risk management also addresses prevention of accident initiators, prevention of core damage, and implementation of effective emergency response procedures. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. Examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management are discussed, with particular attention given to the early phases of accidents. Plans and methods for evaluating accident management strategies that have been identified in the NRC accident management program are discussed. 2 refs., 3 figs.

  12. Nuclear accidents

    SciTech Connect

    Mobley, J.A.

    1982-05-01

    A nuclear accident with radioactive contamination can happen anywhere in the world. Because expert nuclear emergency teams may take several hours to arrive at the scene, local authorities must have a plan of action for the hours immediately following an accident. The site should be left untouched except to remove casualties. Treatment of victims includes decontamination and meticulous wound debridement. Acute radiation syndrome may be an overwhelming sequela.

  13. Comparative analysis of pressure vessel integrity for various LOCA conditions

    NASA Astrophysics Data System (ADS)

    Çolak, Üner; Özdere, Oya

    2001-09-01

    In this study, integrity analysis is performed for a classical four loop PWR pressure vessel fabricated from SA533B type ferritic steel. Pressure vessel behavior is analyzed by deterministic and probabilistic methods under transient conditions, which may cause pressurized thermal shock (PTS). In deterministic analysis, the change of material properties and the mechanical state of the vessel are analyzed against changes in coolant pressure and temperature. Probabilistic analysis is performed to obtain pressure vessel beltline region weld failure probabilities in transient conditions. Overall vessel failure probabilities are evaluated based on the results of deterministic analyses. Computer code VISA-II is utilized for the calculation of vessel failure probabilities. Among three cases considered in this study, a medium break loss of coolant accident induced by a 50 cm2 break in the hot leg yields the highest vessel rupture probability. The maximum nil ductility temperature in all cases is still below the NRC PTS limit.

  14. 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.

  15. 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.

  16. [Genomic instability after exposure to radiation at low doses (in the 10-kilometer zone of the accident at the Chernobyl Atomic Electric Power Station and under laboratory conditions)].

    PubMed

    Pelevina, I I; Gotlib, V Ia; Kudriashova, O V; Serebrianyĭ, A M; Afanas'ev, G G

    1996-01-01

    The results of series investigations of late effects after Chernobyl accident are discussed. Genomic instability induced by chronic irradiation of cultural cells in Chernobyl zone and in laboratory conditions have been studied. It was shown that low level prolonged irradiation result in increase of frequency of cells with micronuclei, giant cells, enhancement of radiosensitivity in descendents of early irradiated cells. Chronic low doses irradiation doesn't induce the adaptive response. Comparative investigation of adaptive response in blood lymphocytes of people (adults and children) living in Moscow and in regions polluted with radionuclides (5-40 ci/km2) after Chernobyl disaster have been conducted. In population from contaminated areas the frequency of individuals with definite adaptive response is decreased and there are individuals with increasing radiosensitivity after irradiation in conditioned dose. Chronic irradiation during living on contaminated areas don't induce the adaptive response.

  17. 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.

  18. Bicycle accidents in childhood.

    PubMed Central

    Nixon, J; Clacher, R; Pearn, J; Corcoran, A

    1987-01-01

    The results of a 10 year study of bicycle fatalities and an eight year study of serious non-fatalities are reported for urban Brisbane (population 1,000,000). There were 845 serious non-fatal bicycle accidents and 46 fatalities during the study. Boys were involved in 86% of accidents. Boys have an accident rate of 134.21 per 100,000 population at risk and a fatality rate of 5.06 per 100,000 at risk. Serious bicycle accidents have increased by 50% in this decade; but considering fatal cases alone, no secular trend was evident over the 10 year period of the study. This suggests that an increase in the overall rate of bicycle accidents has been in part compensated by less serious injuries. In 70% of fatalities children had head injuries, and 87% of fatalities followed a collision between a cyclist and a motor vehicle or a train. Bicycle accidents on the roads most commonly occur to boys aged between 12 and 14 years on a straight road at "mid-block" between 3 and 5 pm in clear weather conditions and in daylight. It is concluded that injuries and fatalities after bicycle accidents can be reduced by protecting children's heads, separating child cyclists from other road traffic, or educating and training both cyclists and other road users in safe behaviour. The compulsory use of helmets and the restriction of access to the roads by child cyclists to reduce injuries are, however, still controversial in many areas. PMID:3109611

  19. 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.

  20. [Travel and accidents].

    PubMed

    Cha, Olivier

    2015-04-01

    Traumatic pathologies are the most frequent medical events to be observed among French travellers. Accidents on the public highway by lack of respect of the fundamental rules of road security, particularly abroad, traffic conditions in bad repair in numerous emergent countries, usually the destination of mass tourism and underdeveloped organization of health care and local urgency help. Sports activities are also a source of accidents. A good physical training is essential. Drowning is a real plague, especially among children due to a lack of vigilance. Preventive measures are simple, keep them constantly in mind and apply them carefully so as to have beautiful memories of our trip back home.

  1. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 3 2011-10-01 2011-10-01 false Reporting accidents. 195.50 Section 195.50 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An...

  2. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 3 2012-10-01 2012-10-01 false Reporting accidents. 195.50 Section 195.50 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An...

  3. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 3 2014-10-01 2014-10-01 false Reporting accidents. 195.50 Section 195.50 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An...

  4. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Reporting accidents. 195.50 Section 195.50 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An...

  5. Rhizopus-associated soft tissue infection in an immunocompetent air-conditioning technician after a road traffic accident: a case report and review of the literature.

    PubMed

    Rabie, Nada B; Althaqafi, Abdulhakeem O

    2012-03-01

    Rhinocerebral or sinopulmonary mucromycosis is a well-recognized human fungal infection found among immunocompromised and diabetic patients. However, the infection is rare among immunocompetent hosts. We are reporting the case of an adult immunocompetent male patient working as an air-conditioning technician. The patient was a victim of a road traffic accident (RTA) and sustained multiple fractures in the proximal part of the left tibia, distal femur, and scapula. Two weeks postoperatively, Rhizopus microspores were isolated from an infected traumatic wound over the distal femur. Surgical debridement was performed, and the patient was started on amphotericin B. Occupational exposure history and workplace environmental sanitation are crucial for the prevention of this potentially fatal yet preventable infection.

  6. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    SciTech Connect

    Myers, B.F.; Morrissey, R.E.

    1980-04-01

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO/sub 2/ and highly enriched (HEU) TRISO UC/sub 2/ particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO/sub 2/ particles and 23.5 and 74% FIMA for UC/sub 2/ particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium.

  7. 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.

  8. 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.

  9. Coupled reactor physics and coolant dynamics of heavy liquid metal coolant systems.

    SciTech Connect

    Cahalan, J. E.; Dunn, F. E.; Taiwo, T. A.

    1999-07-15

    Cooling of advanced nuclear designs with heavy liquid metals such as lead or lead-bismuth eutectic offers the potential for plant simplifications and higher operating efficiencies compared to previously considered liquid metal coolants such as sodium or NaK. Such applications would however also introduce additional safety concerns and design challenges, therefore necessitating a verifiable computational tool for transient design-basis analysis of heavy liquid metal coolant (HLMC) systems. This capability would enable analysts to compare operational and safety characteristics of design alternatives, and to evaluate relative performance advantages with a consistent, deterministic measure.

  10. Estimation of thermal loads on the VVER vessel under conditions of inversion of the stratified molten pool in a severe accident

    NASA Astrophysics Data System (ADS)

    Loktionov, V. D.; Mukhtarov, E. S.

    2016-09-01

    Analysis of the thermal state of molten pools that can be formed on the vessel bottom of the VVER-600 medium-power reactor during a severe anticipated accident with melting of the core is represented. Two types of the molten pool of core materials, with the two-layer and inverse three-layer stratification, are considered. Thermal loads acting on the reactor vessel from the melt are estimated depending on its formation time. Features of the thermal state of the melt in the case of its inverse stratification are analyzed. It is shown that thermal loads on the reactor vessel exceed the critical heat flux (CHF) when forming the two-layer stratified molten pool 10 and 24 h after its shutdown, and the thermal load is close to the corresponding CHF or somewhat exceeds it in 72 h. In the case of the formation of the inverse structure of the melt, one can observe a decrease by more than 2.5 times (in comparison with the two-layer stratified structure) in the thermal load on the reactor vessel in the region of its contact with the upper layer of the steel melt. Analysis of results showed that maximum densities of heat flux to the reactor vessel from the bottom metallic layer with the melt inversion did not exceed corresponding CHFs 24 and 72 h after the reactor shutdown. Because the thermal load on the reactor vessel can be localized in the region of its bottom, where the CHF is relatively small, during the inverse stratification of the melt, there is a need to carry out further in-depth experimental and analytical investigations of conditions for formation of the stratified molten pool and to obtain corrected experimental CHFs for conditions and outlines of cooling the external surface of the VVER-600 vessel in a severe accident.

  11. Weather types and traffic accidents.

    PubMed

    Klaić, Z B

    2001-06-01

    Traffic accident data for the Zagreb area for the 1981-1982 period were analyzed to investigate possible relationships between the daily number of accidents and the weather conditions that occurred for the 5 consecutive days, starting two days before the particular day. In the statistical analysis of low accident days weather type classification developed by Poje was used. For the high accident days a detailed analyses of surface and radiosonde data were performed in order to identify possible front passages. A test for independence by contingency table confirmed that conditional probability of the day with small number of accidents is the highest, provided that one day after it "N" or "NW" weather types occur, while it is the smallest for "N1" and "Bc" types. For the remaining 4 days of the examined periods dependence was not statistically confirmed. However, northern ("N", "NE" and "NW") and anticyclonic ("Vc", "V4", "V3", "V2" and "mv") weather types predominated during 5-days intervals related to the days with small number of accidents. On the contrary, the weather types with cyclonic characteristics ("N1", "N2", "N3", "Bc", "Dol1" and "Dol"), that are generally accompanied by fronts, were the rarest. For 85% days with large number of accidents, which had not been caused by objective circumstances (such as poor visibility, damaged or slippery road etc.), at least one front passage was recorded during the 3-days period, starting one day before the day with large number of accidents.

