WorldWideScience

Sample records for blowdown

  1. Steam generator blowdown system upgrades

    International Nuclear Information System (INIS)

    Lee, Tien P.; Kim, David H.; Jindal, Krishan K.

    2004-01-01

    The steam generator blowdown (SGBD) system is used to remove impurities from the steam generators in order to maintain steam generator (SG) water chemistry within specifications. The original SGBD systems at Diablo Canyon power plant (DCPP) were designed in the early 1970s, and since that time the industry has changed its practices regarding water chemistry. DCPP has operated its SGBD system above its design flow rate. This resulted in a history of high maintenance and unreliable operation. Subsequently, DCPP implemented extensive modifications in order to accommodate the higher industry standard flow rates. These modifications resulted in a more reliable and rugged system. Additionally, significant savings were realized due to an increase in net plant output and a reduction in the required plant makeup water by recovering steam generator blowdown. (author)

  2. LMR steam generator blowdown with RETRAN

    International Nuclear Information System (INIS)

    Wei, T.Y.C.

    1985-01-01

    One of the transients being considered in the FSAR Chapter 15 analyses of anticipated LMR transients is the fast blowdown of a steam generator upon inadvertent actuation of the liquid metal/water reaction mitigation system. For the blowdown analysis, a stand-alone steam generator model for the IFR plant was constructed using RETRAN

  3. BWR drywell behavior under steam blowdown

    International Nuclear Information System (INIS)

    NguyenLe, Q.

    1998-01-01

    Historically, thermal hydraulics analyses on Large Break Loss of Coolant Accidents (LOCA) have been focused on the transients within the reactor or steam generator. Few have studied the effects of steam blowdown on the containment building. This paper discusses some theoretical issues as well as presenting numerical and experimental results of the blowdown tests performed at the Purdue University Multi-Dimensional Integrated Test Assembly (PUMA)

  4. PPOOLEX experiments with two parallel blowdown pipes

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2011-01-15

    This report summarizes the results of the experiments with two transparent blowdown pipes carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through either one or two vertical transparent blowdown pipes to the condensation pool. Five experiments with one pipe and six with two parallel pipes were carried out. The main purpose of the experiments was to study loads caused by chugging (rapid condensation) while steam is discharged into the condensation pool filled with sub-cooled water. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. In the experiments the initial temperature of the condensation pool water varied from 12 deg. C to 55 deg. C, the steam flow rate from 40 g/s to 1 300 g/s and the temperature of incoming steam from 120 deg. C to 185 deg. C. In the experiments with only one transparent blowdown pipe chugging phenomenon didn't occur as intensified as in the preceding experiments carried out with a DN200 stainless steel pipe. With the steel blowdown pipe even 10 times higher pressure pulses were registered inside the pipe. Meanwhile, loads registered in the pool didn't indicate significant differences between the steel and polycarbonate pipe experiments. In the experiments with two transparent blowdown pipes, the steamwater interface moved almost synchronously up and down inside both pipes. Chugging was stronger than in the one pipe experiments and even two times higher loads were measured inside the pipes. The loads at the blowdown pipe outlet were approximately the same as in the one pipe cases. Other registered loads around the pool were about 50-100 % higher than with one pipe. The experiments with two parallel blowdown pipes gave contradictory results compared to the earlier studies dealing with chugging loads in case of multiple pipes. Contributing

  5. After the blowdown: a resource assessment of the Boundary Waters Canoe Area Wilderness, 1999-2003

    Science.gov (United States)

    W. Keith Moser; Mark H. Hansen; Mark D. Nelson; Susan J. Crocker; Charles H. Perry; Bethany Schulz; Christopher W. Woodall

    2007-01-01

    The Boundary Waters Canoe Area Wilderness (BWCAW) was struck by a major windstorm on July 4, 1999. Estimated volume in blowdown areas was up to 29 percent less than in non-blowdown areas. Mean down woody fuel loadings were twice as high in blowdown areas than in non-blowdown areas. Overstory species diversity declined in blowdown areas, but understory diversity,...

  6. Construction of the blowdown and condensation loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, Choon Kyung; Song, Chul Kyung; Cho, Seok; Chun, S. Y.; Chung, Moon Ki

    1997-12-01

    The blowdown and condensation loop (B and C loop) has been constructed to get experimental data for designing the safety depressurization system (SDS) and steam sparger which are considered to implement in the Korea Next Generation Reactor (KNGR). In this report, system description on the B and C loop is given in detail, which includes the drawings and technical specification of each component, instrumentation and control system, and the operational procedures and the results of the performance testing. (author). 7 refs., 11 tabs., 48 figs.

  7. LOFT system structural response during subcooled blowdown

    International Nuclear Information System (INIS)

    Martinell, J.S.

    1978-01-01

    The Loss-of-Fluid Test (LOFT) facility is a highly instrumented, pressurized water reactor test system designed to be representative of large pressurized water reactors (LPWRs) for the simulation of loss-of-coolant accidents (LOCAs). Detailed structural analysis and appropriate instrumentation (accelerometers and strain gages) on the LOFT system provided information for evaluation of the structural response of the LOFT facility for loss-of-coolant experiment (LOCE) induced loads. In general, the response of the system during subcooled blowdown was small with typical structural accelerations below 2.0 G's and dynamic strains less than 150 x 10 - 6 m/m. The accelerations measured at the steam generator and simulated steam generator flange exceeded LOCE design values; however, integration of the accelerometer data at these locations yielded displacements which were less than one half of the design values associated with a safe shutdown earthquake (SSE), which assures structural integrity for LOCE loads. The existing measurement system was adequate for evaluation of the LOFT system response during the LOCEs. The conditions affecting blowdown loads during nuclear LOCEs will be nearly the same as those experienced during the nonnuclear LOCEs, and the characteristics of the structural response data in both types of experiments are expected to be the same. The LOFT system is concluded to be adequately designed and further analysis of the LOFT system with structural codes is not required for future LOCE experiments

  8. PIV measurement at the blowdown pipe outlet

    International Nuclear Information System (INIS)

    Puustinen, M.; Laine, J.; Raesaenen, A.; Pyy, L.; Telkkae, J.

    2013-04-01

    This report summarizes the findings of the PIV measurement tests carried out in January - February 2013 with the scaled down PPOOLEX test facility at LUT. The main objective of the tests was to find out the operational limits of the PIV system regarding suitable test conditions and correct values of different adjustable PIV parameters. An additional objective was to gather CFD grade data for verification/validation of numerical models. Both water and steam injection tests were carried out. PIV measurements with cold water injection succeeded well. Raw images were of high quality, averaging over the whole measurement period could be done and flow fields close to the blowdown pipe outlet could be determined. In the warm water injection cases the obtained averaged velocity field images were harder to interpret, especially if the blowdown pipe was also filled with warm water in the beginning of the measurement period. The absolute values of the velocity vectors seemed to be smaller than in the cold water injection cases. With very small steam flow rates the steam/water interface was inside the blowdown pipe and quite stable in nature. The raw images were of good quality but due to some fluctuation in the velocity field averaging of the velocity images over the whole measured period couldn't be done. Condensation of steam in the vicinity of the pipe exit probably caused these fluctuations. A constant outflow was usually followed by a constant inflow towards the pipe exit. Vector field images corresponding to a certain phase of the test could be extracted and averaged but this would require a very careful analysis so that the images could be correctly categorized. With higher steam flow rates rapid condensation of large steam bubbles created small gas bubbles which were in front of the measurement area of the PIV system. They disturbed the measurements by reflecting laser light like seeding particles and therefore the raw images were of poor quality and they couldn't be

  9. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

    The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

  10. Contribution to the theory of the two phase blowdown phenomenon

    International Nuclear Information System (INIS)

    Hutcherson, M.N.

    1975-12-01

    In order to accurately model the two phase portion of a pressure vessel blowdown, it becomes necessary to understand the bubble growth mechanism within the vessel during the early period of the decompression, the two phase flow behavior within the vessel, and the applicability of the available two phase critical flow models to the blowdown transient. To aid in providing answers to such questions, a small scale, separate effects, isothermal blowdown experiment has been conducted in a small pressure vessel. The tests simulated a full open, double ended, guillotine break in a large diameter, short exhaust duct from the vessel. The vaporization process at the initiation of the decompression is apparently that of thermally dominated bubble growth originating from the surface cavities inside the system. Thermodynamic equilibrium of the remaining fluid within the vessel existed in the latter portion of the decompression. A nonuniform distribution of fluid quality within the vessel was also detected in this experiment. By comparison of the experimental results from this and other similar transient, two phase critical flow studies with steady state, small duct, two phase critical flow data, it is shown that transient, two phase critical flow in large ducts appears to be similar to steady state, two phase critical flow in small ducts. Analytical models have been developed to predict the blowdown characteristics of a system during subcooled decompression, the bubble growth regime of blowdown, and also in the nearly dispersed period of depressurization. This analysis indicates that the system pressure history early in the blowdown is dependent on the internal vessel surface area, the internal vessel volume, and also on the exhaust flow area from the system. This analysis also illustrates that the later period of decompression can be predicted based on thermodynamic equilibrium

  11. PPOOLEX experiments with a modified blowdown pipe outlet

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.; Raesaenen, A.

    2009-08-01

    This report summarizes the results of the experiments with a modified blowdown pipe outlet carried out with the PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through a vertical DN200 blowdown pipe to the condensation pool. Four reference experiments with a straight pipe and ten with the Forsmark type collar were carried out. The main purpose of the experiment series was to study the effect of a blowdown pipe outlet collar design on loads caused by chugging phenomena (rapid condensation) while steam is discharged into the condensation pool. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. During the experiments the initial temperature level of the condensation pool water was either 20-25 or 50-55 deg. C. The steam flow rate varied from 400 to 1200 g/s and the temperature of incoming steam from 142 to 185 deg. C. In the experiments with 20-25 deg. C pool water, even 10 times higher pressure pulses were measured inside the blowdown pipe in the case of the straight pipe than with the collar. In this respect, the collar design worked as planned and removed the high pressure spikes from the blowdown pipe. Meanwhile, there seemed to be no suppressing effect on the loads due to the collar in the pool side in this temperature range. Registered loads in the pool were approximately in the same range (or even a little higher) with the collar as with the straight pipe. In the experiments with 50-55 deg. C pool water no high pressure pulses were measured inside the blowdown pipe either with the straight pipe or with the collar. In this case, more of the suppressing effect is probably due to the warmer pool water than due to the modified pipe outlet. It has been observed already in the earlier experiments with a straight pipe in the POOLEX and PPOOLEX facilities that warm pool water has a diminishing effect on

  12. LOFT blowdown loop piping thermal analysis Class I review

    International Nuclear Information System (INIS)

    Kinnaman, T.L.

    1978-01-01

    In accordance with ASME Code, Section III requirements, all analyses of Class I components must be independently reviewed. Since the LOFT blowdown loop piping up through the blowdown valve is a Class I piping system, the thermal analyses are reviewed. The Thermal Analysis Branch comments to this review are also included. It is the opinion of the Thermal Analysis Branch that these comments satisfy all of the reviewers questions and that the analyses should stand as is, without additional considerations in meeting the ASME Code requirements and ANC Specification 60139

  13. 46 CFR 162.018-5 - Blow-down adjustment and popping tolerance.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 6 2010-10-01 2010-10-01 false Blow-down adjustment and popping tolerance. 162.018-5... Compressed Gas § 162.018-5 Blow-down adjustment and popping tolerance. (a) Safety relief valves shall be so... adjustible blow-down construction shall be adjusted to close after blowing down not more than 5 percent of...

  14. Occurrence of critical heat flux during blowdown with flow reversal

    International Nuclear Information System (INIS)

    Leung, J.C.M.

    1976-04-01

    A small-scale experiment using Freon-11 at 130 0 F and 65 psia in a well-instrumented transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 2 ft length. Inlet and exit void fractions were measured by a capacitance technique. Flow regime transition was observed with high speed photography. A 1-hr contact time between Freon-11 and nitrogen at 130 0 F and 60 psig was found to greatly affect the steady-state subcooled boiling initial conditions. Delay in bubble growth was observed in adiabatic blowdown runs. This was caused by the thermodynamic nonequilibrium conditions required for the unstable bubble growth. For the diabatic runs, equilibrium was more closely approached in the test section during the early phase of blowdown. Critical heat flux did not occur immediately during the flow decay in an approximately 60 msec reversal period. The first or early CHF which occurred at about 400 msec was independent of the blowdown volume and did not propagate upward. An annular flow pattern appeared at the onset of this CHF which occurred only at the lower 8 in. of the heated zone

  15. Chemical approaches to zero blowdown operation (TP93-05)

    International Nuclear Information System (INIS)

    Geiger, G.E.; Ogg, J.; Hatch, M.R.

    1993-03-01

    Zero blowdown operation was evaluated at a cooling tower at the Stanford Linear Accelerator Center in an attempt to eliminate cooling water discharge. Testing was performed with and without acid feed for pH control using a state-of-the-art treatment which contained polymer, phosphonate, and azole. Supplemental additional of a proprietary calcium carbonate scale inhibitor was also evaluated

  16. Membrane distillation of industrial cooling tower blowdown water

    Directory of Open Access Journals (Sweden)

    N.E. Koeman-Stein

    2016-06-01

    Full Text Available The potential of membrane distillation for desalination of cooling tower blowdown water (CTBD is investigated. Technical feasibility is tested on laboratory and pilot scale using real cooling tower blowdown water from Dow Benelux in Terneuzen (Netherlands. Two types of membranes, polytetrafluorethylene and polyethylene showed good performance regarding distillate quality and fouling behavior. Concentrating CTBD by a factor 4.5 while maintaining a flux of around 2 l/m2*h was possible with a water recovery of 78% available for reuse. Higher concentration factors lead to severe decrease in flux which was caused by scaling. Membrane distillation could use the thermal energy that would otherwise be discharged of in a cooling tower and function as a heat exchanger. This reduces the need for cooling capacity and could lead to a total reduction of 37% water intake for make-up water, as well as reduced energy and chemicals demands and greenhouse gas emissions.

  17. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described

  18. Blowdown heat transfer and transient boiling transition in BWR's

    International Nuclear Information System (INIS)

    Sozzi, G.L.; Burnette, G.W.

    1977-01-01

    Experimental results from the NRC/EPRI/GE BWR Blowdown Heat Transfer Program are evaluated in terms of bundle local heat transfer performance and in terms of cross-sectional average bundle thermal-hydraulic fluid conditions. The bundle heat transfer performance was generally found to be nucleate boiling below the two-phase mixture level interface with highly dispersed film boiling or steam cooling heat transfer above the interface. Comparisons are presented for predictions of boiling transition (BT) and post BT heat transfer performance during the blowdown phase of the LOCA experiments. These predictions utilize a drift flux void fraction model. The comparisons show very good agreement of both the onset of BT and the post BT heat transfer. 12 references

  19. Blow.MOD2: a program for blowdown transient calculations

    International Nuclear Information System (INIS)

    Doval, A.

    1990-01-01

    The BLOW.MOD2 program has been developed to calculate the blowdown phase in a pressurized vessel after a break/valve is opened. It is a one volume model where break height and flow area are specified. Moody critical flow model was adopted under saturation conditions for flow calculation through the break. Heat transfer from structures and internals have been taken into account. Long term depressurization results and a more complex model are compared satisfactorily. (Author)

  20. Fluid-structure interactions in PWR vessels during blowdown

    International Nuclear Information System (INIS)

    Schumann, U.; Enderle, G.; Katz, F.; Ludwig, A.; Moesinger, H.; Schlechtendahl, E.G.

    1979-01-01

    For analysis of blowdown loadings and dynamic response of PWR vessel internals several computer codes have been developed at Karlsruhe. The goal is to provide advanced codes which permit a 'best estimate' analysis of the deformations and stresses of the internal structures, in particular the core barrel, such that the safety margins can be evaluated. The stresses reach their maxima during the initial subcooled period of the blowdown in which two-phase phenomena are important in the blowdown pipe only. In this period, the computed results with and without fluid-structural interactions show that the coupling between the water in the downcomer and the rather thin elastic core barrel is of dominant importance. Without coupling the core barrel oscillates with much higher frequencies than with coupling. The amplitudes and stresses are about twice as large initially. Later, the decoupled analysis can result in a meaningless overestimation of the structural response. By comparison of computations for incompressible and for compressible fluid with and without coupling we have found that a correct treatment of the fluid-structure coupling is more important than the description of pressure waves. (orig.)

  1. A simple blowdown code for SUPER-SARA loop conditions

    International Nuclear Information System (INIS)

    Fritz, G.

    1981-01-01

    The Super Sara test programme (SSTP) is aimed to study in pile the fuel and cluster behaviour under two types of accident conditions: - the ''Large break loss of coolant'' condition (LB-Loca), - the ''Severe fuel damage'' (SFD) in a boildown caused by a small break. BIVOL was developed for the LB-Loca situation. This code is made for a loop where essentially two volumes define the thermohydraulics during the blowdown. In the SUPERSARA loop these two volumes are represented by the hot leg and cold leg pipings together with the respective upper and lower plenum of the test section

  2. Thermal effects influencing measurements in a supersonic blowdown wind tunnel

    Directory of Open Access Journals (Sweden)

    Vuković Đorđe S.