  12. Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments

    DOE PAGES

    Parish, Chad M.; Terrani, Kurt A.; Kim, Young -Jin; ...

    2016-11-28

    Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corrodedmore » region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA – which suffered the least corrosion –retained some Al in the corroded region. Lastly, the results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment.« less

  13. Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments

    SciTech Connect

    Parish, Chad M.; Terrani, Kurt A.; Kim, Young -Jin; Koyanagi, Takaaki; Katoh, Yutai

    2016-11-28

    Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corroded region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA – which suffered the least corrosion –retained some Al in the corroded region. Lastly, the results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment.

  14. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    SciTech Connect

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.

  15. Factors affecting vertical distribution of Fukushima accident-derived radiocesium in soil under different land-use conditions.

    PubMed

    Koarashi, Jun; Atarashi-Andoh, Mariko; Matsunaga, Takeshi; Sato, Tsutomu; Nagao, Seiya; Nagai, Haruyasu

    2012-08-01

    The Fukushima Dai-ichi nuclear power plant accident in Japan, triggered by a big earthquake and the resulting tsunami on 11 March 2011, caused a substantial release of radiocesium ((137)Cs and (134)Cs) and a subsequent contamination of soils in a range of terrestrial ecosystems. Identifying factors and processes affecting radiocesium retention in these soils is essential to predict how the deposited radiocesium will migrate through the soil profile and to other biological components. We investigated vertical distributions of radiocesium and physicochemical properties in soils (to 20 cm depth) at 15 locations under different land-use types (croplands, grasslands, and forests) within a 2 km × 2 km mesh area in Fukushima city. The total (137)Cs inventory deposited onto and into soil was similar (58.4±9.6 kBq m(-2)) between the three different land-use types. However, aboveground litter layer at the forest sites and herbaceous vegetation at the non-forested sites contributed differently to the total (137)Cs inventory. At the forest sites, 50-91% of the total inventory was observed in the litter layer. The aboveground vegetation contribution was in contrast smaller (<35%) at the other sites. Another remarkable difference was found in vertical distribution of (137)Cs in mineral soil layers; (137)Cs penetrated deeper in the forest soil profiles than in the non-forested soil profiles. We quantified (137)Cs retention at surface soil layers, and showed that higher (137)Cs retention can be explained in part by larger amounts of silt- and clay-sized particles in the layers. More importantly, the (137)Cs retention highly and negatively correlated with soil organic carbon content divided by clay content across all land-use types. The results suggest that organic matter inhibits strong adsorption of (137)Cs on clay minerals in surface soil layers, and as a result affects the vertical distribution and thus the mobility of (137)Cs in soil, particularly in the forest ecosystems.

  16. Cable condition monitoring research activities at Sandia National Laboratories

    SciTech Connect

    Jacobus, M.J.; Zigler, G.L.; Bustard, L.D.

    1988-01-01

    Sandia National Laboratories is currently conducting long-term aging research on representative samples of nuclear power plant cables. The objectives of the program are to determine the suitability of these cables for extended life (beyond 40 year design basis) and to assess various cable condition monitoring techniques for predicting remaining cable life. The cables are being aged for long times at relatively mild exposure conditions with various condition monitoring techniques to be employed during the aging process. Following the aging process, the cables will be exposed to a sequential accident profile consisting of high dose rate irradiation followed by a simulated design basis loss-of-coolant accident (LOCA) steam exposure. 12 refs., 1 fig., 1 tab.

  17. TACT 1: A computer program for the transient thermal analysis of a cooled turbine blade or vane equipped with a coolant insert. 2. Programmers manual

    NASA Technical Reports Server (NTRS)

    Gaugler, R. E.

    1979-01-01

    A computer program to calculate transient and steady state temperatures, pressures, and coolant flows in a cooled axial flow turbine blade or vane with an impingement insert is described. Coolant-side heat transfer coefficients are calculated internally in the program, with the user specifying either impingement or convection heat transfer at each internal flow station. Spent impingement air flows in a chordwise direction and is discharged through the trailing edge and through film cooling holes. The ability of the program to handle film cooling is limited by the internal flow model. Input to the program includes a description of the blade geometry, coolant-supply conditions, outside thermal boundary conditions, and wheel speed. The blade wall can have two layers of different materials, such as a ceramic thermal barrier coating over a metallic substrate. Program output includes the temperature at each node, the coolant pressures and flow rates, and the coolant-side heat transfer coefficients.

  18. Cryogenic-coolant He-4-superconductor interaction

    NASA Technical Reports Server (NTRS)

    Caspi, S.; Lee, J. Y.; Kim, Y. I.; Allen, R. J.; Frederking, T. H. K.

    1978-01-01

    The thermodynamic and thermal interaction between a type 2 composite alloy and cryo-coolant He4 was studied with emphasis on post quench phenomena of formvar coated conductors. The latter were investigated using a heater simulation technique. Overall heat transfer coefficients were evaluated for the quench onset point. Heat flux densities were determined for phenomena of thermal switching between a peak and a recovery value. The study covered near saturated liquid, pressurized He4, both above and below the lambda transition, and above and below the thermodynamic critical pressure. In addition, friction coefficients for relative motion between formvar insulated conductors were determined.

  19. 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.

  20. 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.

  1. Phased Development of Accident Tolerant Fue

    SciTech Connect

    Bragg-Sitton, Shannon M.; Carmack, W. Jon

    2016-09-01

    The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modeling predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology

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

    SciTech Connect

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

    2012-07-01

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

  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. 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.

  5. Scoping assessments of ATF impact on late–stage accident progression including molten core-concrete interaction

    SciTech Connect

    Farmer, Mitchell T.; Leibowitz, Leonard; Terrani, Kurt A.; Robb, Kevin R.

    2013-12-31

    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. Furthermore, experimental data for this system under severe accident conditions is needed for a proper evaluation and comparison to iron-based claddings.

  6. 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.

  7. Thermal state of the safety system, reactor, side reflector and shielding of the {open_quote}{open_quote}TOPAZ-2{close_quote}{close_quote} system under conditions of fire caused by a launcher accident at the launch pad

    SciTech Connect

    Grinberg, E.I.; Doschatov, V.V.; Nikolaev, V.S.; Sokolov, N.S.; Usov, V.A.

    1996-03-01

    The paper presents some results of calculational analyses performed to determine thermal state of the TOPAZ II safety system structure, radiation shielding, reactor without the side reflector, rods and inserts of the side reflector under conditions of fire at the launch pad when an accident occurs to a launch vehicle. {copyright} {ital 1996 American Institute of Physics.}

  8. Numerical Analysis of Nucleate Boiling on High Heat-Flux and High Subcooling Condition for Reactivity Initiation Accident

    SciTech Connect

    Heo, S.; Koshizuka, S.; Oka, Y.

    2002-07-01

    This paper shows the numerical simulation study on the growth of the bubble in the transient pool boiling using MPS-MAFL method. The growth process of a bubble with the different initial radii is calculated in a high heat-flux and high subcooling condition expected in nuclear reactor core during RIA. The smaller initial radius is, the earlier the growth starts. The initial bubble radius has little effect on the growth initiation time and the bubble departure radius. (authors)

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

    SciTech Connect

    Hutter, Ernest

    1986-01-01

    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.

  10. 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

  11. 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.

  12. Hang-gliding accidents.

    PubMed Central

    Margreiter, R; Lugger, L J

    1978-01-01

    Seventy-five known hang-gliding accidents causing injury to the pilot occurred in the Tyrol during 1973-6. Most occurred in May, June, or September and between 11 am and 3 pm, when unfavourable thermic conditions are most likely. Thirty-four accidents happened during launching, 13 during flight, and 28 during landing, and most were caused by human errors--especially deficient launching technique; incorrect estimation of wind conditions, altitude, and speed; and choice of unfavourable launching and landing sites. Eight pilots were moderately injured, 60 severely (multiply in 24 cases), and seven fatally; fractures of the spine and arms predominated. Six of the 21 skull injuries were fatal. The risk of hang-gliding seems unjustifiably high, and safety precautions and regulations should be adopted to ensure certain standards of training and equipment and to limit flying to favourable sites and times. Images p401-a PMID:624028

  13. 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.

  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. 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.

  16. The basic problems of bed-fence-covers in hospitals for preventing accidents based on the investigation into the actual conditions: for developing the safer bed-fence-cover for elderly patients.

    PubMed

    Matsuoka, Megumi; Konishi, Teuko; Toyoda, Mitsuko; Maie, Kazuo

    2009-12-01

    The basic problems of bed-fence-covers in hospitals were listed for preventing relevant accidents based on the investigation into actual conditions in a hospital in Kawasaki City. There were many elderly aged patients with dementia, higher brain dysfunction or psychosis in the hospital. They sometimes fell into the gaps of bed-fences, resulting in serious accidents. It was due not only to the structure of the bed-fences, but also to the characteristics of patients. Therefore the authors listed up the problems concerning the accidents to recognize them; (i) as physical conditions, (1) they could not move by themselves because of paralysis or decrease of fitness, (2) they could not feel when they were pinched by the gaps because of decrease of sense, (3) they moved irregularly or violently without their intention, and (ii) as mental conditions, (1) they took dangerous behaviors because of dementias, (2) they could not control their behaviors because of higher brain dysfunctions, (3) they could not control their feelings and moved violently because of mental disorders. The authors intend to develop safer bed-fence-covers to prevent these accidents for the elderly patients.

  17. 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.

  18. 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.

  19. 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.

  20. Impact of boron dilution accidents on low boron PWR safety

    SciTech Connect

    Papukchiev, A.; Liu, Y.; Schaefer, A.

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As an inadvertent reduction of the boron concentration during a boron dilution accident could introduce positive reactivity and have a negative impact on PWR safety, design changes to reduce boron concentration in the reactor coolant are of general interest. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) load has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) to 518 ppm. For the assessment of the potential safety advantages, a boron dilution accident due to small break loss-of-coolant-accident (SBLOCA) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The results from the comparative analyses showed that the impact of the boron dilution accident on the new PWR design safety is significantly lower in comparison with the standard design. The new reactor design provided at least 4, 4% higher reactivity margin to recriticality during the whole accident which is equivalent to the negative reactivity worth of additional 63% of all control rods fully inserted in to the core. (authors)

  1. Effects of indoor and outdoor cultivation conditions on (137) Cs concentrations in cultivated mushrooms produced after the Fukushima Daiichi Nuclear Power Plant accident.