    2016-01-01

    Full Text Available During a supersonic run of a blowdown wind tunnel, temperature of air in the test section drops which can affect planned measurements. Adverse thermal effects include variations of the Mach and Reynolds numbers, variation of airspeed, condensation of moisture on the model, change of characteristics of the instrumentation in the model, et cetera. Available data on thermal effects on instrumentation are pertaining primarily to long-run-duration wind tunnel facilities. In order to characterize such influences on instrumentation in the models, in short-run-duration blowdown wind tunnels, temperature measurements were made in the wing-panel-balance and main-balance spaces of two wind tunnel models tested in the T-38 wind tunnel. The measurements showed that model-interior temperature in a run increased at the beginning of the run, followed by a slower drop and, at the end of the run, by a large temperature drop. Panel-force balance was affected much more than the main balance. Ways of reducing the unwelcome thermal effects by instrumentation design and test planning are discussed.

  3. Polyethylene encapsulation of simulated blowdown waste for SEG treatability study

    International Nuclear Information System (INIS)

    Kalb, P.D.

    1993-01-01

    The Environmental and Waste Technology Center is a participating subcontractor in the Scientific Ecology Group (SEG) Treatability Study for Westinghouse Savannah River Co.'s Blowdown Waste. This waste will be generated at the Consolidated Incinerator Facility (CIF) and will consist of the neutralized aqueous scrubber solution from the incinerator. Since the facility is designed to burn low-level radioactive, hazardous, and mixed wastes, the blowdown waste will likely be a mixed waste. Polyethylene encapsulation is an improved treatment method that has been developed at BNL over the last 10 years. Polyethylene is an inert, thermoplastic polymer with a melt temperature of 120 C. The BNL process is a modification of standard plastics extrusion technology that has been utilized successfully by the plastics industry for over 50 years. Polyethylene binder and dry waste material are fed through separate calibrated feeders to the extruder, where the materials are thoroughly mixed, heated to a molten condition, and then extruded into a suitable mold. A monolithic solid waste form results on cooling. The objective of the Phase 1 screening effort was to prepare test specimens of CIF surrogate waste encapsulated in polyethylene for leach testing using EPA's Toxicity Characteristic Leaching Procedure (TCLP). BNL received aqueous CIF surrogate from SEG, pretreated the stimulant for processing, and fabricated TCLP test specimens for analysis at an independent laboratory. Laboratory and processing procedures are described in this letter report

  4. PIV measurement at the blowdown pipe outlet. [Particle Image Velocimetry

    Energy Technology Data Exchange (ETDEWEB)

    Puustinen, M.; Laine, J.; Raesaenen, A.; Pyy, L.; Telkkae, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    2013-04-15

    This report summarizes the findings of the PIV measurement tests carried out in January - February 2013 with the scaled down PPOOLEX test facility at LUT. The main objective of the tests was to find out the operational limits of the PIV system regarding suitable test conditions and correct values of different adjustable PIV parameters. An additional objective was to gather CFD grade data for verification/validation of numerical models. Both water and steam injection tests were carried out. PIV measurements with cold water injection succeeded well. Raw images were of high quality, averaging over the whole measurement period could be done and flow fields close to the blowdown pipe outlet could be determined. In the warm water injection cases the obtained averaged velocity field images were harder to interpret, especially if the blowdown pipe was also filled with warm water in the beginning of the measurement period. The absolute values of the velocity vectors seemed to be smaller than in the cold water injection cases. With very small steam flow rates the steam/water interface was inside the blowdown pipe and quite stable in nature. The raw images were of good quality but due to some fluctuation in the velocity field averaging of the velocity images over the whole measured period couldn't be done. Condensation of steam in the vicinity of the pipe exit probably caused these fluctuations. A constant outflow was usually followed by a constant inflow towards the pipe exit. Vector field images corresponding to a certain phase of the test could be extracted and averaged but this would require a very careful analysis so that the images could be correctly categorized. With higher steam flow rates rapid condensation of large steam bubbles created small gas bubbles which were in front of the measurement area of the PIV system. They disturbed the measurements by reflecting laser light like seeding particles and therefore the raw images were of poor quality and they couldn

  5. A review of progress with analysis of blowdown experiments using RELAP-UK

    International Nuclear Information System (INIS)

    Fayers, F.J.

    1975-10-01

    This paper briefly reviews some of the recent work at AEE Winfrith to establish the validity of the RELAP-UK code by comparison with blowdown experiments. Five sources of experimental data have been used which include two of the Edwards' simple pipe blowdown experiments, the LOFT semi-scale Benchmark Problem No. 2, and the Italian and Japanese blowdown rig results. Various difficulties in the comparison between theory and measurements are highlighted and the steps proposed to resolve the problems are indicated. (author)

  6. Transient critical heat flux and blowdown heat-transfer studies

    Energy Technology Data Exchange (ETDEWEB)

    Leung, J.C.

    1980-05-01

    Objective of this study is to give a best-estimate prediction of transient critical heat flux (CHF) during reactor transients and hypothetical accidents. To accomplish this task, a predictional method has been developed. Basically it involves the thermal-hydraulic calculation of the heated core with boundary conditions supplied from experimental measurements. CHF predictions were based on the instantaneous ''local-conditions'' hypothesis, and eight correlations (consisting of round-tube, rod-bundle, and transient correlations) were tested against most recent blowdown heat-transfer test data obtained in major US facilities. The prediction results are summarized in a table in which both CISE and Biasi correlations are found to be capable of predicting the early CHF of approx. 1 s. The Griffith-Zuber correlation is credited for its prediction of the delay CHF that occurs in a more tranquil state with slowly decaying mass velocity. In many instances, the early CHF can be well correlated by the x = 1.0 criterion; this is certainly indicative of an annular-flow dryout-type crisis. The delay CHF occurred at near or above 80% void fraction, and the success of the modified Zuber pool-boiling correlation suggests that this CHF is caused by flooding and pool-boiling type hydrodynamic crisis.

  7. Modelling of blowdown of steam in the pressurized PPOOLEX facility

    International Nuclear Information System (INIS)

    Paettikangas, T.; Niemi, J.; Timperi, A.

    2009-12-01

    PPOOLEX experiment WLL-04-02 on condensation of vapour is studied with CFD simulations. Wall condensation model has been adapted to an Euler-Euler multiphase model of the Fluent CFD code for this purpose. In addition, a simple direct-contact condensation model has also been included in the code. The main focus of the CFD modelling work was on modelling condensation in the drywell. The amount of condensation found in the CFD calculation was in fair agreement with the experiment. The present simulation was so short that the gas flowing into the wetwell contained significant amount of air. The mole fraction of vapour at the outlet of the vent pipe had the maximum value of about 0.3. Therefore, the noncondensable gas strongly affected the direct-contact condensation in the water pool. Much longer simulations are needed in order to study jugging and condensation oscillations. FSI calculations of the experiments were performed by using the Star-CD, ABAQUS and MpCCI codes. An approximate method that makes possible numerically stable FSI calculations for the experimental facilities was used. The method is based on linear perturbation method which necessitates small structural deformations. The calculations showed that FSI has to be taken into account for the POOLEX facility which has relatively light structures. A way for determining the pressure source for the acoustic model from pressure measured at the pool bottom was also examined. Separation of the pressure component due to wall motion from the blowdown load was attempted by conducting a Fourier analysis on the measured displacement signal. The study showed that in practise sufficiently accurate acceleration signal cannot be obtained this way because the transformed signal gets easily out of phase. A measurement system was proposed which could be used for determining the pressure fluctuations. (author)

  8. Modelling of blowdown of steam in the pressurized PPOOLEX facility

    Energy Technology Data Exchange (ETDEWEB)

    Paettikangas, T.; Niemi, J.; Timperi, A. (VTT Technical Research Centre of Finland (Finland))

    2009-12-15

    PPOOLEX experiment WLL-04-02 on condensation of vapour is studied with CFD simulations. Wall condensation model has been adapted to an Euler-Euler multiphase model of the Fluent CFD code for this purpose. In addition, a simple direct-contact condensation model has also been included in the code. The main focus of the CFD modelling work was on modelling condensation in the drywell. The amount of condensation found in the CFD calculation was in fair agreement with the experiment. The present simulation was so short that the gas flowing into the wetwell contained significant amount of air. The mole fraction of vapour at the outlet of the vent pipe had the maximum value of about 0.3. Therefore, the noncondensable gas strongly affected the direct-contact condensation in the water pool. Much longer simulations are needed in order to study jugging and condensation oscillations. FSI calculations of the experiments were performed by using the Star-CD, ABAQUS and MpCCI codes. An approximate method that makes possible numerically stable FSI calculations for the experimental facilities was used. The method is based on linear perturbation method which necessitates small structural deformations. The calculations showed that FSI has to be taken into account for the POOLEX facility which has relatively light structures. A way for determining the pressure source for the acoustic model from pressure measured at the pool bottom was also examined. Separation of the pressure component due to wall motion from the blowdown load was attempted by conducting a Fourier analysis on the measured displacement signal. The study showed that in practise sufficiently accurate acceleration signal cannot be obtained this way because the transformed signal gets easily out of phase. A measurement system was proposed which could be used for determining the pressure fluctuations. (author)

  9. Effect of air content and mass inflow on the pressure rise in a containment during blowdown

    International Nuclear Information System (INIS)

    Marshall, J.; Holland, P.G.

    1977-01-01

    Experiments were made to investigate conditions arising during blowdown of a vessel filled with saturated steam/water at 7 MPa pressure into a containment vessel. The initial air pressure in the containment vessel was varied from one atmosphere to near vacuum. The initial water content of the high pressure vessel was varied. Pressure and temperature distributions were measured during the blowdown transient and compared with calculations based on a simple lumped-parameter model. The effect of condensation heat transfer on the containment pressure is discussed and attention drawn to the inadequacy of most available data. (Author)

  10. Stress analysis of LOFT steam generator blowdown cross-over line

    International Nuclear Information System (INIS)

    Singh, J.N.

    1978-01-01

    The purpose of this report is to demonstrate compliance of the LOFT Steam Generator Blowdown Cross-Over Piping with the ASME Boiler and Pressure Vessel Code, Section III, Subsection NC. Deadweight, thermal expansion, seismic, LOCE, and LOCA loads have been considered. With the addition of two snubbers, as shown in this report, the system conforms to all requirements

  11. Bench-scale treatability studies for simulated incinerator scrubber blowdown containing radioactive cesium and strontium

    International Nuclear Information System (INIS)

    Coroneos, A.C.; Taylor, P.A.; Arnold, W.D. Jr.; Bostick, D.A.; Perona, J.J.

    1994-12-01

    The purpose of this report is to document the results of bench-scale testing completed to remove 137 Cs and 90 Sr from the Oak Ridge K-25 Site Toxic Substances Control Act (TSCA) Incinerator blowdown at the K-25 Site Central Neutralization Facility, a wastewater treatment facility designed to remove heavy metals and uranium from various wastewaters. The report presents results of bench-scale testing using chabazite and clinoptilolite zeolites to remove cesium and strontium; using potassium cobalt ferrocyanide (KCCF) to remove cesium; and using strontium chloride coprecipitation, sodium phosphate coprecipitation, and calcium sulfate coprecipitation to remove strontium. Low-range, average-range, and high-range concentration blowdown surrogates were used to complete the bench-scale testing

  12. Performance of Blowdown Turbine Driven by Exhaust Gas of Nine-Cylinder Radial Engine

    Science.gov (United States)

    Turner, L Richard; Desmon, Leland G

    1944-01-01

    An investigation was made of an exhaust-gas turbine having four separate nozzle boxes each covering a 90 degree arc of the nozzle diaphragm and each connected to a pair of adjacent cylinders of a nine-cylinder radial engine. This type of turbine has been called a "blowdown" turbine because it recovers the kinetic energy developed in the exhaust stacks during the blowdown period, that is the first part of the exhaust process when the piston of the reciprocating engine is nearly stationary. The purpose of the investigation was to determine whether the blow turbine could develop appreciable power without imposing any large loss in engine power arising from restriction of the engine exhaust by the turbine.

  13. Fluid and structural dynamics calculations to determine core barrel loads during blowdown (EV 3,000)

    International Nuclear Information System (INIS)

    Krieg, R.; Schlechtendahl, E.G.

    1977-01-01

    To begin with, the main physical phenomena in connection with blowdown loads on the care barrel and the computer models used are briefly described. These models have also been used in the design of the HTR test care barrel. The fluid dynamics part of the calculations was carried out using the WHAMMOD and DAPSY codes; for the structural dynamics part, the STRUDL/Dynal code was employed. (orig./RW) [de

  14. Singular and interactive effects of blowdown, salvage logging, and wildfire in sub-boreal pine systems

    Science.gov (United States)

    D'Amato, A.W.; Fraver, S.; Palik, B.J.; Bradford, J.B.; Patty, L.

    2011-01-01

    The role of disturbance in structuring vegetation is widely recognized; however, we are only beginning to understand the effects of multiple interacting disturbances on ecosystem recovery and development. Of particular interest is the impact of post-disturbance management interventions, particularly in light of the global controversy surrounding the effects of salvage logging on forest ecosystem recovery. Studies of salvage logging impacts have focused on the effects of post-disturbance salvage logging within the context of a single natural disturbance event. There have been no formal evaluations of how these effects may differ when followed in short sequence by a second, high severity natural disturbance. To evaluate the impact of this management practice within the context of multiple disturbances, we examined the structural and woody plant community responses of sub-boreal Pinus banksiana systems to a rapid sequence of disturbances. Specifically, we compared responses to Blowdown (B), Fire (F), Blowdown-Fire, and Blowdown-Salvage-Fire (BSF) and compared these to undisturbed control (C) stands. Comparisons between BF and BSF indicated that the primary effect of salvage logging was a decrease in the abundance of structural legacies, such as downed woody debris and snags. Both of these compound disturbance sequences (BF and BSF), resulted in similar woody plant communities, largely dominated by Populus tremuloides; however, there was greater homogeneity in community composition in salvage logged areas. Areas experiencing solely fire (F stands) were dominated by P. banksiana regeneration, and blowdown areas (B stands) were largely characterized by regeneration from shade tolerant conifer species. Our results suggest that salvage logging impacts on woody plant communities are diminished when followed by a second high severity disturbance; however, impacts on structural legacies persist. Provisions for the retention of snags, downed logs, and surviving trees as part

  15. A method for calculating the critical time under blowdown conditions during a SBLOCA

    International Nuclear Information System (INIS)

    Su Guanghui; Yu Zhenwan; Guo Yujun; Zhang Jinling; Qiu Suizheng; Jia Dounan

    1994-01-01

    The critical time is the period from the instant at which the blowdown occurs to the instant when the critical heat flux (CHF) happens. It determines the time of operating the safety protection system when a LOCA occurs in PWR. It is important to calculate the critical time correctly. The weakness of Griffith's method is analyzed and a great improvement is developed. The critical time calculated by the improved method agrees with the experimental values

  16. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Pappx, L.

    1994-01-01

    After modification of Dukovany NPP steam generator feedwater system, the increased concentration of minerals was measured in the cold leg of modified steam generator. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators, has focused this attention on the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of flow distribution in the secondary side of SG was developed. (Author)

  17. PPOOLEX experiments on the dynamics of free water surface in the blowdown pipe

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.; Raesaenen, A.

    2013-04-01

    This report summarizes the results of the thermal stratification and mixing experiments carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through the vertical DN200 blowdown pipe to the condensation pool filled with sub-cooled water. The main objective of the experiments was to obtain verification data for the development of the Effective Momentum Source (EMS) and Effective Heat Source (EHS) models to be implemented in GOTHIC code by KTH. A detailed test matrix and procedure put together on the basis of pre-test calculations was provided by KTH before the experiments. Altogether six experiments were carried out. The experiments consisted of a small steam flow rate stratification period and of a higher flow rate mixing period. The dry well structures were heated up to approximately 130 deg. C before the stratification period was initiated. The initial water bulk temperature in the condensation pool was 13-16 deg. C. During the low steam flow rate (85-105 g/s) period steam condensed mainly inside the blowdown pipe. As a result temperatures remained constant below the blowdown pipe outlet while they increased towards the pool surface layers indicating strong thermal stratification of the wet well pool water. In the end of the stratification period the temperature difference between the pool bottom and surface was 15-30 deg. C depending on the test parameters and the duration of the low flow rate period. In the beginning of the mixing phase the steam flow rate was increased rapidly to 300-425 g/s to mix the pool water totally. Depending on the used steam flow rate and initial pool water temperature it took 150-500 s to achieve total mixing. If the test was continued long enough the water pool began to stratify again after the water bulk temperature had reached ∼50 deg. C despite of steam mass flux belonging to the chugging region of the

  18. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Papp, L.