    PubMed

    Tagami, Keiko; Uchida, Shigeo; Ishii, Nobuyoshi

    2017-01-01

    Radiocesium ((134) Cs and (137) Cs) in mushrooms has been a matter of public concern after the accident at Fukushima Daiichi Nuclear Power Plant. To minimize the internal dose by ingestion of cultivated mushrooms, the Japanese government set a guideline level with respect to the radiocesium concentration in bed-logs and mushroom beds; however, the effects of indoor and outdoor cultivation methods on radiocesium concentrations in cultivated mushrooms were not clear. The effects of indoor and outdoor cultivation on the radiocesium concentrations in mushroom were examined using published food monitoring data. (137) Cs concentration data in Lentinula edodes from the Aizu area in Fukushima Prefecture and seven prefectures outside Fukushima were used for the analysis. No statistically significant (137) Cs concentration differences were found between these two cultivation methods. Using detected (137) Cs data in shiitake, the geometric means from each prefecture were less than one-quarter of the standard limit (100 Bq kg(-1) ) for total radiocesium under both cultivation conditions. It was suspected that re-suspended radiocesium might have been taken up by mushrooms or that radiocesium might have been absorbed into the mushrooms from the soil in the outdoor cultures. However, neither effect was significant for cultivated mushrooms in the areas examined. © 2016 Society of Chemical Industry. © 2016 Society of Chemical Industry.

  2. Use of accident experience in developing criteria for teleoperator equipment

    SciTech Connect

    Vallario, E.J.; Selby, J.M.

    1985-10-01

    The 1961 SL-1 reactor accident in Idaho and the Recuplex accident at Hanford are reviewed to identify problems common to emergency situations, lessons learned from accidents, criteria for emergency equipment, and recommendations for using robotics to solve problems during emergencies. Teleoperator equipment could be used to assess the extent of the damage and the condition of the reactor, retrieve dosimeters, evacuate and treat accident victims, clean up debris and decontaminate accident areas. 2 refs., 9 figs.

  3. 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.

  4. Reclamation and disposal of water-based machining coolants

    SciTech Connect

    Taylor, P.A.

    1982-01-01

    The Oak Ridge Y-12 Plant, which is operated by the Union Carbide Corporation, Nuclear Division for the Department of Energy under US government contract W-7405-eng-26, currently uses about 10{sup 6} L/yr (260,000 gal/yr) of water-based coolants in its machining operations. These coolants are disposed of in a 110,000-L (29,000-gal) activated sludge reactor. The reactor has oxidized an average of 38.6 kg of total organic carbon (TOC) per day with an overall efficiency of 90%. The predominant bacteria in the reactor have been identified once each year for the past three years. Six primary types of water-based coolants are currently used in the machine shops. In order to reduce the coolant usage rate, efforts are being made to introduce one universal coolant into the shops. By using a biocide to limit bacterial deterioration and using a filter and centrifuge system to remove dirt and tramp oils from the coolant, the coolant discard rate can be greatly reduced. 1 tab.

  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. 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

  7. Turbomachine injection nozzle including a coolant delivery system

    DOEpatents

    Zuo, Baifang [Simpsonville, SC

    2012-02-14

    An injection nozzle for a turbomachine includes a main body having a first end portion that extends to a second end portion defining an exterior wall having an outer surface. A plurality of fluid delivery tubes extend through the main body. Each of the plurality of fluid delivery tubes includes a first fluid inlet for receiving a first fluid, a second fluid inlet for receiving a second fluid and an outlet. The injection nozzle further includes a coolant delivery system arranged within the main body. The coolant delivery system guides a coolant along at least one of a portion of the exterior wall and around the plurality of fluid delivery tubes.

  8. 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.

  9. Longer life for glyco-based stationary engine coolants

    SciTech Connect

    Hohlfeld, R.

    1996-07-01

    Large, stationary diesel engines used to compress natural gas that is to be transported down pipelines generate a great deal of heat. Unless this heat is dissipated efficiently, it will eventually cause an expensive breakdown. Whether the coolant uses ethylene glycol or propylene glycol, the two major causes of glycol degradation are heat and oxidation. The paper discusses inhibitors that enhance coolant service life and presents a comprehensive list of do`s and don`ts for users to gain a 20-year coolant life.

  10. 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.

  11. 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.

  12. Evaluation of secondary coolant control design alternatives and their effects on heat removal performance

    SciTech Connect

    Khayat, M.I.; Anderson, J.; Battle, R.; March-Leuba, J.

    1994-03-01

    This report documents a series of calculations that evaluate the performance of the core-inlet temperature controller under different transient conditions and design options. The present analyses show that the core-inlet temperature can be controlled at {approx}45{degrees}C under all transient conditions analyzed using the controller design described in the conceptual design report, which includes variable-speed secondary coolant pumps and variable-speed cooling tower fans. This study also shows that a constant-speed secondary pump would be sufficient to maintain core-inlet temperature <45{degrees}C if this temperature is allowed to drop below the set point during some demanding transients, such as normal startup. The use of secondary loop hot coolant to warm the reactor building was also evaluated; however, optimization of the secondary hot-leg temperature can only be achieved by trading off control of the primary side core-inlet temperature.

  13. 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.

  14. Pressurization of a compartment due to the rupture of coolant piping

    SciTech Connect

    Kot, C.A.; Hsieh, B.J.

    1993-01-01

    The pressurization and venting of enclosed compartments due to the accidental rupture of coolant piping is a safety problem common to many nuclear facilities. The processes associated with such an accident are very complex, involving, in general, transient multiphase flows, interactions and mixing between the incoming flows and the gases in the compartment, and heat transfer with the surroundings. Since pipe rupture is associated with many phenomenological uncertainties, elaborate 3-D thermal-hydraulic modeling and extensive calculational efforts are not warranted for many design applications. It is then more appropriate to rely. on simplified, global analysis approaches which can provide reasonably conservative estimates of the structural loads and flow processes, and which can readily be used in parameter/design studies. The objective of this paper is to present such an approach.

  15. 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.

  16. Underreporting of maritime accidents to vessel accident databases.

    PubMed

    Hassel, Martin; Asbjørnslett, Bjørn Egil; Hole, Lars Petter

    2011-11-01

    Underreporting of maritime accidents is a problem not only for authorities trying to improve maritime safety through legislation, but also to risk management companies and other entities using maritime casualty statistics in risk and accident analysis. This study collected and compared casualty data from 01.01.2005 to 31.12.2009, from IHS Fairplay and the maritime authorities from a set of nations. The data was compared to find common records, and estimation of the true number of occurred accidents was performed using conditional probability given positive dependency between data sources, several variations of the capture-recapture method, calculation of best case scenario assuming perfect reporting, and scaling up a subset of casualty information from a marine insurance statistics database. The estimated upper limit reporting performance for the selected flag states ranged from 14% to 74%, while the corresponding estimated coverage of IHS Fairplay ranges from 4% to 62%. On average the study results document that the number of unreported accidents makes up roughly 50% of all occurred accidents. Even in a best case scenario, only a few flag states come close to perfect reporting (94%). The considerable scope of underreporting uncovered in the study, indicates that users of statistical vessel accident data should assume a certain degree of underreporting, and adjust their analyses accordingly. Whether to use correction factors, a safety margin, or rely on expert judgment, should be decided on a case by case basis.

  17. 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.

  18. Correlation of cylinder-head temperatures and coolant heat rejections of a multicylinder, liquid-cooled engine of 1710-cubic-inch displacement

    NASA Technical Reports Server (NTRS)

    Lundin, Bruce T; Povolny, John H; Chelko, Louis J

    1949-01-01

    Data obtained from an extensive investigation of the cooling characteristics of four multicylinder, liquid-cooled engines have been analyzed and a correlation of both the cylinder-head temperatures and the coolant heat rejections with the primary engine and coolant variables was obtained. The method of correlation was previously developed by the NACA from an analysis of the cooling processes involved in a liquid-cooled-engine cylinder and is based on the theory of nonboiling, forced-convection heat transfer. The data correlated included engine power outputs from 275 to 1860 brake horsepower; coolant flows from 50 to 320 gallons per minute; coolants varying in composition from 100 percent water to 97 percent ethylene glycol and 3 percent water; and ranges of engine speed, manifold pressure, carburetor-air temperature, fuel-air ratio, exhaust-gas pressure, ignition timing, and coolant temperature. The effect on engine cooling of scale formation on the coolant passages of the engine and of boiling of the coolant under various operating conditions is also discussed.

  19. 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...

  20. 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...

  1. Chemical Characterization of Simulated Boiling Water Reactor Coolant

    DTIC Science & Technology

    1990-05-01

    industry to reduce personnel radiation exposure and down-time associated with the operation, mainte- nance and refueling of Light Water Reactor (LWR...AD-A226 654 t t-FILL UIY C CHEMICAL CHARACTERIZATION OF SIMULATED , .BOILING WATER REACTOR COOLANt by Li . . , . , - VERRDON HOLBROOK MASON f ; B.S...CHARACTERIZATION OF SIMULATED BOILING WATER REACTOR COOLANT by VERRDON HOLBROOK MASON Submitted to the Department of Nuclear Engineering on May 9, 1988 in

  2. Investigation of cleaner technologies to minimize automotive coolant wastes

    SciTech Connect

    Randall, P.M.

    1993-01-01

    The U.S. 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 economic issues. In addition, the authors examined the potential for substituting propylene glycol for ethylene glycol based engine coolant formulations. (Copyright (c) 1993 Butterworth-Heinemann Ltd.)

  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. Coupled reactor physics and coolant dynamics of heavy-liquid-metal-coolant systems

    SciTech Connect

    Cahalan, J.E.; Dunn, F.E.; Taiwo, T.A.

    1999-07-01

    Cooling of advanced nuclear designs with heavy liquid metals such as lead or lead-bismuth eutectic offers improved safety characteristics and higher operating efficiencies compared with previously considered liquid-metal coolants such as sodium or NaK. Such applications would, however, also introduce additional safety concerns and design challenges, therefore necessitating a verifiable computational tool for transient design-basis analysis of heavy-liquid-metal-coolant (HLMC) systems. This capability would enable analysts to compare operational and safety characteristics of design alternatives and to evaluate relative performance advantages with a consistent, deterministic measure. No existing computer code provides all the computational capabilities needed for system analysis of HLMC designs. However, the SASSYS-1 computer code, long used for analysis of sodium-cooled designs, has been adapted to provide a scoping reactor kinetics/thermal-hydraulics analysis capability and a framework for future development. Revision of the SASSYS-1 reactor kinetics, thermophysical properties, convective heat transfer, and flow momentum transfer models has permitted preliminary design and safety assessments of HLMC systems. Preliminary results are given here.

  5. Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

    SciTech Connect

    Curtis, Franklin G; Ekici, Kivanc; Freels, James D

    2011-01-01

    The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

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

    NASA Astrophysics Data System (ADS)

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

    1989-08-01

    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.

  7. 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.

  8. 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.

  9. 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.

  10. Radiolysis of the coolant in the VK-50 boiling water reactor

    SciTech Connect

    Zabelin, A.I.; Shmelev, V.E.

    1986-10-01

    Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterized by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.

  11. Nuclear accident dosimetry intercomparison studies.

    PubMed

    Sims, C S

    1989-09-01

    Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shielded spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry.

  12. Severe accident analysis using dynamic accident progression event trees

    NASA Astrophysics Data System (ADS)

    Hakobyan, Aram P.

    In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. One of the principal deficiencies lies in the static nature of conventional APETs. In the conventional event tree techniques, the sequence of events is pre-determined in a fixed order based on the expert judgments. The main objective of this PhD dissertation was to develop a software tool (ADAPT) for automated APET generation using the concept of dynamic event trees. As implied by the name, in dynamic event trees the order and timing of events are determined by the progression of the accident. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. The function of a dynamic APET developed includes prediction of the conditions, timing, and location of containment failure or bypass leading to the release of radioactive material, and calculation of probabilities of those failures. Thus, scenarios that can potentially lead to early containment failure or bypass, such as through accident induced failure of steam generator tubes, are of particular interest. Also, the work is focused on treatment of uncertainties in severe accident phenomena such as creep rupture of major RCS components, hydrogen burn, containment failure, timing of power recovery, etc. Although the ADAPT methodology (Analysis of Dynamic Accident Progression Trees) could be applied to any severe accident analysis code, in this dissertation the approach is demonstrated by applying it to the MELCOR code [1]. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a

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

    SciTech Connect

    Ernst, Frank

    2016-10-13

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

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

    SciTech Connect

    Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.; Woolstenhulme, Nicolas E.; Ban, Heng; Wachs, Daniel M.

    2016-09-01

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

  15. 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.

  16. 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.

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

    NASA Technical Reports Server (NTRS)

    Stepka, Francis S

    1958-01-01

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

  18. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    SciTech Connect

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

    1997-07-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

  19. 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)

  20. 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.

  1. 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.

  2. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems

    SciTech Connect

    Titov, Gene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-04-05

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  3. 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.

  4. Full reactor coolant system chemical decontamination qualification programs

    SciTech Connect

    Miller, P.E.

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  5. 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.

  6. Diesel engine coolant analysis, new application for established instrumentation

    SciTech Connect

    Anderson, D.P.; Lukas, M.; Lynch, B.K.

    1998-09-01

    Rotating disk electrode (RDE) arc emission spectrometers are used in many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long, that most practitioners have probably forgotten that when RDE spectrometers were first introduced more than 40 years ago, they were also routinely used for aqueous samples. This paper describes recent work to calibrate and modify RDE arc emission spectrometers for the analysis of engine coolant samples; a mixture of approximately 50% water and 50% glycol. The technique has been shown to be effective for the analysis of wear metals, contamination and supplemental coolant additives in ethylene and propylene glycol. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and RDE spectrometers will be presented. The data correlates extremely well on new and relatively clean coolants. However, not surprisingly, RDE results are sometimes higher for samples containing particles larger than a few micrometers. This paper suggests that RDE spectrometers are appropriate, and sometimes preferred, for most types of coolants and certain types of aqueous samples. Actual field data is be presented to support the arguments.

  7. 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)

  8. Car accidents determined by stopped cars and traffic flow

    NASA Astrophysics Data System (ADS)

    Yang, Xian-qing; Ma, Yu-qiang

    2002-12-01

    The product of traffic flow and the fraction of stopped cars is proposed to determine the probability Pac for car accidents in the Fukui-Ishibashi model by analysing the necessary conditions of the occurrence of car accidents. Qualitative and quantitative characteristics of the probability Pac can well be explained. A strategy for avoiding car accidents is suggested.

  9. 10 CFR 50.67 - Accident source term.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The... to January 10, 1997, who seek to revise the current accident source term used in their design...

  10. 49 CFR 195.52 - Immediate notice of certain accidents.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 3 2014-10-01 2014-10-01 false Immediate notice of certain accidents. 195.52... TRANSPORTATION OF HAZARDOUS LIQUIDS BY PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.52 Immediate notice of certain accidents. (a) Notice requirements. At the earliest practicable moment...

  11. 49 CFR 195.52 - Immediate notice of certain accidents.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 3 2012-10-01 2012-10-01 false Immediate notice of certain accidents. 195.52... TRANSPORTATION OF HAZARDOUS LIQUIDS BY PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.52 Immediate notice of certain accidents. (a) Notice requirements. At the earliest practicable moment...

  12. 49 CFR 195.52 - Immediate notice of certain accidents.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 3 2013-10-01 2013-10-01 false Immediate notice of certain accidents. 195.52... TRANSPORTATION OF HAZARDOUS LIQUIDS BY PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.52 Immediate notice of certain accidents. (a) Notice requirements. At the earliest practicable moment...

  13. 49 CFR 195.52 - Telephonic notice of certain accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Telephonic notice of certain accidents. 195.52... TRANSPORTATION OF HAZARDOUS LIQUIDS BY PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.52 Telephonic notice of certain accidents. (a) At the earliest practicable moment following discovery of...

  14. 49 CFR 195.52 - Immediate notice of certain accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 3 2011-10-01 2011-10-01 false Immediate notice of certain accidents. 195.52... TRANSPORTATION OF HAZARDOUS LIQUIDS BY PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.52 Immediate notice of certain accidents. (a) Notice requirements. At the earliest practicable moment...

  15. Accident prevention in radiotherapy

    PubMed Central

    Holmberg, O

    2007-01-01

    In order to prevent accidents in radiotherapy, it is important to learn from accidents that have occurred previously. Lessons learned from a number of accidents are summarised and underlying patterns are looked for in this paper. Accidents can be prevented by applying several safety layers of preventive actions. Categories of these preventive actions are discussed together with specific actions belonging to each category of safety layer. PMID:21614274

  16. Professional experience and traffic accidents/near-miss accidents among truck drivers.

    PubMed

    Girotto, Edmarlon; Andrade, Selma Maffei de; González, Alberto Durán; Mesas, Arthur Eumann

    2016-10-01

    To investigate the relationship between the time working as a truck driver and the report of involvement in traffic accidents or near-miss accidents. A cross-sectional study was performed with truck drivers transporting products from the Brazilian grain harvest to the Port of Paranaguá, Paraná, Brazil. The drivers were interviewed regarding sociodemographic characteristics, working conditions, behavior in traffic and involvement in accidents or near-miss accidents in the previous 12 months. Subsequently, the participants answered a self-applied questionnaire on substance use. The time of professional experience as drivers was categorized in tertiles. Statistical analyses were performed through the construction of models adjusted by multinomial regression to assess the relationship between the length of experience as a truck driver and the involvement in accidents or near-miss accidents. This study included 665 male drivers with an average age of 42.2 (±11.1) years. Among them, 7.2% and 41.7% of the drivers reported involvement in accidents and near-miss accidents, respectively. In fully adjusted analysis, the 3rd tertile of professional experience (>22years) was shown to be inversely associated with involvement in accidents (odds ratio [OR] 0.29; 95% confidence interval [CI] 0.16-0.52) and near-miss accidents (OR 0.17; 95% CI 0.05-0.53). The 2nd tertile of professional experience (11-22 years) was inversely associated with involvement in accidents (OR 0.63; 95% CI 0.40-0.98). An evident relationship was observed between longer professional experience and a reduction in reporting involvement in accidents and near-miss accidents, regardless of age, substance use, working conditions and behavior in traffic. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. The effect of constraint on fuel-coolant interactions in a confined geometry

    SciTech Connect

    Park, H.; Corradini, M.L.

    1995-09-01

    A Fuel-Coolant Interaction (FCI or vapor explosion) is the phenomena in which a hot liquid rapidly transfers its internal energy into a surrounding colder and more volatile liquid. The energetics of such a complex multi-phase and multi-component phenomenon is partially determined by the surrounding boundary conditions. As one of the boundary conditions, we studied the effect of constraint on FCIs. The WFCI-D series of experiments were performed specifically to observe this effect. The results from these and our previous WFCI tests as well as those of other investigators are compared.

  2. 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

  3. Knock-limited performance of several internal coolants

    NASA Technical Reports Server (NTRS)

    Bellman, Donald R; Evvard, John C

    1945-01-01

    The effect of internal cooling on the knock-limited performance of an-f-28 fuel was investigated in a CFR engine, and the following internal coolants were used: (1) water, (2), methyl alcohol-water mixture, (3) ammonia-methyl alcohol-water mixture, (4) monomethylamine-water mixture, (5) dimethylamine-water mixture, and (6) trimethylamine-water mixture. Tests were run at inlet-air temperatures of 150 degrees and 250 degrees F. to indicate the temperature sensitivity of the internal-coolant solutions.