    1995-01-01

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed

  19. PPOOLEX experiments on the dynamics of free water surface in the blowdown pipe

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M.; Raesaenen, A. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    2013-04-15

    This report summarizes the results of the thermal stratification and mixing experiments carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through the vertical DN200 blowdown pipe to the condensation pool filled with sub-cooled water. The main objective of the experiments was to obtain verification data for the development of the Effective Momentum Source (EMS) and Effective Heat Source (EHS) models to be implemented in GOTHIC code by KTH. A detailed test matrix and procedure put together on the basis of pre-test calculations was provided by KTH before the experiments. Altogether six experiments were carried out. The experiments consisted of a small steam flow rate stratification period and of a higher flow rate mixing period. The dry well structures were heated up to approximately 130 deg. C before the stratification period was initiated. The initial water bulk temperature in the condensation pool was 13-16 deg. C. During the low steam flow rate (85-105 g/s) period steam condensed mainly inside the blowdown pipe. As a result temperatures remained constant below the blowdown pipe outlet while they increased towards the pool surface layers indicating strong thermal stratification of the wet well pool water. In the end of the stratification period the temperature difference between the pool bottom and surface was 15-30 deg. C depending on the test parameters and the duration of the low flow rate period. In the beginning of the mixing phase the steam flow rate was increased rapidly to 300-425 g/s to mix the pool water totally. Depending on the used steam flow rate and initial pool water temperature it took 150-500 s to achieve total mixing. If the test was continued long enough the water pool began to stratify again after the water bulk temperature had reached {approx}50 deg. C despite of steam mass flux belonging to the chugging region

  20. Mixture level models in Toshiba and General Electric blowdown experimental analysis

    International Nuclear Information System (INIS)

    Gebrim, A.N.

    1993-01-01

    Three different mixture level tracking methods to vertical flow channels were tested in two Blowdown experiments. The aim of the tests is to observe the Computational efficiency and the agreement of their results with the experimental data. The first method has been used in the system code ATHLET. The second one has been used in the system code developed at BNL. The third one is described in a report but there is no notice that it has been tested. The results show that the first and the third method produce good agreement with the experimental data. The third method need a fine nodalization to yield good results. (C.M.)

  1. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L. [Inst. of Material Engineering, Ostrava (Switzerland)

    1995-12-31

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

  2. Multidimensional analysis of fluid flow in the loft cold leg blowdown pipe during a loss-of-coolant experiment

    International Nuclear Information System (INIS)

    Demmie, P.N.; Hofmann, K.R.

    1979-03-01

    A computer analysis of fluid flow in the Loss-of-Fluid Test (LOFT) cold leg blowdown pipe during a loss-of-coolant experiment (LOCE) was performed using the computer program K-FIX/MOD1. The purpose of this analysis was to evaluate the capability of K-FIX/MOD1 to calculate theoretical fluid quantity distributions in the blowdown pipe during a LOCE for possible application to the analysis of LOFT experimental data, the determination of mass flow, or the development of data reduction models. A rectangular section of a portion of the LOFT blowdown pipe containing measurement Station BL-1 was modeled using time-dependent boundary conditions. Fluid quantities were calculated during a simulation of the first 26 s of LOFT LOCE L1-4. Sensitivity studies were made to determine changes in void fractions and velocities resulting from specific changes in the inflow boundary conditions used for this simulation

  3. PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

    International Nuclear Information System (INIS)

    Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

  4. High enthalpy extraction experiments with Fuji-1 MHD blow-down facility

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Y.; Okamura, T.; Yoshikawa, K. [Tokyo Inst. of Technology, Yokohama (JP)] [and others

    1999-07-01

    Recent experimental results of closed cycle MHD electrical power generation with the Fuji-1 blow-down facility are presented. In the experiment with Disk-F4 MHD generator, which was conducted with a modified seed injection system in 1997, an enthalpy extraction ratio of 18.4% was successfully demonstrated with a large output power of 506 kW. This enthalpy extraction ratio is the highest among those achieved with the Fuji-1 facility. The experimental results also revealed the electrical characteristics of the generator installed in the blow-down facility. The decline in the output power and its recovery were observed at the early stage of the power generation run. This fact could be attributed to the attachment of seed material to the generator walls and to its detachment, related to the relatively slow rise in temperature on the wall surface. It was verified for the first time in the Fuji-1 experiment that the reduction of impurity contamination resulted in improvement in the generator performance. (Author)

  5. Modeling study of droplet behavior during blowdown period of large break LOCA based on experimental data

    International Nuclear Information System (INIS)

    Sakaba, Hiroshi; Umezawa, Shigemitsu; Teramae, Tetsuya; Furukawa, Yuji

    2004-01-01

    During LOCA (Loss Of Coolant Accident) in PWR, droplets behavior during blowdown period is one of the important phenomena. For example, the spattering from falling liquid film that flows from upper plenum generates those droplets in core region. The behavior of droplets in such flow has strong effect for cladding temperature behavior because these droplets are able to remove heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface. For safety analysis of LOCA in PWR, it is necessary to evaluate droplet diameter precisely in order to predict fuel cladding temperature changing by the calculation code. Based on the test results, a new droplet behavior model was developed for the MCOBRA/TRC code that predicts the droplet behavior during such LOCA events. Furthermore, the verification calculations that simulated some blowdown tests were performed using by the MCOBRA/TRAC code. These results indicated the validity of this droplet model during blow down cooling period. The experiment was focused on investigating the Weber number of steady droplet in the blow down phenomenon of large break LOCA. (author)

  6. Development of the Variable Atmosphere Testing Facility for Blow-Down Analysis of the Mars Hopper Prototype

    Energy Technology Data Exchange (ETDEWEB)

    Nathan D. Jerred; Robert C. O' Brien; Steven D. Howe; James E. O' Brien

    2013-02-01

    Recent developments at the Center for Space Nuclear Research (CSNR) on a Martian exploration probe have lead to the assembly of a multi-functional variable atmosphere testing facility (VATF). The VATF has been assembled to perform transient blow-down analysis of a radioisotope thermal rocket (RTR) concept that has been proposed for the Mars Hopper; a long-lived, long-ranged mobile platform for the Martian surface. This study discusses the current state of the VATF as well as recent blow-down testing performed on a laboratory-scale prototype of the Mars Hopper. The VATF allows for the simulation of Mars ambient conditions within the pressure vessel as well as to safely perform blow-down tests through the prototype using CO2 gas; the proposed propellant for the Mars Hopper. Empirical data gathered will lead to a better understanding of CO2 behavior and will provide validation of simulation models. Additionally, the potential of the VATF to test varying propulsion system designs has been recognized. In addition to being able to simulate varying atmospheres and blow-down gases for the RTR, it can be fitted to perform high temperature hydrogen testing of fuel elements for nuclear thermal propulsion.

  7. Combining satellite imagery with forest inventory data to assess damage severity following a major blowdown event in northern Minnesota, USA

    Science.gov (United States)

    Mark D. Nelson; Sean P. Healey; W. Keith Moser; Mark H. Hansen

    2009-01-01

    Effects of a catastrophic blowdown event in northern Minnesota, USA were assessed using field inventory data, aerial sketch maps and satellite image data processed through the North American Forest Dynamics programme. Estimates were produced for forest area and net volume per unit area of live trees pre- and post-disturbance, and for changes in volume per unit area and...

  8. Singular and combined effects of blowdown, salvage logging, and wildfire on forest floor and soil mercury pools

    Science.gov (United States)

    Carl P.J. Mitchell; Randall K. Kolka; Shawn. Fraver

    2012-01-01

    A number of factors influence the amount of mercury (Hg) in forest floors and soils, including deposition, volatile emission, leaching, and disturbances such as fire. Currently the impact on soil Hg pools from other widespread forest disturbances such as blowdown and management practices like salvage logging are unknown. Moreover, ecological and biogeochemical...

  9. Singular and combined effects of blowdown, salvage logging, and wildfire on forest floor and soil mercury pools.

    Science.gov (United States)

    Mitchell, Carl P J; Kolka, Randall K; Fraver, Shawn

    2012-08-07

    A number of factors influence the amount of mercury (Hg) in forest floors and soils, including deposition, volatile emission, leaching, and disturbances such as fire. Currently the impact on soil Hg pools from other widespread forest disturbances such as blowdown and management practices like salvage logging are unknown. Moreover, ecological and biogeochemical responses to disturbances are generally investigated within a single-disturbance context, with little currently known about the impact of multiple disturbances occurring in rapid succession. In this study we capitalize on a combination of blowdown, salvage logging and fire events in the sub-boreal region of northern Minnesota to assess both the singular and combined effects of these disturbances on forest floor and soil total Hg concentrations and pools. Although none of the disturbance combinations affected Hg in mineral soil, we did observe significant effects on both Hg concentrations and pools in the forest floor. Blowdown increased the mean Hg pool in the forest floor by 0.76 mg Hg m(-2) (223%). Salvage logging following blowdown created conditions leading to a significantly more severe forest floor burn during wildfire, which significantly enhanced Hg emission. This sequence of combined events resulted in a mean loss of approximately 0.42 mg Hg m(-2) (68% of pool) from the forest floor, after conservatively accounting for potential losses via enhanced soil leaching and volatile emissions between the disturbance and sampling dates. Fire alone or blowdown followed by fire did not significantly affect the total Hg concentrations or pools in the forest floor. Overall, unexpected consequences for soil Hg accumulation and by extension, atmospheric Hg emission and risk to aquatic biota, may result when combined impacts are considered in addition to singular forest floor and soil disturbances.

  10. A simplified dynamic analysis for reactor piping systems under blowdown conditions

    International Nuclear Information System (INIS)

    Chen, M.M.

    1975-01-01

    In the design of pipelines in a nuclear power plant for blowdown conditions, is it customary to conduct dynamic analysis of the piping system to obtain the responses and the resulting stresses. Calculations are repeated for each design modification in piping geometry or supporting system until the design codes are met. The numerical calculations are, in general, very costly and time consuming. Until now, there have been no simple means for calculating the dynamic responses for the design. The proposed method reduces the dynamic calculation to a quasi-static one, and can be beneficially used for the preliminary design. The method is followed by a complete dynamical analysis to improve the final results. The new formulations greatly simplify the numerical computation and provide design guides. When used to design a given piping system, the method saved approximately one order of magnitude of computer time. The approach can also be used for other types of structures

  11. Preliminary condensation pool experiments with steam using DN80 and DN100 blowdown pipes

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2004-03-01

    The report summarizes the results of the preliminary steam blowdown experiments. Altogether eight experiment series, each consisting of several steam blows, were carried out in autumn 2003 with a scaled-down condensation pool test rig designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to evaluate the capabilities of the test rig and the needs for measurement and visualization devices. The experiments showed that a high-speed video camera is essential for visual observation due to the rapid condensation of steam bubbles. Furthermore, the maximum measurement frequency of the current combination of instrumentation and data acquisition system is inadequate for the actual steam tests in 2004. (au)

  12. LOFT transient thermal analysis for 10 inch primary coolant blowdown piping weld

    International Nuclear Information System (INIS)

    Howell, S.K.

    1978-01-01

    A flaw in a weld in the 10 inch primary coolant blowdown piping was discovered by LOFT personnel. As a result of this, a thermal analysis and fracture mechanics analysis was requested by LOFT personnel. The weld and pipe section were analyzed for a complete thermal cycle, heatup and Loss of Coolant Experiment (LOCE), using COUPLE/MOD2, a two-dimensional finite element heat conduction code. The finite element representation used in this analysis was generated by the Applied Mechanics Branch. The record of nodal temperatures for the entire transient was written on tape VSN=T9N054, and has been forwarded to the Applied Mechanics Branch for use in their mechanical analysis. Specific details and assumptions used in this analysis are found in appropriate sections of this report

  13. Fracture mechanics evaluation of some LOFT blowdown system and primary coolant coldleg welds

    International Nuclear Information System (INIS)

    Nagata, P.K.

    1978-01-01

    Fracture mechanics evaluations were performed for three welds in the LOFT blowdown system and one weld in the LOFT primary coolant system. Because the applied stress is not known, a sensitivity analysis was run. The assumed initial defect size was one that had a small probability of being missed; applied stresses of 68.9, 137.8, 206.7, and 344.8 MPa were used. It was found that at the lowest stress (68.9 MPa or 10 ksi) the number of cycles from the initial size to rupture was over 6 x 10 6 . The current calculations indicate that with the worst crack configuration--depth-to-length ratio (a/2c) of 0.10--about 1000 cycles with a peak stress of 227.5 MPa (33 ksi) will be needed to propagate the 0.5 x 5.1 cm (0.2 x 2.0 in.) crack to failure

  14. Guidelines for selecting weak-base versus strong-base anion-exchange resins for the recovery of chromate from cooling tower blowdown

    Energy Technology Data Exchange (ETDEWEB)

    Newman, J.; Reed, L.W.

    1980-01-01

    Guidelines for selecting weak-base versus strong-base anion-exchange resins for the recovery of chromate from cooling tower blowdown are given, together with actual operating data on large-scale industrial systems based on strong-base anion-exchange resins, data from a similar pilot system based on weak-base anion resin, and the chemical costs for operating both systems for a cooling tower blowdown containing 2500 ppm total dissolved solids and 20 ppm chromata.

  15. Project description: ORNL PWR blowdown heat transfer separate-effects program, Thermal-Hydraulic Test Facility (THTF)

    International Nuclear Information System (INIS)

    1976-02-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results will be obtained from the Thermal-Hydraulic Test Facility (THTF), a large nonnuclear pressurized-water loop that incorporates a 49-rod electrically heated bundle. Supporting experiments will be carried out in two additional test loops - the Forced Convection Test Facility (FCTF), a small high-pressure facility in which single heater rods can be tested in annular geometry; and an air-water loop which is used to evaluate two-phase flow-measuring instrumentation

  16. Calculation of sample problems related to two-phase flow blowdown transients in pressure relief piping of a PWR pressurizer

    International Nuclear Information System (INIS)

    Shin, Y.W.; Wiedermann, A.H.

    1984-02-01

    A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients

  17. PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 103

    Energy Technology Data Exchange (ETDEWEB)

    Clemons, V.D.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-03-07

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 103, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system.

  18. Debris transport evaluation during the blow-down phase of a LOCA using computational fluid dynamics

    International Nuclear Information System (INIS)

    Park, Jong Pil; Jeong, Ji Hwan; Kim, Won Tae; Kim, Man Woong; Park, Ju Yeop

    2011-01-01

    Highlights: → We conducted CFD simulation on the spreading of the coolant in the containment after a break of the hot leg. It is used to estimate the dispersion of the debris within the containment. → It was assumed that the small and fine debris is transported by the discharge flow so that a fraction of the small and fine debris transport can be estimated based on the amount of water. → The break flow was assumed to be a homogeneous two-phase mixture without phase separation. Isenthalpic expansion of the break flow was used to specify the inlet boundary condition of the break flow. → The fraction of the small and fine debris transported to the upper part is 73%; this value is close to the value calculated using 1D lumped-parameter codes by the USNRC and the KINS, respectively, while 48% more than the value shown in the NEI 04-07. - Abstract: The performance of the emergency recirculation water sump under the influence of debris accumulation following a loss-of-coolant accident (LOCA) has long been of safety concern. Debris generation and transport during a LOCA are significantly influenced by the characteristics of the ejected coolant flow. One-dimensional analyses previously have been attempted to evaluate the debris transport during the blow-down phase but the transport evaluation still has large uncertainties. In this work, a computational fluid dynamics (CFD) analysis was utilized to evaluate small and fine debris transport during the blow-down phase of a pressurized water reactor, OPR1000. The coolant ejected from the ruptured hot-leg was assumed to expand in an isenthalpic process. The transport of small and fine debris was assumed to be dominated by water-borne transport, and the transport fractions for the upper and lower parts of the containment were quantified based on the CFD analysis. It was estimated that 73% of small and fine debris is transported to the upper part of the containment. This value is close to the values estimated by nuclear

  19. LOFT L2-3 scoping blowdown experiment safety analyses A, B, C, CD, CE, and CF

    International Nuclear Information System (INIS)

    Perryman, J.L.; Keeler, C.D.; Good, E.G.; Saukkoriipi, L.O.

    1978-01-01

    The consequences of various postulated single failures during loss-of-fluid test (LOFT) Loss-of-Coolant Experiment (LOCE) L2-3 were analyzed for the LOCE L2-3 blowdown scoping experiment safety analysis (ESA). The following six of the LOCE L2-3 blowdown scoping ESA were performed: (1) Analysis A--failure of high-pressure injection system (HPIS) A; (2) Analysis B--failure of accumulator A; (3) Analysis C--failure of low-pressure injection system (LPIS) A, assuming accumulator A initial conditions of 550 psig and 84 0 F; (4) Analysis CD--failure of LPIS A, assuming accumulator A initial conditions of 550 psig and 150 0 F; (5) Analysis CE--failure of LPIS A, assuming accumulator A initial conditions of 650 psig and 150 0 F; and (6) Analysis CF--failure of LPIS A, assuming accumulator A initial conditions of 650 psig and 115 0 F. RELAP4/MOD5 models with conservative off-nominal initial conditions, and evaluation model (EM) options were used during the blowdown-refill phase of the analyses. RELAP4/MOD6 models with conservative input parameters and code options selected were used for the reflood phase of the analyses

  20. Further development of drag bodies for the measurement of mass flow rates during blowdown experiments

    International Nuclear Information System (INIS)

    Brockmann, E.; John, H.; Reimann, J.