  4. Columbia's uninsulated water coolant lines in the payload bay

    NASA Technical Reports Server (NTRS)

    1997-01-01

    Yesterday, NASA decided to postpone for 24-hours the launch of Columbia on mission STS-83 due to a requirement to add additional thermal insulation to water coolant lines in the orbiter's payload bay. The water coolant lines are seen here winding their way around the window on the left. Managers determined that the lines, which cool various electronics on the orbiter, were not properly insulated and could possibly freeze during Columbia's 16-days in space. Columbia's launch is now set for 2:00 p.m. EST on Friday, April 4, 1997.

  5. A Review of Inherent Safety Characteristics of Metal-Alloy SFR Fuel Against Postulated Accidents

    SciTech Connect

    Sofu, Tanju

    2015-04-01

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

  6. Thermal-Hydraulic Performance of the TREAT Multi-SERTTA for Reactivity Initiated Accident Experiments

    SciTech Connect

    Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.; Woolstenhulme, Nicolas E.; Bess, John D.; O'Brien, Robert C.; Ban, Heng; Wachs, Daniel M.

    2016-08-01

    Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.

  7. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect

    Rebak, Raul B.

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  8. 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.

  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. 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.

  11. Analysis data on samples from the TMI-2 reactor-coolant system and reactor-coolant bleed tank

    SciTech Connect

    Nitschke, R.L.

    1982-05-01

    Two liquid samples from the Three Mile Island Unit 2 (TMI-2) Reactor Coolant System (RCS) and three liquid samples from the three Reactor Coolant Bleed Tanks (RCBT) were taken during the time period March 29, 1979 to August 14, 1980. The samples were analyzed for radionuclide concentrations by two independent laboratories, Exxon Nuclear Idaho Co., Inc. (ENICO) and EG and G Idaho, Inc. at the Idaho National Engineering Laboratory (INEL). The RCS sample taken on March 29, 1979 was also analyzed by Science Applications, Inc. (SAI). This report presents the methods used and the results of these analyes. 14 tables.

  12. 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

  13. Coolant-Control Valves For Fluid-Sampling Probes

    NASA Technical Reports Server (NTRS)

    Schultz, Donald F.

    1989-01-01

    Small built-in leaks prevent overheating. Downstream flow-control globe valve replaced with modified gate valve. Modification consists of drilling small hole through valve gate, so valve never turned completely off. This "leaky" valve provides enough flow of coolant to prevent overheating causing probe to fail. Principle also applied to automatic control system by installing small bypass line around control valve.

  14. 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.

  15. Coolants with selective optical filtering characteristics for ruby laser applications

    NASA Technical Reports Server (NTRS)

    Mc Devitt, F. R.; Rasquin, J. R.

    1968-01-01

    Coolant-filtering medium developed consists of a solution of copper sulfate in a 4-1 volumetric mixture of ethanol and methanol. This solution should be a useful addition to ruby laser systems, particularily in large pulse or Q switching applications.

  16. 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…

  17. 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...

  18. 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...

  19. 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...

  20. 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...

  1. 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…

  2. Substitution of Tolyltriazole for Mercaptobenzothiazole in Military Coolant Inhibitor Formulations

    DTIC Science & Technology

    coolant specifications (0.4% of the sodium salt ). 0.15% was found to be the optimum percentage of NaTT....antifreeze inhibitors with different percentages of sodium tolyltriazole (NaTT). The NaTT caused foaming in the tests, but a silicone type antifoam

  3. 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.

  4. Antimony tartrate corrosion inhibitive composition for coolant systems

    SciTech Connect

    Payerle, N.E.

    1987-08-11

    An automobile coolant concentrate is described comprising (a) a liquid polyhydric alcohol chosen from the group consisting of ethylene glycol, propylene glycol, diethylene glycol and mixtures thereof, and (b) corrosion inhibitors in a corrosion inhibitory amount with respect to corrosion of lead-containing solders, the corrosion inhibitors comprising (i) an alkali metal antimony tartrate, and (ii) an azole compound.

  5. 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.

  6. 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.

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

    SciTech Connect

    Trambauer, K.

    1997-07-01

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

  9. Repository preclosure accident scenarios

    SciTech Connect

    Yook, H.R.; Arbital, J.G.; Keeton, J.M.; Mosier, J.E.; Weaver, B.S.

    1984-09-01

    Waste-handling operations at a spent-fuel repository were investigated to identify operational accidents that could occur. The facility was subdivided, through systems engineering procedures, into individual operations that involve the waste and one specific component of the waste package, in one specific area of the handling facility. From this subdivision approximately 600 potential accidents involving waste package components were identified and then discussed. Supporting descriptive data included for each accident scenario are distance of drop, speed of collision, weight of package component, and weight of equipment involved. The energy of impact associated with each potential accident is calculated to provide a basis for comparison of the relative severities of all the accidents. The results and conclusions suggest approaches to accident consequence mitigation through waste package and facility design. 35 figures, 9 tables.

  10. Geometric and Road Environmental Effects against Total Number of Traffic Accidents in Kendari

    NASA Astrophysics Data System (ADS)

    Kurdin, M. Akbar; Welendo, La; Annisa, Nur

    2017-05-01

    From the large number of traffic accidents that occurred, the carrying of Kendari as the biggest contributor to accidents in the Southeast. The number of accidents in Kendari row since 2011 was recorded at 18 accidents due to the influence of geometric road, in 2012 registered at 13 accident and in 2013 amounted to 6 accidents, with accident data because of the influence Geometric recorded for 3 consecutive years the biggest contributor to accidents because of the influence of geometric is Abeli districts. This study aimed to determine the road which common point of accident-prone (Black spot) in Kecamatan Abeli as accident-prone areas in Kendari, analyze the influence of geometric and road environment against accidents on roads in Kecamatan Abeli, provide alternative treatment based on the causes of accidents on the location of the accident-prone points (blackspot) to reduce the rate of traffic accidents. From the results of a study of 6 curve the accident-prone locations, that the curve I, II, and VI is the “Black Spot” influenced by the amount and condition of traffic accidents, while at the curve II, a traffic accident that occurred also be caused by unsafe geometric where the type of geometric should be changed from Spiral-Spiral type to Spiral-Circle-Spiral type. This indicates geometric effect on the number of accidents.

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

    DOEpatents

    Chainer, Timothy J.; Coico, Patrick A.; Graybill, David P.; Iyengar, Madhusudan K.; Kamath, Vinod; Kochuparambil, Bejoy J.; Schmidt, Roger R.; Steinke, Mark E.

    2015-09-15

    A method is provided for fabricating a cooling apparatus for cooling an electronics rack, which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures, and a tube. The heat exchanger is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of coolant-carrying tube sections, each tube section having a coolant inlet and outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

  12. [Hanggliding accidents. Distribution of injuries and accident analysis].

    PubMed

    Ballmer, F T; Jakob, R P

    1989-12-01

    Paragliding--a relatively new sport to Switzerland--brought 23 patients with 48 injuries (38% lower limb and 29% spinal) within a period of 8 months to the Inselspital University hospital in Berne. The aim of the study in characterizing these injuries is to formulate some guidelines towards prevention. With over 90% of accidents occurring at either take off or landing, emphasis on better training for the beginner is proposed with strict guidelines for the more experienced pilot flying in unfavourable conditions.

  13. 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.

  14. Accident mortality among children

    PubMed Central

    Swaroop, S.; Albrecht, R. M.; Grab, B.

    1956-01-01

    The authors present statistics on mortality from accidents, with special reference to those relating to the age-group 1-19 years. For a number of countries figures are given for the proportional mortality from accidents (the number of accident deaths expressed as a percentage of the number of deaths from all causes) and for the specific death-rates, per 100 000 population, from all causes of death, from selected causes, from all causes of accidents, and from various types of accident. From these figures it appears that, in most countries, accidents are becoming relatively increasingly prominent as a cause of death in childhood, primarily because of the conquest of other causes of death—such as infectious and parasitic diseases, which formerly took a heavy toll of children and adolescents—but also to some extent because the death-rate from motor-vehicle accidents is rising and cancelling out the reduction in the rate for other causes of accidental death. In the authors' opinion, further epidemiological investigations into accident causation are required for the purpose of devising quicker and more effective methods of accident prevention. PMID:13383361

  15. [Accidents with the "paraglider"].

    PubMed

    Lang, T H; Dengg, C; Gabl, M

    1988-09-01

    With a collective of 46 patients we show the details and kinds of accidents caused by paragliding. The base for the casuistry of the accidents was a questionnaire which was answered by most of the injured persons. These were questions about the theoretical and practical training, the course of the flight during the different phases, and the subjective point of view of the course of the accident. The patterns of the injuries showed a high incidence of injuries of the spinal column and high risks for the ankles. At the end, we give some advice how to prevent these accidents.

  16. Transport Aircraft Accident Dynamics

    DTIC Science & Technology

    1982-03-01

    SA Contracti’ Report 165850 S,">.AA 11’ýport No. VL)T-FAR-CT-8Z-T0 i0 u; S𔃺/3!D35" *. T -ansport Aircraft Accident Dynamics Reproduced From Srt...City, New Jersey 08405 84 04-. , NASA Contractor Report FAA Report No. Dbr-MA -C7 ,-70 TRANSPORT AIRCRAFT. ACCIDENT DYNAMICS I A. COMINSKY, et al...transport aircraft accidents , the association between structural systems and accident injuries and the identification of typical scenarios. This report

  17. Comparison of the Aerospace Systems Test Reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

    SciTech Connect

    Warinner, D.K.

    1983-01-01

    This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 10/sup 4/ s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code.

  18. Forced flow evaporator for unusual gravity conditions

    NASA Technical Reports Server (NTRS)

    Niggemann, Richard E. (Inventor); Ellis, Wilbert E. (Inventor)

    1987-01-01

    Low efficiency heat transfer in evaporators subject to unusual gravitational conditions is avoided through the use of a spiral evaporator conduit 12 receiving at an inlet 14 a vaporizable coolant at least partly in the liquid phase. Flow of the coolant through the conduit 12 demists the coolant by centrifuging the liquid phase against a pressurre wall 44 of the conduit 12. Vapor flow 40 induces counterrotating vortices 46, 48 which circulate the liquid phase coolant around the interior of the conduit 12 to wet all surfaces thereof.