    1983-01-01

    Drag bodies have already been used for sometime for the measurement of mass flow rates in blowdown experiments. Former research concerning the drag body behaviour in non-homogeneous two-phase flows frequently dealt with special effects by means of theoretical models only. For pipe flows most investigations were conducted for ratios of drag plate area to pipe cross section smaller 0.02. The present paper gives the results of experiments with drag bodies in a horizontal, non-homogeneous two-phase pipe flow with slip, which were carried through under the sponsorship of the German Ministry for Research and Technology (BMFT). Special interest was layed on the behaviour of the drag coefficient in stationary flows and at various cross sectional ratios. Both design and response of various drag bodies, which were developed at the Battelle-Institut, were tested in stationary and instationary two-phase flows. The influences of density and velocity profiles as well as the drag body position were studied. The results demonstrate, that the drag body is capable of measuring mass flow rates in connection with a gamma densitometer also in non-homogeneous two-phase flows. Satisfying results could be obtained, using simply the drag coefficient which was determined from single-phase flow calibrations

  1. Thermo-hydraulic consequence of pressure suppression containment vessel during blowdown, 2

    International Nuclear Information System (INIS)

    Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki

    1980-01-01

    As a part of the safety research works for the integral-type marine reactor, an analytical code SUPPAC-2V was developed to simulate the thermo-hydraulic consequence of a pressure suppression containment system during blowdown and the code was applied to the Model Experimental Facility of the Safety of Integral Type Marine Reactors (explained already in Part 1). SUPPAC-2V is much different from existing codes in the following points. A nonhomogeneous model for the gaseous region in the drywell, a new correlation for condensing heat transfer coefficient at drywell wall based on existing data and approximation of air bubbles in wetwell water by one dimensional bubble rising model are adopted in this code. In comparing calculational results with experimental results, values of predominant input parameters were evaluated and discussed. Moreover, the new code was applied also to the NSR-7 marine reactor, conceptually designed at the Shipbuilding Research Association in Japan, of which suppression system had been already analysed by CONTEMPT-PS. (author)

  2. Steam generator blowdown heat exchangers degradations operational experience on EDF French NPP fleet

    International Nuclear Information System (INIS)

    Praud, M.; Doyen, F.; Wintergest, M.; Jourdain, W.; Roussillon, M.; Zidane, A.; Mayos, M.

    2015-01-01

    The main function of the Steam Generator Blowdown System (SGBS) is to purify the secondary fluid from all kinds of pollutions: corrosion products from the secondary system, consequences of raw water pollutions through condenser's leakage, potential radiochemical pollutions resulting from Primary-to-Secondary leaks. The topic of this paper is to present the main SGBS dysfunctions linked to the degradation of the tubular heat exchangers, which sometimes can lead to integrity failure, through corrosion phenomenon. The degradation mechanisms have been characterized by various visual inspections and destructive examinations performed on pulled tubes and bundles. It appears that SGBS tubes suffer two main forms of corrosion. First, for the non-regenerative heat exchangers, where external surface of tubes is exposed to intermediate fluid, alkaline corrosion under tube sheet or shell-side baffles may occur. Caustic attack results from Na 3 PO 4 decomposition by thermal or chemical process. Secondly, mainly for regenerative heat exchangers, pitting and under-deposits corrosion linked to lay-up conditions during outages. This kind of attack is the root cause of a potential 'domino effect': a steam jet from the leaking tube can induce mechanical and/or erosion on many tubes located in its vicinity. Concerning the external degradation by caustic corrosion, only design modifications and strong monitoring of the raw water inlet may able to limit the occurrence of tube perforation. The lay-up guidelines should be carefully followed to mitigate internal corrosion: a controlled atmosphere (limited humidity) and cleanliness of the tube (avoiding deposits formation on the bottom line) seem to be the main parameters

  3. Effect of the BBR reactor venting system on the thermohydraulic behaviour of the overall system during blowdown

    International Nuclear Information System (INIS)

    Klenke, H.

    1975-01-01

    It was shown that the venting systems used in the BBR-DWR facilities are very effective in meeting the problems of 'steam bindings' which arise in connection with ruptures on the cold side. In double-end ruptures, their influence is insignificant during the blow-down phase whereas in the re-flood phase the core flooding rate is considerably increased. In small ruptures, the water volume in the core is already increased by the venting systems during the 'pool boiling' phase, i.e. after interruption of the forced circulation. (orig./AK) [de

  4. Application of signature analysis for determining the operational readiness of motor-operated valves under blowdown test conditions

    International Nuclear Information System (INIS)

    Haynes, H.D.

    1988-01-01

    In support of the NRC-funded Nuclear Plant Aging Research (NPAR) program, Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of Motor-Operated Valves (MOVs). As part of this work, ORNL participated in the Gate Valve Flow Interruption Blowdown (GVFIB) tests carried out in Huntsville, Alabama. The GVFIB tests were intended primarily to determine the behavior of motor-operated gate valves under the temperature, pressure, and flow conditions expected to be experienced by isolation valves in Boiling Water Reactors (BWRs) during a high energy line break (blowdown) outside of containment. In addition, the tests provided an excellent opportunity to evaluate signature analysis methods for determining the operational readiness of the MOVs under those accident conditions. ORNL acquired motor current and torque switch shaft angular position data on two test MOVs during various times of the GVFIB tests. The reduction in operating ''margin'' of both MOVs due to the presence of additional valve running loads imposed by high flow was clearly observed in motor current and torque switch angular position signatures. In addition, the effects of differential pressure, fluid temperature, and line voltage on MOV operations were observed and more clearly understood as a result of utilizing signature analysis techniques. 1 ref.; 16 figs

  5. The CNEN Helium-Caesium Blow-Down MPD Facility and Experiments with a Prototype Duct

    International Nuclear Information System (INIS)

    Bertolini, E.; Toschi, R.; Lindley, B.C.; Brown, R.; McNab, I.R.

    1966-01-01

    The CNEN blow-down loop has been designed to study a helium-caesium MPD generator with particular regard to non-equilibrium ionization effects. An operating condition of the loop is: gas mass flow 0.2 kg/sec, seed fraction 1 at, wt.%, useful pulse duration 20 sec, stagnation temperature 2000°K, stagnation pressure 5 atm abs, thermal power 1.6 MW, Mach number 0.6, magnetic field 4 Wb/m2, total impurity level less than 100 ppm. A sufficiently wide range of the stagnation conditions can be obtained with the present arrangement of the loop (temperature up to 2000*K, pressure from slightly sub-atmospheric to 6atmabs, gas mass flow from 50 g/sec to 400 g/sec, seed fraction from 0.1 to 2 at. wt.%. The storage heater is an alumina pebble bed electrically heated with tungsten elements and thermally insulated with zirconia fibre; the thermal capacity at 2000°K is about 1000 MJ. Pure helium is obtained by evaporation of liquid helium at between 4.5 and 5°K; liquid caesium is injected into a limited section of the pebble bed to provide a mixture of the two gases uniform in density and temperature. The duct is made of boron nitride (5 cm x 3 cm x 22 cm) with 25 pairs of tantalum electrodes whose geometry (electrode width 3 mm, segmentation pitch 9 mm) should prevent current leakage between adjacent electrodes; the duct walls and transfer can be pre-heated up to 1700°K. A magnetic field of 4 Wb/m 2 is obtained with a pulsed cryogenic magnet with pulse duration of 6 sec. Two series of experiments have been completed to assess the feasibility of the helium-caesium heating system and the generator duct. Heating system experiments, (a) Compatibility of alumina with tungsten, tantalum and caesium, with thermal cycling at 2000°K; (b) Purification of zirconia fibre and its behaviour at high temperature, with thermal cycling at 2000°K; (c) Capability of an alumina pebble bed of evaporating, heating and mixing caesium with flowing helium at 2000°K, with blow-down tests. Generator

  6. Steam quality determination by self-tracers present in the BOP blowdown water and steam

    International Nuclear Information System (INIS)

    Rodriguez Ivanna; Chocron Mauricio; Miceli Fiorella; Saucedo, Ramona

    2012-09-01

    Steam quality determination (the ratio between steam mass flow rate and the sum of steam and liquid water mass flow rate) or its reciprocal named moisture carryover, is a magnitude of importance in fossil fired and nuclear power plants as well. This is due to that, the steam quality participates in the determination of the power transferred from the primary to the secondary circuit (ASME PTC, Gross Heat Input) and the performance of the secondary circuit, in the efficiency of the liquid separators located in the dome of the recirculating type steam generators and finally because the drops carryover implies mechanical wear in the turbine blades and transport of impurities as well. It is after the above mentioned reasons, that several standardized procedures exist (ASTM) and international institutions devoted to the properties of water and steam and applications in power plants release recommendations on the steam quality (IAPWS). Even though, the measurement is still a subject of new publications (Thomas et al., NPC10). In general, the determination methods make use of the addition of a tracer, stable alkaline element or isotope, which has to be later quantified by an analytical or radiochemical technique. It also means keeping the BOP under specified conditions during the test. Chemicals dosing of is not always accepted considering that ions used as tracers concentrate in the steam generator media and modify the water chemistry conditions. This is more pronounced in old devices with presence of fouling and sludge piles on the tube sheet. In the present work a technique based on the concentration of the ions currently existing in the cycle: Blowdown Water (BDW) and condensate (MSR-MSC) of the main steam used as heating fluid in the Moisture Separator Reheater (MSR). The latter ensures a total representative sample of the two phase stream, unlike sampling in the main steam sampling line. Those ion concentrations participate in the calculation of the steam quality. To

  7. Water Hammer Analysis using RELAP5/MOD 3.3 for Yonggwang Nuclear Power Unit 1 and 2 Blowdown System

    International Nuclear Information System (INIS)

    Lee, Sang Il; Kim, Hea Zoo; Chu, Jung Ho; Ahn, Se Hong; Jung, Chang Ho

    2010-01-01

    Water hammer can be defined as a rapid pressure step occurring in the liquid in a closed pipe caused by a sudden change in the liquid velocity. This pressure acts for a period which is twice the transit time of sonic wave in the pipe. Generally, water hammer can occur in any thermal-hydraulic systems like nuclear power plant and is extremely dangerous for nuclear power plant piping system since, if the pressure induced exceeds the pressure range of the pipe given by the manufacturer, it can lead to the failure of the piping system integrity. For Yonggwang nuclear power unit 1 and 2, water hammer occurred repeatedly on the outlet piping of regenerative heat exchanger of steam generator blowdown system. Thus, design modification was performed to prevent the water hammer and the analysis of effect on water hammer before and after design modification was performed to verify the validity of the design modification

  8. Two-dimensional numerical experiments with DRIX-2D on two-phase-water-flows referring to the HDR-blowdown-experiments

    International Nuclear Information System (INIS)

    Moesinger, H.

    1979-08-01

    The computer program DRIX-2D has been developed from SOLA-DF. The essential elements of the program structure are described. In order to verify DRIX-2D an Edwards-Blowdown-Experiment is calculated and other numerical results are compared with steady state experiments and models. Numerical experiments on transient two-phase flow, occurring in the broken pipe of a PWR in the case of a hypothetic LOCA, are performed. The essential results of the two-dimensional calculations are: 1. The appearance of a radial profile of void-fraction, velocity, sound speed and mass flow-rate inside the blowdown nozzle. The reason for this is the flow contraction at the nozzle inlet leading to more vapour production in the vicinity of the pipe wall. 2. A comparison between modelling in axisymmetric and Cartesian coordinates and calculations with and without the core barrel show the following: a) The three-dimensional flow pattern at the nozzle inlet is poorly described using Cartesian coordinates. In consequence a considerable difference in pressure history results. b) The core barrel alters the reflection behaviour of the pressure waves oscillating in the blowdown-nozzle. Therefore, the core barrel should be modelled as a wall normal to the nozzle axis. (orig./HP) [de

  9. Validation of the CATHENA channel model for the post blowdown analysis for the CS28-1 experiment, II - transient

    International Nuclear Information System (INIS)

    Rhee, B.W.; Park, J.H.

    2006-01-01

    To form a licensing bases for the new methodology of fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis has been developed, and tested for a high temperature thermal-chemical experiment CS28-1. Pursuant to the objective of this study the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain good accuracy in the temperature and H 2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results for the FESs of three fuel rings and the pressure tube were quite good as proven in the Figs. 3∼6. However this raises a question how the transient FES and pressure tube temperature can be predicted so well in spite of the insufficient justification of using the 'non-participating medium assumption' for the CO 2 gas gap. Through this study, it was found that the radiation heat transfer model of CATHENA among FES of three rings and the pressure tube as well as the exothermic metal-water reaction model based on the Urbanic-Heidrick correlation are quite accurate and sound. Also it was found that an accurate prediction of the initial condition of the experiment is very important for the accurate prediction of the whole transient as it serves as the starting point of the transient. (author)

  10. Experimental investigation of void distribution in Suppression Pool during the initial blowdown period of a Loss of Coolant Accident using air–water two-phase mixture

    International Nuclear Information System (INIS)

    Rassame, Somboon; Griffiths, Matthew; Yang, Jun; Lee, Doo Yong; Ju, Peng; Choi, Sung Won; Hibiki, Takashi; Ishii, Mamoru

    2014-01-01

    Highlights: • Basic understanding of the venting phenomena in the SP during a LOCA was obtained. • A series of experiment is carried out using the PUMA-E test facility. • Two phases of experiments, namely, an initial and a quasi-steady phase were observed. • The maximum void penetration depth was experienced during the initial phase. - Abstract: During the initial blowdown period of a Loss of Coolant Accident (LOCA), the non-condensable gas initially contained in the BWR containment is discharged to the pressure suppression chamber through the blowdown pipes. The performance of Emergency Core Cooling System (ECCS) can be degraded due to the released gas ingestion into the suction intakes of the ECCS pumps. The understanding of the relevant phenomena in the pressure suppression chamber is important in analyzing potential gas intrusion into the suction intakes of ECCS pumps. To obtain the basic understanding of the relevant phenomena and the generic data of void distribution in the pressure suppression chamber during the initial blowdown period of a LOCA, tests with various blowdown conditions were conducted using the existing Suppression Pool (SP) tank of the integral test facility, called Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility, a scaled downcomer pipe installed in the PUMA-E SP, and air discharge pipe system. Two different diameter sizes of air injection pipe (0.076 and 0.102 m), a range of air volumetric flux (7.9–24.7 m/s), initial void conditions in an air injection pipe (fully void, partially void, and fully filled with water) and different air velocity ramp rates (1.0, 1.5, and 2.0 s) are used to investigate the impact of the blowdown conditions to the void distribution in the SP. Two distinct phases of experiments, namely, an initial and a quasi-steady phase were observed. The maximum void penetration depth was experienced during the initial phase. The quasi-steady phase provided less void

  11. Evaluation of SG blowdown demineralizer performance by replacement of ammonia with ethanolamine as a PWR secondary pH control agent

    International Nuclear Information System (INIS)

    Rhee, I.H.; Yim, S.J.

    2002-01-01

    Four Korean PWR plants have adopted ethanolamine (ETA) as a secondary pH control agent to increase the pH at the liquid phase, which raises the pH in the SG blowdown system. The run time of the SG blowdown demineralizer can be reduced by the increased number of ionic chemical species primarily due to ETA. Contrary to the possible prevention of SG degradation, the replacement of ammonia with ETA results in the water chemistry difficulties and more frequent generation of spent resin. A comprehensive experimental data set for binary, ternary, quaternary, and quinary cation and anion adsorption was developed from small-volume batch tests at total cation or anion concentrations of 0.01 and 0.05 N to obtain the selectivity coefficients of many cations and anions normally present in the PWR secondary system water. In addition, the kinetic study using the bench-scale column was performed to examine the breakthrough point of an ion and to calculate the ratio of inlet to outlet concentration at the column, so called Decontamination Factor, in the different background electrolyte. The batch equilibrium tests indicated that the ion selectivity is higher for an ion of higher valence and is not uniform in the different composition and ionic strength. The preference of an ion on ion exchange resin rather tends to be lower with higher ionic strength. The leakage of an ion from the ion exchange column is not also uniform in time in the various composition and total concentration. Therefore the ion selectivity and breakthrough time are different in ammonia and ethanolamine background electrolytes. The run time of SG blowdown demineralizer can be shorter than it can be expected due to the elevated ionic strength as well as the increased dissolved solids. The quantitative run time can be estimated by such ion exchange models as semi-empirical mass action and surface complexation models. The demineralizer can be used longer by increasing the ratios of cation to anion exchange resins in

  12. Evaluation of SG blowdown demineralizer performance by replacement of ammonia with ethanolamine as a PWR secondary pH control agent

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H. [Department of Materials and Chemical Engineering, Soonchunhyang Univ. (Korea, Republic of); Yim, S.J. [Operation Management Team, Korea Hydro and Nuclear Power Co. Ltd., Seoul (Korea, Republic of)

    2002-07-01

    Four Korean PWR plants have adopted ethanolamine (ETA) as a secondary pH control agent to increase the pH at the liquid phase, which raises the pH in the SG blowdown system. The run time of the SG blowdown demineralizer can be reduced by the increased number of ionic chemical species primarily due to ETA. Contrary to the possible prevention of SG degradation, the replacement of ammonia with ETA results in the water chemistry difficulties and more frequent generation of spent resin. A comprehensive experimental data set for binary, ternary, quaternary, and quinary cation and anion adsorption was developed from small-volume batch tests at total cation or anion concentrations of 0.01 and 0.05 N to obtain the selectivity coefficients of many cations and anions normally present in the PWR secondary system water. In addition, the kinetic study using the bench-scale column was performed to examine the breakthrough point of an ion and to calculate the ratio of inlet to outlet concentration at the column, so called Decontamination Factor, in the different background electrolyte. The batch equilibrium tests indicated that the ion selectivity is higher for an ion of higher valence and is not uniform in the different composition and ionic strength. The preference of an ion on ion exchange resin rather tends to be lower with higher ionic strength. The leakage of an ion from the ion exchange column is not also uniform in time in the various composition and total concentration. Therefore the ion selectivity and breakthrough time are different in ammonia and ethanolamine background electrolytes. The run time of SG blowdown demineralizer can be shorter than it can be expected due to the elevated ionic strength as well as the increased dissolved solids. The quantitative run time can be estimated by such ion exchange models as semi-empirical mass action and surface complexation models. The demineralizer can be used longer by increasing the ratios of cation to anion exchange resins in

  13. Lead sulfate nano- and microparticles in the acid plant blow-down generated at the sulfuric acid plant of the El Teniente mine, Chile.

    Science.gov (United States)

    Barassi, Giancarlo M; Klimsa, Martin; Borrmann, Thomas; Cairns, Mathew J; Kinkel, Joachim; Valenzuela, Fernando

    2014-12-01

    The acid plant 'blow-down' (also called weak acid) produced at El Teniente mine in Chile was characterized. This liquid waste (tailing) is generated during the cooling and cleaning of the smelter gas prior to the production of sulfuric acid. The weak acid was composed of a liquid and a solid phase (suspended solids). The liquid phase of the sample analyzed in this study mainly contained Cu (562 mg L(-1)), SO4(2-) (32 800 mg L(-1)), Ca (1449 mg L(-1)), Fe (185 mg L(-1)), As (6 mg L(-1)), K (467 mg L(-1)) and Al (113 mg L(-1)). Additionally, the sample had a pH-value and total acidity of 0.45 and 2970 mg L(-1) as CaCO3, respectively. Hence, this waste was classified as extremely acidic and with a high metal content following the Ficklin diagram classification. Elemental analysis using atomic absorption, inductively coupled plasma, X-ray diffraction and electron microscopy showed that the suspended solids were anglesite (PbSO4) nano- and microparticles ranging from 50 nm to 500 nm in diameter.

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

    Science.gov (United States)

    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

  15. Mathematical aspects of reactor blowdown

    International Nuclear Information System (INIS)

    Esposito, V.J.

    1975-01-01

    To simulate a hypothetical loss of coolant accident, a large number of equations describing various thermal-hydraulic phenomena must be solved. A review is presented of some of the existing computational methods used for this simulation. A summary of techniques (multi-dimensional) being considered for more detailed investigation is included. (28 references) (U.S.)

  16. Development of a test facility for analyzing supercritical fluid blowdown

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M., E-mail: thiagodbtr@gmail.com [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Silva, Mario A.B. da, E-mail: mabs500@gmail.com [Universidade Federal de Pernambuco (CTG/UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The generation IV nuclear reactors under development mostly use supercritical fluids as the working fluid because higher temperatures improve the thermal efficiency. Supercritical fluids are used by modern nuclear power plants to achieve thermal efficiencies of around 45%. With water as the supercritical working fluid, these plants operate at a high temperature and pressure. However, experiments on supercritical water are limited by technical and financial difficulties. These difficulties can be overcome by using model fluids, which have more feasible supercritical conditions and exhibit a lower critical pressure and temperature. Experimental research is normally used to determine the conditions under which model fluids represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine model fluids that can represent supercritical fluids in a transient state. This paper presents an application of fractional scale analysis to determine the simulation parameters for a depressurization test facility. Carbon dioxide (CO{sub 2}) and R134a gas were considered as the model fluids because their critical point conditions are more feasible than those of water. The similarities of water (prototype), CO{sub 2} (model) and R134a (model) for depressurization in a pressure vessel were analyzed. (author)

  17. Development of a test facility for analyzing supercritical fluid blowdown

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2015-01-01

    The generation IV nuclear reactors under development mostly use supercritical fluids as the working fluid because higher temperatures improve the thermal efficiency. Supercritical fluids are used by modern nuclear power plants to achieve thermal efficiencies of around 45%. With water as the supercritical working fluid, these plants operate at a high temperature and pressure. However, experiments on supercritical water are limited by technical and financial difficulties. These difficulties can be overcome by using model fluids, which have more feasible supercritical conditions and exhibit a lower critical pressure and temperature. Experimental research is normally used to determine the conditions under which model fluids represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine model fluids that can represent supercritical fluids in a transient state. This paper presents an application of fractional scale analysis to determine the simulation parameters for a depressurization test facility. Carbon dioxide (CO 2 ) and R134a gas were considered as the model fluids because their critical point conditions are more feasible than those of water. The similarities of water (prototype), CO 2 (model) and R134a (model) for depressurization in a pressure vessel were analyzed. (author)

  18. A substructure method to compute the 3D fluid-structure interaction during blowdown

    International Nuclear Information System (INIS)

    Guilbaud, D.; Axisa, F.; Gantenbein, F.; Gibert, R.J.

    1983-08-01

    The waves generated by a sudden rupture of a PWR primary pipe have an important mechanical effect on the internal structures of the vessel. This fluid-structure interaction has a strong 3D aspect. 3D finite element explicit methods can be applied. These methods take into account the non linearities of the problem but the calculation is heavy and expensive. We describe in this paper another type of method based on a substructure procedure: the vessel, internals and contained fluid are axisymmetrically described (AQUAMODE computer code). The pipes and contained fluid are monodimensionaly described (TEDEL-FLUIDE Computer Code). These substructures are characterized by their natural modes. Then, they are connected to another (connection of both structural and fluid nodes) the TRISTANA Computer Code. This method allows to compute correctly and cheaply the 3D fluid-structure effects. The treatment of certain non linearities is difficult because of the modal characterization of the substructures. However variations of contact conditions versus time can be introduced. We present here some validation tests and comparison with experimental results of the litterature

  19. Code-to-code comparison for blowdown transients at supercritical conditions

    Energy Technology Data Exchange (ETDEWEB)

    Manera, Annalisa [Paul Scherrer Institute (Switzerland); Antoni, Olivier [CEA Grenoble (France)

    2008-07-01

    The supercritical water reactor is one of the designs selected for further evaluation by the Gen-IV International Forum. In the framework of the EU FP6 project HPLWR-2 (High Performance Light Water Reactor - Phase 2 [1]) a concept of the supercritical water reactor is developed. For the safety assessment of new HPLWR concepts, the capabilities of thermal-hydraulic codes to cope with water at supercritical conditions and especially with the transition from supercritical to sub-critical conditions are of crucial importance. The CATHARE 2 and the RELAP5 codes are foreseen to be employed in the HPLWR-2 project, in order to perform safety analyses for the HPLWR design. Therefore, tests have been carried out to assess the capabilities of the selected system codes. (orig.)

  20. Blowdown mass flow measurements during the Power Burst Facility LOC-11C test

    International Nuclear Information System (INIS)

    Broughton, J.M.; MacDonald, P.E.

    1979-01-01

    An interpretation and evaluation of the two-phase coolant mass flow measurements obtained during Test LOC-11C performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL) are presented. Although a density gradient existed within the pipe between 1 and 6 s, the homogeneous flow model used to calculate the coolant mass flow from the measured mixture density, momentum flux, and volumetric flow was found to be generally satisfactory. A cross-sectional average density was determined by fitting a linear density gradient through the upper and lower chordal densities obtained from a three-beam gamma densitometer and then combining the result with the middle beam density. The integrated measured coolant mass flow was subsequently found to be within 5% if the initial mass inventory of the PBF loss-of-coolant accident (LOCA) system. The posttest calculations using the RELAP4/MOD6 computer code to determine coolant mass flow for Test LOC-11C also agreed well with the measured data

  1. Application of a non-equilibrium drift flux model to two-phase blowdown experiments

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1976-08-01

    A vapor drift-flux model has been applied to the discharge of two-phase mixtures under choked flow conditions, including equilibrium as well as non-equilibrium vapor generation models. The system of four conservation equations is being solved, using the method of characteristics. Closed form expressions have been obtained for the propagation velocities from approximate solutions of the system's characteristic determinant. Treatment of the phase change front as a discontinuity, similar to the treatment of shocks in single phase gas dynamics, permitted very accurate solutions. Good agreement with experimental data is shown

  2. Coarse woody debris and soil respiration 6 years post-tornado in a Piedmont forest blowdown

    Science.gov (United States)

    Oldfield, C.; Peterson, C. J.

    2017-12-01

    Severe wind disturbances can rapidly change carbon pools and fluxes in forests, causing a site to switch from a carbon sink to a source in a matter of minutes. Moreover, salvage logging after a disturbance can result in disturbed and compacted soil, altered woody debris carbon pools, and seedling mortality, all of which may further alter carbon dynamics beyond that caused by the disturbance itself. We measured down dead wood and soil respiration in the summer of 2017 at Boggs Creek Recreation Area in the Piedmont of northeast Georgia, the site of a severe tornado in 2011. Down dead wood and soil respiration were compared in control (intact forest), salvaged, and unsalvaged areas. Megagrams per hectare of down dead wood was significantly higher in the unsalvaged condition than the control or salvage logging condition (ANOVAs, p<0.05 in both cases). Conversely, the volume of down dead wood was not significantly different in the control when compared to the salvage logging condition (p=0.99). Soil respiration was significantly higher in the salvage logged condition than the control (p<0.05), but was not significantly different between the unsalvaged condition and the control (p=0.30) or the unsalvaged condition and the salvaged condition (p=0.58). This research shows that wind disturbances have a lasting impact on the amount of down dead wood in a forest, and salvage logging may lead to greater soil respiration years after the initial disturbance, both of which will influence the time elapsed before a disturbed forest switches from carbon source to carbon sink. Further research is needed to determine the duration of these effects, along with the carbon consequences for other forest carbon pools.

  3. Fuel assembly loads during a hypothetical blowdown event in a PWR

    International Nuclear Information System (INIS)

    Stabel, J.; Bosanyi, B.; Kim, J.D.

    1991-01-01

    As a consequence of a hypothetical sudden break of the main coolant pipe of a PWR, RPV-internals and fuel assemblies (FA's) are undergoing horizontal and vertical motions. FA's may impact against each other, against core shroud or against lower core support. The corresponding impact loads must be absorbed by the FA spacer grids and guide thimbles. In this paper FA-loads are calculated with and without consideration of Fluid-Structure-Interaction (FSI) effects for assumed different break sizes of the main coolant pipe. The analysis has been performed for a hypothetical cold leg break of a typical SIEMENS-4 loop plant. For this purpose the codes DAPSY/DAISY (GRS, Germany) were coupled with the structural code KWUSTOSS (SIEMENS). It is shown that the FA loads obtained in calculations with consideration of FSI effects are by a factor of 2-4 lower than those obtained in the corresponding calculations without consideration of FSI. (author)

  4. Integrating remote sensing and forest inventory data for assessing forest blowdown in the boundary waters canoe area wilderness

    Science.gov (United States)

    Mark D. Nelson; W. Keith Moser

    2007-01-01

    The USDA Forest Service's Forest Inventory and Analysis (FIA) program conducts strategic inventories of our Nation's forest resources. There is increasing need to assess effects of forest disturbance from catastrophic events, often within geographic extents not typically addressed by strategic forest inventories. One such event occurred within the Boundary...

  5. Measurement of iodine released in a blowdown accident in the HTR-Modul. Final report on flow tests

    International Nuclear Information System (INIS)

    Zentis, A.

    1993-01-01

    A passive measuring device has been designed which consists of several filter cartridges of differnt length, and which is placed into the depressurization channel of the reactor. The dependence of the rate of flow through the filter on the flow rate in the depressurization channel must be known in order to be able to derive from the radioactivity deposited and measured in the filters a value indicating the total amount of iodine released. The report explains the basic principles of design of the instrument and of the experiments, and gives an interpretation of results of the flow tests in the AVA (aerodynamic testing facility) at Interatom. These flow tests have shown that it is feasible to determine the order of magnitude of iodine emissions with the given method and instrument. (orig./HP) [de

  6. 2D fluid flow in the downcomer and dynamic response of the core barrel during PWR blowdown

    International Nuclear Information System (INIS)

    Katz, F.; Krieg, R.; Ludwig, A.; Schlechtendahl, E.G.; Stoelting, K.

    1977-01-01

    As a part of the HDR program, methods for coupled fluid-structural dynamics are being developed. On the fluid side the 2D finite difference code YAQUI has been modified (it became YAQUIR) and adapted to describe the fluid dynamics in the downcomer of PWR's. On the structural side for determination of the dynamic core barrel response the code CYLDY2 has been developed. In this code the core barrel is treated as a thin cylindrical shell fixed at the upper end and ring stiffened at the lower end. The mass of the lower end ring also simulated a part of the core mass. Both models have been successfully tested. Coupling has been achieved for a simplified structural model proving the correctness of the coupling procedure. The structural model CYLDY2 is based on Fluegge's shell equations and uses variational principles. The solution is a superposition of steady-state and transient eigenfunctions. Results indicate that for the relatively thin-walled core barrel of the HDR-experiments in most cases the local deformations are somewhat higher than the global deformation (beam model). The coupling of YAQUIR and CYLDY2 is performed by imbedding the structural model in the fluid model. Fluid velocities are parallel to the fluid/structure interface. The structure desplacements define the time and space dependent thickness of the two-dimensional fluid layer (2 1/2-dimensional model)

  7. 2D fluid flow in the downcomer and dynamic response of the core barrel during PWR blowdown

    International Nuclear Information System (INIS)

    Katz, F.; Krieg, R.; Ludwig, A.; Schlechtendahl, E.G.; Stoelting, K.

    1977-01-01

    As a part of the HDR program, methods for coupled fluid/structure dynamics are being developed. On the fluid side the 2D finite difference code YAQUI has been modified and adapted to describe the fluid dynamics in the downcomer of PWR's. On the structural side for determination of the dynamic core barrel response the code CYLDY2 has been developed. In this code the core barrel is treated as a thin cylindrical shell fixed at the upper end and ring stiffened at the lower end. The mass of the lower end ring also simulates a part of the core mass. Both models have been successfully tested. Coupling has been achieved for a simplified structural model proving the correctness of the coupling procedure. YAQUIR is a significantly modified version of the code YAQUI. The coupling of YAQUIR and CYLDY2 is performed by imbedding the structural model in the fluid model. Fluid velocities are parellel to the fluid/structure interface. The structure displacements define the time and space dependent thickness of the two-dimensional fluid layer. While coupling of the complete CYLDY2 model with YAQUIR is still underway, results have been obtained with a simple axisymmetric structural model. For an axisymmetric test case three forms of pressure fluctuations have been observed: 1) radial oscillations dominated by the local compressibility of the water, 2) axial compression/expansion waves in the water considerably different from those obtained for a rigid barrel, 3) bulk axial water oscillations dominated by the global compressibility of the core barrel. (Auth.)

  8. Analysis of hide-out return and the composition of the steam generators' combined blowdown under ethanolamine chemistry

    International Nuclear Information System (INIS)

    Rodriguez Aliciardi, M; Belloni, M; Croatto, F; Ferrari, F; Herrera, C; Mendizabal, M; Montes, J; Saucedo, R; Rodriguez, I; Chocron, M

    2012-01-01

    Among the techniques recommended for the surveillance of the steam generators (SG's) are (i) the follow-up of the corrosion products in the feed water and blown-down water and the corresponding accumulation mass balance and (ii) the follow-up of the ionic impurities normally present in the main condensate and feed water that can enter in the cycle either through condenser leakages or even with the make-up water. This is, no matter how well the cycle is constructed or the make-up water plant behaves, given the continuous phase change present in the cycle, a concentration equilibrium is established. Moreover, a synergistic effect exists between both, corrosion products and impurities, (i) and (ii), because impurities concentrate in crevices and deposits creating a potentially corrosion risky environment for the SG's tubes. Hide-out return determination consists in the measurement of ions' concentrations, present in the blown-down water which are eluted in the shutdown during the power and temperature reduction. The integrated values allow graphical representations for performing several evaluations like: (i) the performance make-up water plant during a period, (ii) the presence of condenser in-leakages, (iii) the SG's deposits estimation and (iv) the internal chemistry of the SG's liquid phase inside the sludge pile. CNE, a CANDU 6 pressurized heavy-water reactor (PHWR) located in Embalse, Cordoba Province, Argentina, is close to reach the end-of-design-life and a life extension project is currently ongoing. In the plant also the chemistry of the BOP has been modified from a morpholine to an ethanolamine chemistry (Fernandez et. al, NPC'2010, October 3-7, Quebec City, Canada) to enhance protection of SG's internals and new SG's will be installed. Then, it is of interest to set up a hide-out return procedure and analysis of the combined blown-down crud collected during the shut down as a base line for the second period of operation. The results of the sampling performed during the 2011 Plant Programmed Outage (PO) are presented, compared to previous data and discussed (author)

  9. Subcooled decompression analysis of the ROSA and the LOFT semiscale blowdown test data with the digital computer code DEPCO-MULTI

    International Nuclear Information System (INIS)

    Namatame, Ken; Kobayashi, Kensuke

    1975-12-01

    In the ROSA (Rig of Safety Assessment) program, the digital computer code DEPCO-SINGLE and DEPCO-MULTI (Subcooled Decompression Process in Loss-of-Coolant Accident - Single Pipe and - Multiple Pipe Network) were prepared to study thermo-hydraulic behavior of the primary coolant in subcooled decompression of the PWR LOCA. The analytical results with DEPCO-MULTI on the subcooled decompression phenomena are presented for ROSA-I, ROSA-II and LOFT 500, 600, 700 and 800 series experiments. The effects of space mesh length, elasticity of pressure boundary materials and simplification for computational piping system on the computed result are described. This will be the final work on the study of the subcooled decompression analysis as for the ROSA program, and the authors wish that the present code shall further be examined with the data of much advanced experiments. (auth.)