  19. Development and Validation of Accident Models for FeCrAl Cladding

    SciTech Connect

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean

    2016-08-01

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

  20. An ISP-27 accident scenario for analysis of Krsko Nuclear Power Plant SBLOCA

    SciTech Connect

    Petelin, S.; Mavko, B.; Gortnar, O.; Parzer, I.

    1994-12-31

    The reactor safety analysis group of Jozef Stefan Institute (IJS) has participated in analyses of International Standard Problem 27 (ISP-27), which was based on test 9.1 b performed at the BETHSY experimental facility (France). In addition, we realized the ISP-27 transient scenario in the analysis of a small-break loss-of-coolant accident (SBLOCA) for Krsko nuclear power plant (NPP). The objective was to evaluate the effectiveness of the ISP-27 proposed accident management procedure for a real NPP and to compare the physical phenomena known from experimental background with the phenomena predicted by simulation of a real plant transient.

  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. TMI-2 accident: core heat-up analysis

    SciTech Connect

    Ardron, K.H.; Cain, D.G.

    1981-01-01

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions.

  3. Experimental evaluation of machining parameters in machining of 7075 aluminium alloy with cryogenic liquid nitrogen coolant

    NASA Astrophysics Data System (ADS)

    Muthuraman, V.; Arunkumar, S.

    2017-03-01

    The experimental results on investigation on the turning of 7075 aluminium alloy, using Cryogenic Liquid Nitrogen (LN2) as a coolant was analyzed in this paper. The influence of the cryogenic LN2 coolant compared with that of the conventional coolant on the cuting performance parameters, such as the cutting force, cutting temperature, andbsurface finish was analysed and investigated. The use of the cryogenic liquid nitrogen coolant influenced the cutting temperature and the cutting force by about 17 to 29% and 11 to 20% reduction respectively. The surface finish value of the machined workpiece is about 15 to 23% better than that of the conventional coolant.

  4. CFD Analyses of Air-Ingress Accident for VHTRs

    NASA Astrophysics Data System (ADS)

    Ham, Tae Kyu

    -ingress mechanism and to utilize the CFD simulation in the analysis of the phenomenon. Previous air-ingress studies simulated the depressurization process using simple assumptions or 1-D system code results. However, recent studies found flow oscillations near the end of the depressurization which could influence the next stage of the air-ingress accident. Therefore, CFD simulations were performed to examine the air-ingress mechanisms from the depressurization through the establishment of local natural circulation initiate. In addition to the double-guillotine break scenario, there are other scenarios that can lead to an air-ingress event such as a partial break were in the cross vessel with various break locations, orientations, and shapes. These additional situations were also investigated. The simulation results for the OSU test facility showed that the discharged helium coolant from a reactor vessel during the depressurization process will be mixed with the air in the containment. This process makes the density of the gas mixture in the containment lower and the density-driven air-ingress flow slower because the density-driven flow is established by the density difference of the gas species between the reactor vessel and the containment. In addition, for the simulations with various initial and boundary conditions, the simulation results showed that the total accumulated air in the containment collapsed within 10% standard deviation by: 1. multiplying the density ratio and viscosity ratio of the gas species between the containment and the reactor vessel and 2. multiplying the ratio of the air mole fraction and gas temperature to the reference value. By replacing the gas mixture in the reactor cavity with a gas heavier than the air, the air-ingress speed slowed down. Based on the understanding of the air-ingress phenomena for the GT-MHR air-ingress scenario, several mitigation measures of air-ingress accident are proposed. The CFD results are utilized to plan experimental

  5. Analysis of main steam isolation valve leakage in design basis accidents using MELCOR 1.8.6 and RADTRAD.

    SciTech Connect

    Salay, Michael; Kalinich, Donald A.; Gauntt, Randall O.; Radel, Tracy E.

    2008-10-01

    Analyses were performed using MELCOR and RADTRAD to investigate main steam isolation valve (MSIV) leakage behavior under design basis accident (DBA) loss-of-coolant (LOCA) conditions that are presumed to have led to a significant core melt accident. Dose to the control room, site boundary and LPZ are examined using both approaches described in current regulatory guidelines as well as analyses based on best estimate source term and system response. At issue is the current practice of using containment airborne aerosol concentrations as a surrogate for the in-vessel aerosol concentration that exists in the near vicinity of the MSIVs. This study finds current practice using the AST-based containment aerosol concentrations for assessing MSIV leakage is non-conservative and conceptually in error. A methodology is proposed that scales the containment aerosol concentration to the expected vessel concentration in order to preserve the simplified use of the AST in assessing containment performance under assumed DBA conditions. This correction is required during the first two hours of the accident while the gap and early in-vessel source terms are present. It is general practice to assume that at {approx}2hrs, recovery actions to reflood the core will have been successful and that further core damage can be avoided. The analyses performed in this study determine that, after two hours, assuming vessel reflooding has taken place, the containment aerosol concentration can then conservatively be used as the effective source to the leaking MSIV's. Recommendations are provided concerning typical aerosol removal coefficients that can be used in the RADTRAD code to predict source attenuation in the steam lines, and on robust methods of predicting MSIV leakage flows based on measured MSIV leakage performance.

  6. 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.

  7. 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.

  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. 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.

  10. Optimization of fast breeder reactors employing innovative liquid metal coolants

    SciTech Connect

    Pilarski, Stevan

    2007-07-01

    In this paper we propose a comparative assessment of fast breeder reactor core concepts employing Pb, Pb- Mg and Pb-{sup 7}Li as primary coolants and oxide and nitride fuels. Starting from a common reference core to make the comparison relevant, each coolant candidate is associated to an optimized design that takes into account its specific physical properties. For each core, we perform a neutronic analysis and an assessment of its safety potential. In comparison with the case of Pb, the use of Pb-Mg and Pb-{sup 7}Li increases the void reactivity effect. On the other hand, the breeding gain also increases, and the Doppler effect is enhanced, leading to a favorable behaviour concerning safety. (author)

  11. Peculiarities of evolution of shock waves generated by boiling coolant

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

    Simulation of compression wave generation and evolution at the disk target was performed for the case of explosive-type boiling of coolant; the boiling is initiated by endwall rupture of a high-pressure pipeline. The calculations were performed for shock wave amplitude at different times and modes of pipe rupture. The simulated pressure of a target-reflected shock wave is different from the theoretical value for ideal gas; this discrepancy between simulation and theory becomes lower at higher distances of flow from the nozzle exit. Comparative simulation study was performed for flow of two-phase coolant with account for slip flow effect and for different sizes of droplets. Simulation gave the limiting droplet size when the single-velocity homogeneous flow model is valid, i.e., the slip flow effect is insignificant.

  12. Columbia Accident Investigation Board

    NASA Image and Video Library

    2003-02-13

    Members of the Columbia Accident Investigation Board examine pieces of Columbia debris in the RLV Hangar. The debris was shipped from the collection point at Barksdale Air Force Base, Shreveport, La. As part of the ongoing investigation into the tragic accident that claimed Columbia and her crew of seven, workers will attempt to reconstruct the orbiter inside the RLV.

  13. Farm accidents in children.

    PubMed

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

    1992-07-04

    To examine the problem of accidental injury to children on farms. 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. 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. Children aged under 16. Death or injury after farm related accidents. 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. 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.

  14. 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…

  15. 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.

  16. 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

  17. 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.

  18. Rotating solid radiative coolant system for space nuclear reactors

    SciTech Connect

    Apley, W.J.; Babb, A.L.

    1988-05-01

    The RING power system described in this paper is proposed as a primary or emergency heat rejection system for advanced space reactor power applications. The system employs a set of four (4) counter-rotating, 90 degree offset, coolant-carrying rings. The rings (segmented, corrugated, finned, thin-walled pipes, filled with liquid lithium) pass through a cavity heat exchanger and reradiate the absorbed heat to the space environment. 25 refs., 6 figs., 3 tabs.

  19. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    DTIC Science & Technology

    2006-12-01

    Nuclear Data File/B-Version V FENDL Fusion Evaluated Nuclear Data Library HLMC Heavy Liquid Metal Coolant LBE Lead-Bismuth Eutectic LMFBR ...According to a Generation IV Nuclear Energy Systems report29 released February 21, 2006, the core outlet temperature and peak cladding temperature...24 sodium for Liquid Metal Fast Breeder Reactor ( LMFBR ) systems worldwide.32 The pumping power requirements for lead, LBE and tin are much

  20. Effect of reactor coolant radioactivity upon configuration feasibility for a nuclear electric propulsion vehicle

    NASA Technical Reports Server (NTRS)

    Soffer, L.; Wright, G. N.

    1973-01-01

    A preliminary shielding analysis was carried out for a conceptual nuclear electric propulsion vehicle designed to transport payloads from low earth orbit to synchronous orbit. The vehicle employed a thermionic nuclear reactor operating at 1575 kilowatts and generated 120 kilowatts of electricity for a round-trip mission time of 2000 hours. Propulsion was via axially directed ion engines employing 3300 pounds of mercury as a propellant. The vehicle configuration permitted a reactor shadow shield geometry using LiH and the mercury propellant for shielding. However, much of the radioactive NaK reactor coolant was unshielded and in close proximity to the power conditioning electronics. An estimate of the radioactivity of the NaK coolant was made and its unshielded dose rate to the power conditioning equipment calculated. It was found that the activated NaK contributed about three-fourths of the gamma dose constraint. The NaK dose was considered a sufficiently high fraction of the allowable gamma dose to necessitate modifications in configuration.

  1. Accidents on hospital wards.

    PubMed Central

    Levene, S; Bonfield, G

    1991-01-01

    Eight hospitals reported 781 non-iatrogenic accidents occurring to patients and visitors under 16 years of age during an 18 month period up to October 1989. Accidents more often involved boys and children aged 3 to 5 years old. Falls from a height, slips, and striking accidents were common by day and falls by night. A total of 41% of accidents to inpatients occurred when parents were present. Only three accidents were serious. Altogether 27% involved beds and cots, and only one consequent injury was more than minor. Data collected routinely in case of medicolegal action can be presented in a form that may facilitate preventative work. Potentially remediable causes for concern include falls from beds and cots and the use of makeshift equipment. PMID:1929510

  2. 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.

  3. 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.