  10. 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.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. 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

    International Nuclear Information System (INIS)

    Wachs, D. M.

    1998-01-01

    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

  12. Application of NEA/CSNI standard problem 3 (blowdown and flow reversal in the IETA-1 rig) to the validation of the RELAP-UK Mk IV code

    International Nuclear Information System (INIS)

    Bryce, W.M.

    1977-10-01

    NEA/CSNI Standard Problem 3 consists of the modelling of an experiment on the IETI-1 rig, in which there is initially flow upwards through a feeder, heated section and riser. The inlet and outlet are then closed and a breach opened at the bottom so that the flow reverses and the rig depressurises. Calculations of this problem by many countries using several computer codes have been reported and show a wide spread of results. The purpose of the study reported here was the following. First, to show the sensitivity of the calculation of Standard Problem 3. Second, to perform an ab initio best estimate calculation using the RELAP-UK Mark IV code with the standard recommended options, and third, to use the results of the sensitivity study to show where tuning of the RELAP-UK Mark IV recommended model options was required. This study has shown that the calculation of Standard Problem 3 is sensitive to model assumptions and that the use of the loss-of-coolant accident code RELAP-UK Mk IV with the standard recommended model options predicts the experimental results very well over most of the transient. (U.K.)

  13. Installation Assessment of Headquarters, Walter Reed Army Medical Center, Washington, DC and Noncontiguous Sections Forest Glen, Silver Spring, Maryland and Glen Haven, Wheaton, Maryland.

    Science.gov (United States)

    1984-06-01

    treated with sodium hexametaphosphate, sodium hydroxide, and tannin to reduce scaling and corrosion potential. Boilers are fueled as noted in Table 2.1...sulfides, phosphates, and tannins . The blowdown is unmetered for flow; blowdown occurs for approximately 2 minutes per day, 7 days a week. The blowdown...1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50

  14. Provision and Updating of Estimates of Reliability Parameters for Use in Reliability Analyses of Safety-Instrumented Systems

    OpenAIRE

    Bjartnes, Magnus Woll

    2012-01-01

    Safety-instrumented systems are implemented in the industry to prevent accidents to occur and escalate. The blowdown system on a oil production ship is one example of such system. If a fire breaks out on the ship, the blowdown system's role is to remove the flammable gases from the current production lines on the ship. This is done by opening of the blowdown valves, that are installed on the different production lines. In this thesis, the blowdown system on a new Teekay ship, and espesically ...

  15. 40 CFR 437.2 - General definitions.

    Science.gov (United States)

    2010-07-01

    ... sludges, metal-finishing rinse water and sludges, chromate wastes, blow-down water and sludges from air pollution control, spent anodizing solutions, incineration air pollution control wastewaters, waste liquid..., air pollution control scrubber blow-down, laboratory-derived wastewater, on-site landfill wastewaters...

  16. Numerical Simulation of Pulse Detonation Rocket-Induced MHD Ejector (PDRIME) Concepts for Advanced Propulsion Systems

    Science.gov (United States)

    2012-02-28

    the pressure at the nozzle exit drops during blowdown, the shock then slows down, and eventually, the ionized air in the bypass section starts tomove ...exit drops during blowdown, the shock then slows down, and eventually, the ionized air in the bypass section starts tomove downstream.At this point

  17. mckay correspondence in quasi-sl quasitoric orbifolds

    Indian Academy of Sciences (India)

    57

    McKay correspondence we have in mind. More precisely, the .... blowdown maps. In these notes, we establish the invariance of Betti numbers of. Chen-Ruan cohomology of quasi-SL quasitoric orbifolds under crepant blowdowns. ...... [1] Walter L. Baily, Jr The Decomposition Theorem for V-Manifolds American Journal of.

  18. Process fluid cooling system

    International Nuclear Information System (INIS)

    Farquhar, N.G.; Schwab, J.A.

    1977-01-01

    A system of heat exchangers is disclosed for cooling process fluids. The system is particularly applicable to cooling steam generator blowdown fluid in a nuclear plant prior to chemical purification of the fluid in which it minimizes the potential of boiling of the plant cooling water which cools the blowdown fluid

  19. Analysis of thermal fluctuations in the semiscale tests to determine flow transit delay times using a transfer function cross-correlation technique

    International Nuclear Information System (INIS)

    Raptis, A.C.; Popper, G.F.

    1977-08-01

    On April 14, 1976, EG and G performed the Semiscale Blowdown 29-1 experiment to try to establish the feasibility of using a transit time flowmeter (TTF) to measure transient blowdown two-phase flow rates. The recorded signals from that experiment were made available to and analyzed by the Argonne National Laboratory using the transfer function cross-correlation technique. The theoretical background for the transfer function method of analysis and the results of the data analysis are presented. Histograms of transit time during the blowdown are shown and topics for further investigation are identified

  20. Hypersonic Tunnel Facility (HTF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Hypersonic Tunnel Facility (HTF) is a blow-down, non-vitiated (clean air) free-jet wind tunnel capable of testing large-scale, propulsion systems at Mach 5, 6,...

  1. 40 CFR 98.250 - Definition of source category.

    Science.gov (United States)

    2010-07-01

    ...; asphalt blowing operations; blowdown systems; storage tanks; process equipment components (compressors... plants (i.e., hydrogen plants that are owned or under the direct control of the refinery owner and...

  2. 40 CFR 63.1360 - Applicability.

    Science.gov (United States)

    2010-07-01

    ... that are subject to subpart F of this part; (3) Production of ethylene; (4) Coal tar distillation; and... safety showers; (v) Noncontact steam boiler blowdown and condensate; (vi) Laundry water; (vii) Vessels...

  3. Feasibility of using microencapsulated phase change materials as filler for improving low temperature performance of rubber sealing materials.

    Science.gov (United States)

    Tiwari, Avinash; Shubin, Sergey N; Alcock, Ben; Freidin, Alexander B; Thorkildsen, Brede; Echtermeyer, Andreas T

    2017-11-01

    The feasibility of a novel composite rubber sealing material to improve sealing under transient cooling (in a so-called blowdown scenario) is investigated here. A composite of hydrogenated nitrile butadiene rubber (HNBR) filled with Micro Encapsulated Phase Change Materials (MEPCM) is described. The fillers contain phase change materials that release heat during the phase transformation from liquid to solid while cooling. This exotherm locally heats the rubber and may improve the function of the seal during a blowdown event. A representative HNBR-MEPCM composite was made and the critical thermal and mechanical properties were obtained by simulating the temperature distribution during a blowdown event. Simulations predict that the MEPCM composites can delay the temperature decrease in a region of the seal during the transient blowdown. A sensitivity analysis of material properties is also presented which highlights possible avenues of improvement of the MEPCMs for sealing applications.

  4. LOFT reactor vessel 2900 downcomer stalk instrument penetration flange stress analysis

    International Nuclear Information System (INIS)

    Finicle, D.P.

    1978-01-01

    The LOFT Reactor Vessel 290 0 Downcomer Stalk Instrument Penetration Flange Stress Analysis has been completed using normal operational and blowdown loading. A linear elastic analysis was completed using simplified hand analysis techniques. The analysis was in accordance with the 1977 ASME Boiler and Pressure Vessel Code, Section III, for a Class 1 component. Loading included internal pressure, bolt preload, and thermal gradients due to normal operating and blowdown

  5. 40 CFR 63.7740 - What are my monitoring requirements?

    Science.gov (United States)

    2010-07-01

    ...) You must at all times monitor the 3-hour average pH of the scrubber blowdown using CPMS according to the requirements in § 63.7741(e)(2) or measure and record the pH of the scrubber blowdown once per production cycle using a pH probe and meter according to the requirements in § 63.7741(e)(3). (h) For one or...

  6. Industrial Water Analysis Program: A Critical Study.

    Science.gov (United States)

    1983-09-01

    more readily removed by blowdown tannin decreases sludge accumulation and scale for- mation in the boiler . . . . Tannin has as a corro- sive control... tannin : Quebracho tannin contributes to several water treatment objectives. One principal contribution is to keep sludge fluid so that it can be...carried by the cir- culating boiler water and more readily removed by blowdown. This results from the collodial property of tannin which causes the sludge

  7. Radioactive waste treatment apparatus

    International Nuclear Information System (INIS)

    Abrams, R.F.; Chellis, J.G.

    1983-01-01

    Radioactive waste treatment apparatus is disclosed in which the waste is burned in a controlled combustion process, the ash residue from the combustion process is removed and buried, the gaseous effluent is treated in a scrubbing solution the pH of which is maintained constant by adding an alkaline compound to the solution while concurrently extracting a portion of the scrubbing solution, called the blowdown stream. The blowdown stream is fed to the incinerator where it is evaporated and the combustibles in the blowdown stream burned and the gaseous residue sent to the scrubbing solution. Gases left after the scrubbing process are treated to remove iodides and are filtered and passed into the atmosphere

  8. pH control in RCW loops at ORGDP

    International Nuclear Information System (INIS)

    Spann, B.M.

    1982-01-01

    The pH of RCW is a vital parameter in limiting corrosion and scaling in cooling water systems. At ORGDP, pH has in the past been adjusted by the addition of sulfuric acid to makeup water from the softeners. Since each of the RCW loops is independent and operates at varying flow rates, temperatures, and cycles of concentration that require different quantities of sulfuric acid for pH adjustment, acid addition only to the makeup water header has resulted in unsatisfactory pH control. This has resulted in increased blowdown requirements to maintain acceptable sulfate levels in the RCW. To minimize blowdown, system improvements are to be provided by an FY 1981 GPP project which will permit independent pH adjustment in each of the six RCW and blowdown resoftening loops

  9. Computational modeling and analysis of heavy water losses in boiler blow down with different positions of BBW-V100 at KANUPP

    International Nuclear Information System (INIS)

    Maqbool, M. U.

    2012-01-01

    The term blowdown is referred to the boilers and steam generators. Blowing down water from the steam generators maintains the chemistry of the feedwater and helps prevent scaling or sludge formation. In a nuclear power plant, the primary loop contains some activity in the form of tritium content. In boilers, primary and secondary systems interface and due to the pressure difference there is always a chance of mixing of primary and secondary fluids in event of tube leak. This primary fluid i.e., heavy water in our case can be lost through the blowdown lines after mixing with the feedwater. This thesis is a computational work for the determination of heavy water losses through the blowdown lines. (author)

  10. An analysis methodology for hot leg break mass and energy release

    International Nuclear Information System (INIS)

    Song, Jin Ho; Kwon, Young Min; Kim, Taek Mo; Chung, Hae Yong; Lee, Sang Jong

    1996-07-01

    An analysis methodology for the hot leg break mass and energy release is developed. For the blowdown period a modified CEFLASH-4A analysis is suggested. For the post-blowdown period a new computer model named COMET is developed. Differently from previous post-blowdown analysis model FLOOD3, COMET is capable of analyzing both cold leg and hot leg break cases. The cold leg break model is essentially same as that of FLOOD3 with some improvements. The analysis results by the newly proposed hot leg break model in the COMET is in the same trend as those observed in scaled-down integral experiment. And the analyses results for the UCN 3 and 4 by COMET are qualitatively and quantitatively in good agreement with those predicted by best-estimate analysis by using RELAP5/MOD3. Therefore, the COMET code is validated and can be used for the licensing analysis. 6 tabs., 82 figs., 9 refs. (Author)

  11. Large-scale wind disturbances promote tree diversity in a Central Amazon forest.

    Science.gov (United States)

    Marra, Daniel Magnabosco; Chambers, Jeffrey Q; Higuchi, Niro; Trumbore, Susan E; Ribeiro, Gabriel H P M; Dos Santos, Joaquim; Negrón-Juárez, Robinson I; Reu, Björn; Wirth, Christian

    2014-01-01

    Canopy gaps created by wind-throw events, or blowdowns, create a complex mosaic of forest patches varying in disturbance intensity and recovery in the Central Amazon. Using field and remote sensing data, we investigated the short-term (four-year) effects of large (>2000 m(2)) blowdown gaps created during a single storm event in January 2005 near Manaus, Brazil, to study (i) how forest structure and composition vary with disturbance gradients and (ii) whether tree diversity is promoted by niche differentiation related to wind-throw events at the landscape scale. In the forest area affected by the blowdown, tree mortality ranged from 0 to 70%, and was highest on plateaus and slopes. Less impacted areas in the region affected by the blowdown had overlapping characteristics with a nearby unaffected forest in tree density (583 ± 46 trees ha(-1)) (mean ± 99% Confidence Interval) and basal area (26.7 ± 2.4 m(2) ha(-1)). Highly impacted areas had tree density and basal area as low as 120 trees ha(-1) and 14.9 m(2) ha(-1), respectively. In general, these structural measures correlated negatively with an index of tree mortality intensity derived from satellite imagery. Four years after the blowdown event, differences in size-distribution, fraction of resprouters, floristic composition and species diversity still correlated with disturbance measures such as tree mortality and gap size. Our results suggest that the gradients of wind disturbance intensity encompassed in large blowdown gaps (>2000 m(2)) promote tree diversity. Specialists for particular disturbance intensities existed along the entire gradient. The existence of species or genera taking an intermediate position between undisturbed and gap specialists led to a peak of rarefied richness and diversity at intermediate disturbance levels. A diverse set of species differing widely in requirements and recruitment strategies forms the initial post-disturbance cohort, thus lending a high resilience towards wind

  12. LOFT test support branch data abstract report: one-sixth scale model BWR jet pump test

    International Nuclear Information System (INIS)

    Crapo, H.S.

    1979-01-01

    Pump performance data are presented for a 1/6 scale model jet pump in tests conducted at the LOFT Test Support Blowdown Facility. Steady-state subcooled pump characterization tests were performed over a wide range of forward and reverse flow conditions, both at room temperature, and at elevated temperature (555 0 K). Blowdown tests were also performed to obtain two-phase performance data in configurations simulating the flow patterns in the intact and broken loops of a BWR during a recirculation line break transient

  13. Ecological impact of chloro-organics produced by chlorination of cooling tower waters

    Energy Technology Data Exchange (ETDEWEB)

    Jolley, R L; Cumming, R B; Pitt, W W; Taylor, F G; Thompson, J E; Hartmann, S J

    1977-01-01

    Experimental results of the initial assessment of chlorine-containing compounds in the blowdown from cooling towers and the possible mutagenic activity of these compounds are reported. High-resolution liquid chromatographic separations were made on concentrates of the blowdown from the cooling tower at the High Flux Isotope Reactor (HFIR) and from the recirculating water system for the cooling towers at the Oak Ridge Gaseous Diffusion Plant (ORGDP), Oak Ridge, Tennessee. The chromatograms of chlorinated cooling waters contained numerous uv-absorbing and cerate-oxidizable constituents that are now being processed through a multicomponent identification procedure. Concentrates of the chlorinated waters are also being examined for mutagenic activity.

  14. 40 CFR 471.11 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ...-forming wet air pollution control scrubber blowdown. Subpart A—BPT Pollutant or pollutant property Maximum... solutions. Subpart A—BPT Pollutant or pollutant property Maximum for any 1 day Maximum for monthly average mg/off-kg (pounds per million off-pound) of lead-tin-bismuth rolled with soap solutions Antimony 0...

  15. 40 CFR 471.12 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ...) Shot-forming wet air pollution control scrubber blowdown. Subpart A—BAT Pollutant or pollutant property... 0.030 Lead 0.010 0.005 (b) Rolling spent soap solutions. Subpart A—BAT Pollutant or pollutant... lead-tin-bismuth rolled with soap solutions Antimony 0.120 0.055 Lead 0.018 0.009 (c) Drawing spent...

  16. Decomposition of log crepant birational morphisms between log terminal surfaces

    OpenAIRE

    Fukuda, Shigetaka

    1999-01-01

    We prove that every log crepant birational morphism between log terminal surfaces is decomposed into log-flopping type divisorial contraction morphisms and log blow-downs. Repeating these two kinds of contractions we reach a minimal log minimal surface from any log minimal surface.