  4. 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

  5. [Accidents and injuries at work].

    PubMed

    Standke, W

    2014-06-01

    In the case of an accident at work, the person concerned is insured by law according to the guidelines of the Sozialgesetzbuch VII as far as the injuries have been caused by this accident. The most important source of information on the incident in question is the accident report that has to be sent to the responsible institution for statutory accident insurance and prevention by the employer, if the accident of the injured person is fatal or leads to an incapacity to work for more than 3 days (= reportable accident). Data concerning accidents like these are sent to the Deutsche Gesetzliche Unfallversicherung (DGUV) as part of a random sample survey by the institutions for statutory accident insurance and prevention and are analyzed statistically. Thus the key issues of accidents can be established and used for effective prevention. Although the success of effective accident prevention is undisputed, there were still 919,025 occupational accidents in 2011, with clear gender-related differences. Most occupational accidents involve the upper and lower extremities. Accidents are analyzed comprehensively and the results are published and made available to all interested parties in an effort to improve public awareness of possible accidents. Apart from reportable accidents, data on the new occupational accident pensions are also gathered and analyzed statistically. Thus, additional information is gained on accidents with extremely serious consequences and partly permanent injuries for the accident victims.

  6. Study on the effect of the impeller and diffuser blade number on reactor coolant pump performances

    NASA Astrophysics Data System (ADS)

    Long, Y.; Yin, J. L.; Wang, D. Z.; Li, T. B.

    2016-05-01

    In this paper, CFD approach was employed to study how the blade number of impeller and diffuser influences reactor coolant pump performances. The three-dimensional pump internal flow channel was modelled by pro/E software, Reynolds-averaged Naiver-Stokes equations with the k-ε turbulence model were solved by the computational fluid dynamics software CFX. By post-processing on the numerical results, the performance curves of reactor coolant pump were obtained. The results are as follows, with the blade number of the impeller increasing, the head of the pump with different diffuser universally increases in the 8Q n∼1.2Q n conditions, and at different blade number of the diffuser, the head increases with the blade number of the impeller increasing. In 1.0Q n condition, when the blades number combination of impeller and diffuser chooses 4+16, 7+14 and 6+18, the head curves exist singular points. In 1.2Q n condition, the head curve still exists singular point in 6+18. With the blade number of the impeller increasing, the efficiency of the pump with different diffuser universally decreases in the 0.8Q n and 1.0Q n conditions, but in 1.2Q n condition, the efficiency of the pump with different diffuser universally increases. In 1.0Q n condition, the impellers of 4 and 5 blades are better. When the blade number combination of impeller and diffuser choose 4+11, 4+17, 4+18, 5+12, 5+17 and 5+18, the efficiencies relatively have higher values. With the blade number of the impeller increasing, the hydraulic shaft power of the pump with different diffuser universally increases in the 0.8Q n∼1.2Q n conditions, and with the blade number of the diffuser increasing, the power of different impeller overall has small fluctuation, but tends to be uniform. This means the increase of the diffuser blade number has less influence on shaft power.The influence on the head and flow by the matching relationship of the blades number between impeller and diffuser is very complicated, which

  7. 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.

  8. [Severe parachuting accident. Analysis of 122 cases].

    PubMed

    Krauss, U; Mischkowsky, T

    1993-06-01

    Based on a population of 122 severely injured patients the causes of paragliding accidents and the patterns of injury are analyzed. A questionnaire is used to establish a sport-specific profile for the paragliding pilot. The lower limbs (55.7%) and the lower parts of the spine (45.9%) are the most frequently injured parts of the body. There is a high risk of multiple injuries after a single accident because of the tremendous axial power. The standard of equipment is good in over 90% of the cases. Insufficient training and failure to take account of geographical and meteorological conditions are the main determinants of accidents sustained by paragliders, most of whom are young. Nevertheless, 80% of our patients want to continue paragliding. Finally some advice is given on how to prevent paragliding accidents and injuries.

  9. 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. © 2010 The Author(s). Journal compilation © Overseas Development Institute, 2010.

  10. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    NASA Astrophysics Data System (ADS)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  11. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    SciTech Connect

    Waris, Abdul Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki; Aji, Indarta K.

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  12. [A monitoring system for work-related accidents in Piracicaba, São Paulo, Brazil].

    PubMed

    Cordeiro, Ricardo; Vilela, Rodolfo Andrade Gouveia; de Medeiros, Maria Angélica Tavares; Gonçalves, Cláudia Giglio de Oliveira; Bragantini, Clarice Aparecida; Varolla, Antenor J; Celso, Stephan

    2005-01-01

    The authors report on the development of a work accident monitoring system in Piracicaba, São Paulo State, Brazil, with the following characteristics: information feeding the system is obtained in real time directly from accident treatment centers; the system has universal monitoring, covering all work-related accidents in Piracicaba, regardless of the nature of the worker's employment conditions, place of work, or place of residence; health surveillance and promotion of health initiatives are triggered by identification of sentinel events; spatial distribution analysis of work-related accidents is a basic tool in designing accident awareness strategies and accident prevention policies. The system was implemented in November 2003 and by October 2004 had identified 5,320 work-related accidents, or a 3.8% annual proportional incidence of work-related accidents in the municipal area. We illustrate spatial analysis of registered work-related accidents and present a detailed investigation of one example of a serious accident.

  13. Fluid flow analysis of the SSME high pressure fuel and oxidizer turbine coolant systems

    NASA Technical Reports Server (NTRS)

    Teal, G. A.

    1989-01-01

    The objective is to provide improved analysis capability for the Space Shuttle Main Engine (SSME) high pressure fuel and oxidizer turbine coolant systems. Each of the systems was analyzed to determine fluid flow rate and thermodynamic and transport properties at all key points in the systems. Existing computer codes were used as a baseline for these analyses. These codes were modified to provide improved analysis capability. The major areas of improvement are listed. A review of the drawings was performed, and pertinent geometry changes were included in the models. Improvements were made in the calculation of thermodynamic and transport properties for a mixture of hydrogen and steam. A one-dimensional turbine model for each system is included as a subroutine to each code. This provides a closed loop analysis with a minimum of required boundary conditions as input. An improved labyrinth seal model is included in the high pressure fuel turbine coolant model. The modifications and the analysis results are presented in detail.

  14. Corrosion of Ferritic Steels in High Temperature Molten Salt Coolants for Nuclear Applications

    SciTech Connect

    Farmer, J; El-Dasher, B; de Caro, M S; Ferreira, J

    2008-11-25

    Corrosion of ferritic steels in high temperature molten fluoride salts may limit the life of advanced reactors, including some hybrid systems that are now under consideration. In some cases, the steel may be protected through galvanic coupling with other less noble materials with special neutronic properties such a beryllium. This paper reports the development of a model for predicting corrosion rates for various ferritic steels, with and without oxide dispersion strengthening, in FLiBe (Li{sub 2}BeF{sub 4}) and FLiNaK (Li-Na-K-F) coolants at temperatures up to 800 C. Mixed potential theory is used to account for the protection of steel by beryllium, Tafel kinetics are used to predict rates of dissolution as a function of temperature and potential, and the thinning of the mass-transfer boundary layer with increasing Reynolds number is accounted for with dimensionless correlations. The model also accounts for the deceleration of corrosion as the coolants become saturated with dissolved chromium and iron. This paper also reports electrochemical impedance spectroscopy of steels at their corrosion potentials in high-temperature molten salt environments, with the complex impedance spectra interpreted in terms of the interfacial charge transfer resistance and capacitance, as well as the electrolyte conductivity. Such in situ measurement techniques provide valuable insight into the degradation of materials under realistic conditions.

  15. 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.

  16. 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.

  17. [Multicenter paragliding accident study 1990].

    PubMed

    Lautenschlager, S; Karli, U; Matter, P

    1992-01-01

    During the period from 1.1.90 until 31.12.90, 86 injuries associated with paragliding were analyzed in a prospective study in 12 different Swiss hospitals with reference to causes, patterns, and frequencies. The injuries showed a mean score of over 2 and were classified as severe. Most frequent spine injuries (36%) and lesions of the lower extremity (35%) with a high risk of the ankles were diagnosed. One accident was fatal. 60% of the accidents happened during landing, 26% during launching and 14% during flight. Half of the pilots were affected during their primary training course. Most accidents were caused by inflight error of judgement--especially incorrect estimation of wind conditions--and further the choice of unfavourable landing sites. In contrast to previous injury-reports, only one equipment failure could be noted, but often the equipment was not corresponding with the experience and the weight of the pilot. To reduce the frequency of paragliding-injuries an accurate choice of equipment and an increased attention to environmental factors is mandatory. Furthermore an education-program regarding the attitude and intelligence of the pilot should be included in training courses.

  18. Analysis of Maximum Reasonably Foreseeable Accidents for the Yucca Mountain Draft Environmental Impact Statement (DEIS)

    SciTech Connect

    S.B. Ross; R.E. Best; S.J. Maheras; T.I. McSweeney

    2001-08-17

    Accidents could occur during the transportation of spent nuclear fuel and high-level radioactive waste. This paper describes the risks and consequences to the public from accidents that are highly unlikely but that could have severe consequences. The impact of these accidents would include those to a collective population and to hypothetical maximally exposed individuals (MEIs). This document discusses accidents with conditions that have a chance of occurring more often than 1 in 10 million times in a year, called ''maximum reasonably foreseeable accidents''. Accidents and conditions less likely than this are not considered to be reasonably foreseeable.