  17. Innovations in fuels management: Demonstrating success in treating a serious threat of wildfire in Northern Minnesota

    Science.gov (United States)

    Dennis Neitzke

    2007-01-01

    This case study illustrates the positive effects of strategic fuels treatments in continuous heavy fuels. In 1999, a severe windstorm blew down close to 1,000 square miles of forest land in northern Minnesota and Canada. As much as 400,000 acres of the blowdown occurred in the Boundary Waters Canoe Area Wilderness. Fire experts were invited to assess the hazardous...

  18. 40 CFR 471.72 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ... shall be no discharge of process wastewater pollutants. (e) Surface treatment spent baths. Subpart G—BAT...—Subpart G—BAT. There shall be no discharge of process wastewater pollutants. (b) Extrusion tool contact....752 (g) Wet air pollution control scrubber blowdown. Subpart G—BAT Pollutant or pollutant property...

  19. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  20. A validation of ATR LOCA thermal-hydraulic code with a statistical approach

    International Nuclear Information System (INIS)

    Mochizuki, Hiroyasu

    2000-01-01

    When cladding temperatures are measured for a blowdown experiment, cladding temperatures at the same elevation in the fuel bundle have usually some differences due to eccentricity of the fuel bundle and other reasons such as biased two-phase flow. In the present paper, manufacturing tolerances and uncertainties of thermal-hydraulics are incorporated into a LOCA code that is applied with the statistical method. The present method was validated with the results of different blowdown experiments conducted using the 6 MW blowdown facility simulating the Advanced Thermal Reactor (ATR). In the present statistical method, the code was modified to run fast in order to calculate the blowdown thermal-hydraulics a lot of times with the code using different sets of input data. These input data for sizes and empirical correlations are prepared by the effective Monte-Carlo method based on the distribution functions deduced by the measured manufacturing errors and the uncertainties of thermal hydraulics. The calculated curves express uncertainties due to the different input deck. The uncertainty band and tendency of the cladding temperature were dependent on the beak sizes in the experiment. The measured results were traced by the present method. (author)

  1. 75 FR 54961 - Final Supplemental Environmental Impact Statement, Single Nuclear Unit at the Bellefonte Plant...

    Science.gov (United States)

    2010-09-09

    ... facility, a wind farm, a methane- gas cofiring facility, and several small solar photovoltaic facilities... use the existing natural draft cooling towers, water intake channel and pumping station, blowdown... including wind, solar, biomass, and hydropower. All of these energy resources have a place in TVA's plans...

  2. A modular assembly method of a feed and thruster system for Cubesats

    NARCIS (Netherlands)

    Louwerse, M.C.; Jansen, Henricus V.; Elwenspoek, Michael Curt

    2010-01-01

    A modular assembly method for devices based on micro system technology is presented. The assembly method forms the foundation for a miniaturized feed and thruster system as part of a micro propulsion unit working as a simple blow-down system of a rocket engine. The micro rocket is designed to be

  3. Three dimensional analysis of turbulent steam jets in enclosed structures: a CFD approach

    International Nuclear Information System (INIS)

    Ishii, M.; NguyenLe, Q.

    1999-01-01

    This paper compares the three-dimensional numerical simulation with the experimental data of a steam blowdown event in a light water reactor containment building. The temperature and pressure data of a steam blowdown event was measured at the Purdue University Multi-Dimensional Integrated Test Assembly (PUMA), a scaled model of the General Electric simplified Boiling Water Reactor. A three step approach was used to analyze the steam jet behavior. First, a 1-Dimensional, system level RELAP5/Mod3.2 model of the steam blowdown event was created and the results used to set the initial conditions for the PUMA blowdown experiments. Second, 2-Dimensional CFD models of the discharged steam jets were computed using PHOENICS, a commercially available CFD package. Finally, 3-Dimensional model of the PUMA drywell was created with the boundary conditions based on experimental measurements. The results of the 1-D and 2-D models were reported in the previous meeting. This paper discusses in detail the formulation and the results of the 3-Dimensional PHOENICS model of the PUMA drywell. It is found that the 3-D CFD solutions compared extremely well with the measured data

  4. The National Shipbuilding Research Program. Trailer Mounted Water Recovery and Reuse System

    Science.gov (United States)

    2000-11-30

    blowdown reclamation, dye purification, coolant recovery, desalting of brackish wastewater, and marginal reduction of biological and chemical oxygen... electrostatic ppt. Zn, TSS Sewer NASSCO 5,700 gal/yr (1998) Zn WWTF Note: Blank cells indicate that data was not available at this time. Key: gal

  5. 40 CFR 421.96 - Pretreatment standards for new sources.

    Science.gov (United States)

    2010-07-01

    ... wastewater pollutants in metallurgical acid plant blowdown introduced into a POTW shall not exceed the following values: Subpart I—Metallurgical Acid Plant—PSNS Pollutant or pollutant property Maximum for any 1... GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Metallurgical Acid Plants...

  6. 40 CFR 421.95 - Pretreatment standards for existing sources.

    Science.gov (United States)

    2010-07-01

    ... standards for existing sources. The mass of wastewater pollutants in metallurgical acid plant blowdown introduced into a POTW shall not exceed the following values: Subpart I—Metallurgical Acid Plant—PSES...) EFFLUENT GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Metallurgical Acid...

  7. The efficacy of salvage logging in reducing subsequent fire severity in conifer-dominated forests of Minnesota, USA

    Science.gov (United States)

    Fraver, S.; Jain, T.; Bradford, J.B.; D'Amato, A.W.; Kastendick, D.; Palik, B.; Shinneman, D.; Stanovick, J.

    2011-01-01

    Although primarily used to mitigate economic losses following disturbance, salvage logging has also been justified on the basis of reducing fire risk and fire severity; however, its ability to achieve these secondary objectives remains unclear. The patchiness resulting from a sequence of recent disturbances-blowdown, salvage logging, and ildfire- provided an excellent opportunity to assess the impacts of blowdown and salvage logging on wildfire severity. We used two fire-severity assessments (tree-crown and forest-floor characteristics) to compare post-wildfire conditions among three treatment combinations (Blowdown-Salvage-Fire, Blowdown-Fire, and Fire only). Our results suggest that salvage logging reduced the intensity (heat released) of the subsequent fire. However, its effect on severity (impact to the system) differed between the tree crowns and forest floor: tree-crown indices suggest that salvage logging decreased fire severity (albeit with modest statistical support), while forest-floor indices suggest that salvage logging increased fire severity. We attribute the latter finding to the greater exposure of mineral soil caused by logging operations; once exposed, soils are more likely to register the damaging effects of fire, even if fire intensity is not extreme. These results highlight the important distinction between fire intensity and severity when formulating post-disturbance management prescriptions. ?? 2011 by the Ecological Society of America.

  8. Effects of multiple interacting disturbances and salvage logging on forest carbon stocks

    Science.gov (United States)

    Bradford, J.B.; Fraver, S.; Milo, A.M.; D'Amato, A.W.; Palik, B.; Shinneman, D.J.

    2012-01-01

    Climate change is anticipated to increase the frequency of disturbances, potentially impacting carbon stocks in terrestrial ecosystems. However, little is known about the implications of either multiple disturbances or post-disturbance forest management activities on ecosystem carbon stocks. This study quantified how forest carbon stocks responded to stand-replacing blowdown and wildfire, both individually and in combination with and without post-disturbance salvage operations, in a sub-boreal jack pine ecosystem. Individually, blowdown or fire caused similar decreases in live carbon and total ecosystem carbon. However, whereas blowdown increased carbon in down woody material and forest floor, fire increased carbon in standing snags, a difference that may have consequences for long-term carbon cycling patterns. Fire after the blowdown caused substantial additional reduction in ecosystem carbon stocks, suggesting that potential increases in multiple disturbance events may represent a challenge for sustaining ecosystem carbon stocks. Salvage logging, as examined here, decreased carbon stored in snags and down woody material but had no significant effect on total ecosystem carbon stocks.

  9. 40 CFR 125.93 - What special definitions apply to this subpart?

    Science.gov (United States)

    2010-07-01

    .... The water is usually sent to a cooling canal or channel, lake, pond, or tower to allow waste heat to... losses that have occurred due to blowdown, drift, and evaporation. Cooling water means water used for... systems sometimes employ canals/channels, ponds, or non-recirculating cooling towers to dissipate waste...

  10. Analysis of the large break loss of coolant accidents in nuclear power plants by the computer code RELAP4/MOD6

    International Nuclear Information System (INIS)

    Gregoric, M.; Stritar, A.

    1983-01-01

    The safety analysis of the nuclear power plant Krsko by the code RELAP4/MOD6 is described. Methodology for the safety evaluation for the case of the Large LOCA is introduced. The problems encountered during the analysis of the blowdown phase of the accident are described. Some results of double ended cold leg LOCA analysis for different break sizes are shown. (author)

  11. Specific experiments carried out in Germany in order to demonstrate the safety of existing structures

    International Nuclear Information System (INIS)

    Krutzik, Norbert

    2002-01-01

    Specific experiments are carried out in Germany in order to demonstrate the safety of existing NPPs. HDR research program includes operational loads testing (pressure test, pressure and temperature test, thermal shock, fatigue); extreme loads (earthquake, aircraft crash, external explosion); internal emergency loads (blowdown, hydrogen combustion, fire, thermal shock, water hammer, condensation loads)

  12. Review of literature on catalytic recombination of hydrogen--oxygen

    International Nuclear Information System (INIS)

    Homsy, R.V.; Glatron, C.A.

    1968-01-01

    The results are reported of a literature search for information concerning the heterogeneous, gas phase, catalytic hydrogen-oxygen recombination. Laboratory scale experiments to test the performance of specific metal oxide catalysts under conditions simulating the atmosphere within a nuclear reactor containment vessel following a loss-of-coolant blowdown accident are suggested

  13. Wind Disturbance Produced Changes in Tree Species Assemblage in the Peruvian Amazon

    Science.gov (United States)

    Rifai, S. W.; Chambers, J. Q.; Negron Juarez, R. I.; Ramirez, F.; Tello, R.; Alegria Muñoz, W.

    2010-12-01

    Wind disturbance has been a frequently overlooked abiotic cause of mass tree mortality in the Amazon basin. In the Peruvian Amazon these wind disturbances are produced by meteorological events such as convective systems. Downbursts for example produce short term descendent wind speeds that can be in excess of 30 m s-1. These are capable of producing tree blowdowns which have been reported to be as large as 33 km2 in the Amazon basin. We used the chronosequence of Landsat Satellite imagery to find and locate where these blowdowns have occurred in the Loreto region of the Peruvian Amazon. Spectral Mixture Analysis was used to estimate the proportion landcover of green vegetation, non-photosynthetic vegetation (NPV), soil and shade in each pixel. The change in NPV was calculated by subtracting the NPV signal in the Landsat image prior to the blowdown occurrence, from the image following the disturbance. Our prior research has established a linear relationship between tree mortality and change in NPV. It is hypothesized that these mass tree mortality events result in changes in the tree species assemblage of affected forests. Here we present preliminary tree species assemblage data from two sites in the Peruvian Amazon near Iquitos, Peru. The site (ALP) at the Allpahuayo Mishana reserve (3.945 S, 73.455 W) is 30 km south of Iquitos, Peru, and hosts the remnants of a 50 ha blowdown that occurred in either 1992 or 1993. Another site (NAPO) on the Napo river about 60 km north of Iquitos, is the location of an approximately 300 ha blowdown that occurred in 1998. At each site, a 3000 m x 10 m transect encompassing non disturbed and disturbed areas was installed, and trees greater than 10 cm diameter at breast height were measured for diameter, height and were identified to the species. Stem density of trees with diameter at breast height > 10 cm, and tree height appear to be similar both inside and outside the blowdown affected areas of the forests at both sites. At the ALP

  14. Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

    International Nuclear Information System (INIS)

    DeMars, R.V.; Steinke, R.R.

    1975-01-01

    Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

  15. ROSA-II test data report, 8

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Adachi, Hiromichi; Okazaki, Motoaki; Sobajima, Makoto; Shiba, Masayoshi

    1977-09-01

    Results of the ROSA-II test simulating a loss-of-coolant accident (LOCA) and effect of the emergency core cooling system (ECCS) in a pressurized water reactor (PWR) are presented including the test conditions and interpretations of the phenomena for test runs 324, 325 and 326. Each test was made in a double-ended guillotine break of the cold leg. Effects of the following parameters on the coolant behavior in both blowdown and reflooding phases were examined in comparing between runs, previous test with the same ECCS injection position, etc: 1) increase of the downcomer gap, 2) increase of ECCS injection flow rate, 3) enlargement of the outlet flow area of pump in the blowdown loop. Increasing the downcomer gap, the effect of counter current flow limit (CCFL) decreased and water accumulation increased. By increasing the ECC water injection rate by 1.5 times the normal, the rate of water accumulation increased by 6 times the normal. (auth.)

  16. Multi-circulation boiler design

    Energy Technology Data Exchange (ETDEWEB)

    MacKenzie, Malcolm [Babcock and Wilcox Canada Ltd (Canada)

    2011-07-01

    Steam assisted gravity drainage (SAGD) requires steam, which was historically supplied by once-through steam generators (OTSG). OTSGs are now often replaced by drum boilers for steam generation. Drum boilers have a higher capacity than OTSGs and produce reduced blowdown. However, the volume of contaminants in feedwater is generally significant and ASME criteria require high-quality feedwater. The current study investigates a new design for a multi-circulation boiler that would allow cleaner steam to be generated from lower quality feedwater. In drum boilers, water circulates in one zone. By analyzing circulation fundamentals (notably heat flux and void fraction), the design of multiple circulation zones was envisaged to separate the clean water from the contaminated water. Working with distinct circulation zones would allow the quality of the steam from low quality feedwater to be improved, all the while maintaining the blowdown. The new design would result in a boiler which is easier to clean and would meet ASME criteria.

  17. Vitrification of hazardous and mixed wastes

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Pickett, J.B.; Ramsey, W.G.

    1992-01-01

    Solidification of hazardous/mixed wastes into glass is being examined at the Savannah River Site. The first hazardous/mixed wastes glassified at SRS have been (1) incinerator and (2) nickel plating line (F006) wastes. Solidification of incinerator blowdown and mixtures of incinerator blowdown and incinerator bottom kiln ash have been achieved in Soda (Na 2 O) - Lime (CaO) - Silica (SiO 2 ) glass (SLS) at waste loadings of up to 50 wt%. Solidification of nickel-plating line waste sludges containing depleted uranium have also been achieved in both SLS and borosilicate glasses at waste loadings of 75 wt%. This corresponds to volume reductions of 97% and 81%, respectively. Further studies will examine glassification of: ion exchange zeolites, inorganic filter media, asbestos, glass fiber filters, contaminated soil, cementitious, or other materials in need of remediation

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Informing hazardous zones for on-board maritime hydrogen liquid and gas systems

    Energy Technology Data Exchange (ETDEWEB)

    Blaylock, Myra L. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Pratt, Joseph William [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Bran Anleu, Gabriela A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Proctor, Camron [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2018-01-01

    The significantly higher buoyancy of hydrogen compared to natural gas means that hazardous zones defined in the IGF code may be inaccurate if applied to hydrogen. This could place undue burden on ship design or could lead to situations that are unknowingly unsafe. We present dispersion analyses to examine three vessel case studies: (1) abnormal external vents of full blowdown of a liquid hydrogen tank due to a failed relief device in still air and with crosswind; (2) vents due to naturally-occurring boil-off of liquid within the tank; and (3) a leak from the pipes leading into the fuel cell room. The size of the hydrogen plumes resulting from a blowdown of the tank depend greatly on the wind conditions. It was also found that for normal operations releasing a small amount of "boil- off" gas to regulate the pressure in the tank does not create flammable concentrations.

  20. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1975-01-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

  1. 40 CFR 471.14 - Pretreatment standards for existing sources (PSES).

    Science.gov (United States)

    2010-07-01

    ...) Shot-forming wet air pollution control scrubber blowdown. Subpart A—PSES Pollutant or pollutant... emulsions Antimony 0.067 0.030 Lead 0.010 0.005 (b) Rolling spent soap solutions. Subpart A—PSES Pollutant... off-pounds) of lead-tin-bismuth rolled with soap solutions Antimony 0.120 0.055 Lead 0.018 0.009 (c...

  2. Monitoring PWR reactor vessel liquid level with SPNDs during LOCAs

    International Nuclear Information System (INIS)

    Adams, J.P.

    1982-01-01

    Data from in-core self-powered neutron detectors taken during two nuclear loss-of-coolant accident simulations have been correlated with core moderator density changes. The detector current attenuation has been calculated during blowdown and reflood phases of the simulation. Based on these data, it is concluded that these detectors could be used to monitor reactor vessel liquid level during loss-of-coolant accidents in pressurized water reactors

  3. Feasibility Study of Coal Gasification/Fuel Cell/Cogeneration Project. Scranton, Pennsylvania Site. Project Description,

    Science.gov (United States)

    1985-11-01

    Estimated Water Emissions 143 7-4 Estimated Solid Wastes 144 7-5 Composition of Blowdown from Stretford Process 145 7-6 Summary of Environmental... processes are a costly alternative for the sulfur recovery process due to the high CO2 concentration in the gas (26% Vol). Therefore, a Stretford ...oxidation Stretford Sulfur Removal Process is used for the removal of H2S to the required level. The shifted gas stream is directed to venturi contactor, T

  4. Feasibility Study of Coal Gasification/Fuel Cell/Cogeneration Project. Fort Hood, Texas Site. Project Description,

    Science.gov (United States)

    1985-07-01

    Estimated Water Emissions14 7-4 Estimated Solid Wastes 149 7-5 Composition of Blowdown from Stretford Process 150 -1774A~ vi LIST OF TA~BLES (Cont’d...a costly alternative for the sulfur recovery process due to the high C02 concentration in the gas (25% Vol). Therefore, a Stretford liquid oxidation...The sulfur is separated from the solution, which is regenerated by air-sparging and recycled. Because the Stretford process cannot remove COS, a

  5. Transient two-phase flow

    International Nuclear Information System (INIS)

    Hsu, Y.Y.