  19. Summary on the depressurization from supercritical pressure conditions

    SciTech Connect

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

    2012-07-01

    When a fluid discharges from a high pressure and temperature system, a 'choking' or critical condition occurs, and the flow rate becomes independent of the downstream pressure. During a postulated loss of coolant accident (LOCA) of a water reactor the break flow will be subject to this condition. An accurate estimation of the critical flow rate is important for the evaluation of the reactor safety, because this flow rate controls the loss of coolant inventory and energy from the system, and thus has a significant effect on the accident consequences[1]. In the design of safety systems for a super critical water reactor (SCWR), postulated LOCA transients are particularly important due to the lower coolant inventory compared to a typical PWR for the same power output. This lower coolant inventory would result in a faster transient response of the SCWR, and hence accurate prediction of the critical discharge is mandatory. Under potential two-phase conditions critical flow is dominated by the vapor content or quality of the vapor, which is closely related with the onset of vaporization and the interfacial interaction between phases [2]. This presents a major challenge for the estimation of the flow rate due to the lack of the knowledge of those processes, especially under the conditions of interest for the SCWR. According to the limited data of supercritical fluids, the critical flows at conditions above the pseudo-critical point seem to be fairly stable and consistent with the subcritical homogeneous equilibrium model (HEM) model predictions, while having a lower flow rate than those in the two-phase region. Thus the major difficulty in the prediction of the depressurization flow rates remains in the region where two phases co-exist at the top of the vapor dome. In this region, the flow rate is strongly affected by the nozzle geometry and tends to be unstable. Various models for this region have been developed with different assumptions, e.g. the HEM and Moody model [3

  20. Effect of coolant chemistry on PWR radiation transport processes. Progress report on reactor loop studies

    SciTech Connect

    Brown, D.J.; Flynn, G.; Haynes, J.W.; Kitt, G.P.; Large, N.R.; Lawson, D.; Mead, A.P.; Nichols, J.L.; Woodwark, D.R.

    1986-05-01

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. There are strong indications that the in-core deposition behavior of corrosion product species is not fully accounted for by the solubility model based on nickel ferrite; boric acid plays a role apart from its influence on pH, and corrosion products are adsorbed to some extent in the zirconium oxide film on the fuel cladding. In DWL, soluble species appear to be dominant in deposition processes. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. 13 figs.

  1. Scoping studies of vapor behavior during a severe accident in a metal-fueled reactor

    SciTech Connect

    Spencer, B.W.; Marchaterre, J.F.

    1985-04-15

    Scoping calculations have been performed examining the consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel. The principal gas and vapor species released are shown to be Xe, Cs,and bond sodium contained within the fuel porosity. Fuel vapor pressure is insignificant, and there is no energetic fuel-coolant interaction for the conditions considered. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core (although reactor-material experiments are needed to confirm these high condensation rates). If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the implication is that the ability of vapor expansion to perform appreciable work on the system is largely eliminated. Furthermore, the ability of an expanding vapor bubble to transport fuel and fission product species to the cover gas region where they may be released to the containment is also largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool.

  2. Study on recriticality of fuel debris during hypothetical severe accidents in the Advanced Neutron Source reactor

    SciTech Connect

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.; Shin, S.T.

    1995-09-01

    A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of k{sub eff} on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough.

  3. Anthropotechnological analysis of industrial accidents in Brazil.

    PubMed Central

    Binder, M. C.; de Almeida, I. M.; Monteau, M.

    1999-01-01

    The Brazilian Ministry of Labour has been attempting to modify the norms used to analyse industrial accidents in the country. For this purpose, in 1994 it tried to make compulsory use of the causal tree approach to accident analysis, an approach developed in France during the 1970s, without having previously determined whether it is suitable for use under the industrial safety conditions that prevail in most Brazilian firms. In addition, opposition from Brazilian employers has blocked the proposed changes to the norms. The present study employed anthropotechnology to analyse experimental application of the causal tree method to work-related accidents in industrial firms in the region of Botucatu, São Paulo. Three work-related accidents were examined in three industrial firms representative of local, national and multinational companies. On the basis of the accidents analysed in this study, the rationale for the use of the causal tree method in Brazil can be summarized for each type of firm as follows: the method is redundant if there is a predominance of the type of risk whose elimination or neutralization requires adoption of conventional industrial safety measures (firm representative of local enterprises); the method is worth while if the company's specific technical risks have already largely been eliminated (firm representative of national enterprises); and the method is particularly appropriate if the firm has a good safety record and the causes of accidents are primarily related to industrial organization and management (multinational enterprise). PMID:10680249

  4. Vegetable oils: Liquid coolants for solar heating and cooling applications

    NASA Astrophysics Data System (ADS)

    Ingley, H. A.

    1980-02-01

    Vegetable oils, renewable byproducts of agriculture processes, were investigated for possible use as liquid coolants. Several thermophysical properties of four vegetable oils were investigated. Vapor pressures, specific heat, viscosity, density, and thermal conductivity were determined over a range of temperatures for corn, soybeans, 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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)

  10. Effects of quenched randomness induced by car accidents on traffic flow in a cellular automata model.

    PubMed

    Yang, Xian-Qing; Ma, Yu-Qiang; Zhao, Yue-Min

    2004-10-01

    In this paper we numerically study the impact of quenched disorder induced by car accidents on traffic flow in the Nagel-Schreckenberg (NS) model. Car accidents occur when the necessary conditions proposed by [J. Phys. A 30, 3329 (1997)

  11. 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.

  12. Analysis of automobile radiator performance with ethylene glycol/water and propylene glycol/water coolants

    SciTech Connect

    Gollin, M.; Bjork, D.

    1996-12-31

    The heat transfer and hydraulic performance of the following coolants was examined in five automobile radiators in a wind tunnel: 100% water; 100% propylene glycol; 70/30 propylene glycol/water (volume); 50/50 propylene glycol/water (volume); 70/30 ethylene glycol/water (volume); 50/50 ethylene glycol water (volume). The results of these studies are presented to demonstrate the relative performance of these coolant mixtures in terms of heat transfer, coolant pressure drop and radiator effectiveness for a range of coolant and air flowrates. It is concluded that the most effective of the coolants in transferring heat in the test radiators was water, followed by 50/50 ethylene glycol/water, 50/50 propylene glycol/water, 70/30 ethylene glycol/water, 70/30 propylene glycol and, finally, 100% propylene glycol. There will be a negligible differences between the performance of a radiator using a 50/50 propylene glycol/water coolant and a 50/50 ethylene glycol/water coolant. It is estimated that, with 50/50 propylene glycol coolant replacing 50/50 ethylene glycol/water, the temperature of the coolant throughout the cooling loop will increase by approximately 5%. The effect that the flow regime (fully turbulent/transition/laminar) has upon the performance of a given radiator/coolant combination was found to be significant. The design of the coolant passages in radiators can affect the onset of fully turbulent flow in the coolant passages in a radiator.

  13. Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

    SciTech Connect

    Rempe, Joy Lynn; Knudson, Darrell Lee

    2014-09-01

    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. 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 initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. 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. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report 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: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, 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 to instruments easily removed from the TMI-2 plant for evaluations. As documented

  14. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems: Preprint

    SciTech Connect

    Titov, Eugene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-03-22

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  15. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    SciTech Connect

    Williams, D.F.

    2006-03-24

    transfer and nuclear performance metrics. Lighter salts also tend to have more favorable (larger) moderating ratios, and thus should have a more favorable coolant-voiding behavior in-core. Heavy (high-Z) salts tend to have lower heat capacities and thermal conductivities and more significant activation and transmutation products. However, all of the salts are relatively good heat-transfer agents. A detailed discussion of each property and the combination of properties that served as a heat-transfer metric is presented in the body of this report. In addition to neutronic metrics, such as moderating ratio and neutron absorption, the activation properties of the salts were investigated (Table C). Again, lighter salts tend to have more favorable activation properties compared to salts with high atomic-number constituents. A simple model for estimating the reactivity coefficients associated with a reduction of salt content in the core (voiding or thermal expansion) was also developed, and the primary parameters were investigated. It appears that reasonable design flexibility exists to select a safe combination of fuel-element design and salt coolant for most of the candidate salts. Materials compatibility is an overriding consideration for high-temperature reactors; therefore the question was posed whether any one of the candidate salts was inherently, or significantly, more corrosive than another. This is a very complex subject, and it was not possible to exclude any fluoride salts based on the corrosion database. The corrosion database clearly indicates superior container alloys, but the effect of salt identity is masked by many factors which are likely more important (impurities, redox condition) in the testing evidence than salt identity. Despite this uncertainty, some reasonable preferences can be recommended, and these are indicated in the conclusions. The reasoning to support these conclusions is established in the body of this report.

  16. 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.

  17. Accident progression event tree analysis for postulated severe accidents at N Reactor

    SciTech Connect

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. ); Medford, G.T. )

    1990-06-01

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied.

  18. Tobit analysis of vehicle accident rates on interstate highways.

    PubMed

    Anastasopoulos, Panagiotis Ch; Tarko, Andrew P; Mannering, Fred L

    2008-03-01

    There has been an abundance of research that has used Poisson models and its variants (negative binomial and zero-inflated models) to improve our understanding of the factors that affect accident frequencies on roadway segments. This study explores the application of an alternate method, tobit regression, by viewing vehicle accident rates directly (instead of frequencies) as a continuous variable that is left-censored at zero. Using data from vehicle accidents on Indiana interstates, the estimation results show that many factors relating to pavement condition, roadway geometrics and traffic characteristics significantly affect vehicle accident rates.

  19. Hard metal lung disease: importance of cobalt in coolants.

    PubMed Central

    Sjögren, I; Hillerdal, G; Andersson, A; Zetterström, O

    1980-01-01

    Four patients were found to react to occupational exposure to grinding of hard metal (tungsten carbide). Three of the patients had symptoms and signs compatible with an allergic alveolitis, the symptoms disappearing and the chest radiograph clearing when they were absent from work for a few months. Re-exposure to the offending agent led to new signs and symptoms. The first patient was re-exposed twice and each time reacted a little more seriously. After the last episode her chest radiograph has not cleared completely, in contrast to the first two times. The fourth patient had more typical occupational asthma. All the cases occurred in the part of the factory where air concentrations of cobalt were the lowest. The cobalt there is dissolved in the coolant necessary for grinding the hard metal. It occurs mainly in the ionised form, which is known to react with proteins and therefore presumably acts as a hapten. Protective measures, including choosing a coolant with minimal ability to dissolve cobalt and an effective exhaust system, should minimise the risk of this occupational disease in the future. PMID:7444839

  20. Fitness for service assessment of coolant channels of Indian PHWRs

    NASA Astrophysics Data System (ADS)

    Sinha, R. K.; Sinha, S. K.; Madhusoodanan, K.

    2008-12-01

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17 n/m 2 s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.