    1974-01-01

    The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)

  6. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The key computer programs currently used by the analysis procedure of the ORNL-PWR Blowdown Heat Transfer Separate Effects Program are overviewed with particular emphasis placed on their interrelationships. The major modeling and calculational programs, COBRA, ORINC, ORTCAL, PINSIM, and various versions of RELAP4, are summarized and placed into the perspective of the procedure. The supportive programs, REDPLT, ORCPLT, BDHTPLOT, OXREPT, and OTOCI, and their uses are described

  7. The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

    OpenAIRE

    Jang Hyung-Wook; Lee Sang-Yong; Oh Seung-Jong; Kim Woong-Bae

    2017-01-01

    The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during ...

  8. ANALYSIS OF A STATION BLACKOUT SCENARIO WITH AN ATLAS TEST

    OpenAIRE

    YEON-SIK KIM; XIN-GUO YU; KYOUNG-HO KANG; HYUN-SIK PARK; SEOK CHO; KI-YONG CHOI

    2013-01-01

    A station blackout experiment called SBO-01 was performed at the ATLAS facility. From the SBO-01 test, the station blackout scenario can be characterized into two typical phases: A first phase characterized by decay heat removal through secondary safety valves until the SG dryouts, and a second phase characterized by an energy release through a blowdown of the primary system after the SG dryouts. During the second phase, some physical phenomena of the change over a pressurizer function, i.e.,...

  9. Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory, August 1976

    International Nuclear Information System (INIS)

    Fee, G.G.

    1976-10-01

    Technical highlights are presented for the following activities: heavy section steel technology, fission product beta and gamma energy release, LOCA release from LWR fuel, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, Zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, design criteria for piping and nozzles, and dose conversion factors for inhalation of radionuclides

  10. AGOR 28

    Science.gov (United States)

    2015-11-20

    between transducer and transceiver rooms. • HVAC - Vendor conducted operational checks of all fans, fan controls, heaters etc. Vendor is...Ducts • 555-007-3 Blowdown & Operational Test of CO2 System – Hazmat Lockers and Tunnel Thruster Motor w/ USCG in attendance 7. Captain...come back and make the swap 6 AUX Machinery Room bilge cleaned and paint touched up. Callenberg continues work balancing HVAC system

  11. Reuse of Treated Internal or External Wastewaters in the Cooling Systems of Coal-Based Thermoelectric Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Radisav Vidic; David Dzombak; Ming-Kai Hsieh; Heng Li; Shih-Hsiang Chien; Yinghua Feng; Indranil Chowdhury; Jason Monnell

    2009-06-30

    This study evaluated the feasibility of using three impaired waters - secondary treated municipal wastewater, passively treated abandoned mine drainage (AMD), and effluent from ash sedimentation ponds at power plants - for use as makeup water in recirculating cooling water systems at thermoelectric power plants. The evaluation included assessment of water availability based on proximity and relevant regulations as well as feasibility of managing cooling water quality with traditional chemical management schemes. Options for chemical treatment to prevent corrosion, scaling, and biofouling were identified through review of current practices, and were tested at bench and pilot-scale. Secondary treated wastewater is the most widely available impaired water that can serve as a reliable source of cooling water makeup. There are no federal regulations specifically related to impaired water reuse but a number of states have introduced regulations with primary focus on water aerosol 'drift' emitted from cooling towers, which has the potential to contain elevated concentrations of chemicals and microorganisms and may pose health risk to the public. It was determined that corrosion, scaling, and biofouling can be controlled adequately in cooling systems using secondary treated municipal wastewater at 4-6 cycles of concentration. The high concentration of dissolved solids in treated AMD rendered difficulties in scaling inhibition and requires more comprehensive pretreatment and scaling controls. Addition of appropriate chemicals can adequately control corrosion, scaling and biological growth in ash transport water, which typically has the best water quality among the three waters evaluated in this study. The high TDS in the blowdown from pilot-scale testing units with both passively treated mine drainage and secondary treated municipal wastewater and the high sulfate concentration in the mine drainage blowdown water were identified as the main challenges for blowdown

  12. A facility for the experimental investigation of single substance two phase flow

    International Nuclear Information System (INIS)

    Maeder, P.F.; Dickinson, D.A.; Nikitopoulos, D.E.; DiPippo, R.

    1985-01-01

    The paper describes a research facility dedicated to single-substance two-phase flow. The working fluid is dichlorotetrafluoroethane (or refrigerant R-114), allowing both operation at manageable pressures, temperatures and flowrates, and application of results to practical situations through similarity. Operation is in the blowdown mode. The control and data acquisition systems are fully automated and computer controlled. A range of flow conditions from predominantly liquid flow to high velocity, high void fraction choked flow can be attained

  13. Monthly highlights for Office of Nuclear Regulatory research programs at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Fee, G.G.

    1975-04-01

    Summaries are given of the following programs: heavy section steel technology, fission product beta and gamma energy release, LOCA release from LWR fuel, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer--separate effects, Zircaloy fuel cladding collapse studies, Zr metal--water oxidation kinetics, transient vaporization of LMFBR fuel, and HTGR safety analysis and research. Technical highlights and cost/budget reports are included. (U.S.)

  14. Use of a genetic algorithm to solve two-fluid flow problems on an NCUBE multiprocessor computer

    International Nuclear Information System (INIS)

    Pryor, R.J.; Cline, D.D.

    1992-01-01

    A method of solving the two-phase fluid flow equations using a genetic algorithm on a NCUBE multiprocessor computer is presented. The topics discussed are the two-phase flow equations, the genetic representation of the unknowns, the fitness function, the genetic operators, and the implementation of the algorithm on the NCUBE computer. The efficiency of the implementation is investigated using a pipe blowdown problem. Effects of varying the genetic parameters and the number of processors are presented

  15. Aspirated High Pressure Compressor

    Science.gov (United States)

    2006-08-01

    suggested by the General Electric Company based on an advanced military engine concept intended for high speed flight. The resultant high flight...section, consisting of the two rotors, each on an computed points. The peak pressure ratio is 3.08 and the peak independent, electric motor-driven...instrumentation along Moto . .. ,oto2 with the test error analysis were given by Onnee [14). SFlywheel 2 BLOWDOWN OPERATION T--’..Exit-t-rottle

  16. Documentation of CATHENA input files for the APOLLO computer

    International Nuclear Information System (INIS)

    1988-06-01

    Input files created for the VAX version of the CATHENA two-fluid code have been modified and documented for simulation on the AECB's APOLLO computer system. The input files describe the RD-14 thermalhydraulic loop, the RD-14 steam generator, the RD-12 steam generator blowdown test facility, the Stern Laboratories Cold Water Injection Facility (CWIT), and a CANDU 600 reactor. Sample CATHENA predictions are given and compared with experimental results where applicable. 24 refs

  17. Experimental investigation on the behavior of pressure suppression containment systems by the SOPRE-1 facility

    International Nuclear Information System (INIS)

    Cerullo, N.; Delli Gatti, A.; Marinelli, M.; Mazzini, M.; Mazzoni, A.; Sbrana, A.; Todisco, P.

    1977-01-01

    The SOPRE-1 test facility is an integral model (scale 1:13) of a MARK II pressure suppression containment system. It was set up at the University of Pisa in order to study the pressure-temperature transient in pressure suppression containment systems during LOCAs. Knowledge of this transient is necessary to perform a correct structural analysis of reactor containment. The containment system behavior is studied by changing the principal parameters which affect the transient (blow-down mass and energy release, suppression pool water temperature, vent pipe number and submergence, heat transfer coefficients). The first series of tests involved: A) 13 tests with break area of 1.8 cm 2 , B) 8 tests with break area of 20.0 cm 2 . The following experimental conditions were changed: position of the simulated break (from liquid or steam zone), water pressure (20-85 Kg/cm 2 ) and mass (45-70 Kg) in the vessel model. Tests A): the CONTEMPT codes correctly forecast the pressure-temperature history, both in dry- and in wet-well. Tests B): the experimental runs have shown that increasing of blow-down flowrate produces dry-well pressure spatial differences and anomalous vent pipe behavior. This results in damped oscillations of dry- and wet-well pressure, probably due to alternating air bubble over-expansion and collapse, and in vent pipe opening and reclosing. Dry-well pressure maxima at the end of blow-down are greater than those forecasted by currently applied codes: these codes use an homogeneous model, and do not take into account the above mentioned dynamic phenomena. In some tests other interesting phenomena were observed, such as some local pressure peaks in the suppression pool greater than dry-well pessure maxima at the end of blow-down. At present, all these phenomena are under study; they could be important for the structural analysis of containment systems

  18. Shut-down conditions, emergency cooling and essential services

    International Nuclear Information System (INIS)

    Belda, W.

    1977-01-01

    1) Introduction: Summary of system technology and reactor protection equipment. 2) Definitions. 3) LOCA: a) blowdown and refilling phase; b) jet and reaction forces; c) flow and heat transfer behavior in the core; d) behavior of the heater rods; e) core melting. 4) Protection against and during LOCA: a) general measures; b) break of a primary coolant pipe; c) break of a small pipe; d) break of a secondary pipe. (orig.) [de

  19. Pressure suppression pool mixing in passive advanced BWR plants

    International Nuclear Information System (INIS)

    Gamble, Robert E.; Nguyen, Thuy T.; Shiralkar, Bharat S.; Peterson, Per F.; Greif, Ralph; Tabata, H.

    2001-01-01

    In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data

  20. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  1. Thermal-hydraulics of the PFB/LOFT lead rod loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.; Garner, R.W.; MacDonald, P.E.; Cox, W.R.

    1980-01-01

    Results of the four PBF/LOFT Lead Rod sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods subjected to a series of nuclear blowdown tests, and to determine if subjecting deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature versus system pressure response with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements. Tests LLR-3, LLR-5, LLR-4, and LLR-4A were performed at system conditions of 595 0 K coolant inlet temperature, 15.5 MPa system pressure, and 41, 46, 57 and 56 kW/m test rod peak linear powers, respectively, at initiation of blowdown. Cladding temperatures during the tests ranged from 870 to 1260 0 K

  2. The probability of containment failure by direct containment heating in surry

    Energy Technology Data Exchange (ETDEWEB)

    Pilch, M.M.; Allen, M.D.; Bergeron, K.D.; Tadios, E.L.; Stamps, D.W. [Sandia National Labs., Albuquerque, NM (United States); Spencer, B.W. [Argonne National Lab., IL (United States); Quick, K.S.; Knudson, D.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1995-05-01

    In a light-water reactor core melt accident, if the reactor pressure vessel (RPV) fails while the reactor coolant system (RCS) at high pressure, the expulsion of molten core debris may pressurize the reactor containment building (RCB) beyond its failure pressure. A failure in the bottom head of the RPV, followed by melt expulsion and blowdown of the RCS, will entrain molten core debris in the high-velocity steam blowdown gas. This chain of events is called a high-pressure melt ejection (HPME). Four mechanisms may cause a rapid increase in pressure and temperature in the reactor containment: (1) blowdown of the RCS, (2) efficient debris-to-gas heat transfer, (3) exothermic metal-steam and metal-oxygen reactions, and (4) hydrogen combustion. These processes, which lead to increased loads on the containment building, are collectively referred to as direct containment heating (DCH). It is necessary to understand factors that enhance or mitigate DCH because the pressure load imposed on the RCB may lead to early failure of the containment.

  3. Computer modeling of lime-soda softening of cooling waters

    International Nuclear Information System (INIS)

    Chen, J.C.Y.

    1986-01-01

    A computer model is developed to fully describe the lime soda ash softening process. This process has a long history of being used to remove calcium and magnesium hardness from cooling waters in order to prevent scaling on heat exchangers. Softening of makeup water and/or a sidestream from the recirculating water will allow a reduction in blowdown. In the extreme case, zero blowdown may be accomplished to conserve cooling waters and to save the costs of disposing of blowdown. Cooling waters differ from most natural waters in having higher temperature and higher concentration of dissolved solids, and, therefore, a higher ionic strength. These factors plus the effects of complex formation are taken into consideration in the development of the computer model. To determine the composition of a softened water, the model assumes that an equilibrium state is reached in a reactor, and employs the equations of mass action and mass balance. The resulting nonlinear simultaneous equations are then linearized by Taylor series expansion and solved by the multidimensional Newton-Raphson method. The computer predictions are compared to the results of laboratory studies using synthetic waters

  4. Vibration phenomena in large scale pressure suppression tests

    International Nuclear Information System (INIS)

    Aust, E.; Boettcher, G.; Kolb, M.; Sattler, P.; Vollbrandt, J.

    1982-01-01

    Structure und fluid vibration phenomena (acceleration, strain; pressure, level) were observed during blow-down experiments simulating a LOCA in the GKSS full scale multivent pressure suppression test facility. The paper describes first the source related excitations during the two regimes of condensation oscillation and of chugging, and deals then with the response vibrations of the facility's wetwell. Modal analyses of the wetwell were run using excitation by hammer and by shaker in order to separate phenomena that are particular to the GKSS facility from more general ones, i.e. phenomena specific to the fluid related parameters of blowdown and to the geometry of the vent pipes only. The lowest periodicities at about 12 and 16 Hz stem from the vent acoustics. A frequency of about 36 to 38 Hz prominent during chugging seems to result from the lowest local models of two of the wetwell's walls when coupled by the wetwell pool. Further peaks found during blowdown in the spectra of signals at higher frequencies correspond to global vibration modes of the wetwell. (orig.)

  5. Simulation of Fission Product Liftoff Behavior During Depressurization Transients

    International Nuclear Information System (INIS)

    Tak, Nam-il; Yoon, Churl; Lee, Sung Nam

    2016-01-01

    As one of crucial technologies for the NHDD project, the development of the GAMMA-FP code is on-going. The GAMMA-FP code is targeted for fission product transport analysis under accident conditions. A well-known experiment named COMEDIE considered two important phenomena, i.e., fission product plateout and liftoff, for fission product transport within the primary circuit of a prismatic high temperature gas cooled reactor. The accumulated fission products on the structural material via the plateout can be liftoff during a blowdown phase after a pipe break accident. Since the fission product liftoff can increase a radioactivity risk, it is important to predict the amount of fission product liftoff during depressurization accidents. In this work, a model for fission product liftoff is implemented into the GAMMA-FP code and the GAMMA-FP code with the implemented model is validated using the COMEDIE blowdown test data. The results of GAMMA-FP show that the GAMMA-FP code can reliably simulate a pressure transient during blowdown phase after a pipe break accident. In addition, a reasonable amount of fission product liftoff was predicted by the GAMMA-FP code. The maximum difference between the measured and predicted liftoff fraction was less than a factor of 10. More in-depth study is required to increase the accuracy of prediction for a fission product liftoff

  6. Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.

    1983-01-01

    System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

  7. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

    2016-11-30

    between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. To predict the stability boundary theoretically, linear stability analysis in the frequency domain was performed at four sections of the natural circulation test loop. The flashing phenomena in the chimney section was considered as an axially uniform heat source. And the dimensionless characteristic equation of the pressure drop perturbation was obtained by considering the void fraction effect and outlet flow resistance in the core section. The theoretical flashing boundary showed some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium was recommended to improve the accuracy of flashing instability boundary. As another part of the funded research, flow instabilities of a PWR-type SMR under low pressure and low power conditions were investigated experimentally as well. The NuScale reactor design was selected as the prototype for the PWR-type SMR. In order to experimentally study the natural circulation behavior of NuScale iii reactor during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. The three-level scaling method is used as well to obtain the scaling ratios derived from various non-dimensional numbers. The design of the ideally scaled facility (ISF) was initially accomplished based on these scaling ratios. Then the engineering scaled facility (ESF) was designed and constructed based on the ISF by considering engineering limitations including laboratory space, pipe size, and pipe connections etc. PWR-type SMR experiments were performed in this well-scaled test facility to investigate the potential thermal hydraulic flow instability during the blowdown events, which might occur during the loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) of the prototype PWR

  8. Asbestos in cooling-tower waters

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.A.G.

    1977-12-01

    Fill material in natural- or mechanical-draft cooling towers can be manufactured from a variety of materials, including asbestos cement or asbestos paper. To aid in the environmental impact assessment of cooling towers containing these asbestos types of fill, information on these materials was obtained from cooling-tower vendors and users. Samples of makeup, basin, and blowdown waters at a number of operating cooling towers were obtained, and identification and enumeration of asbestos in the samples were performed by transmission electron microscopy, selected-area electron diffraction, and energy-dispersive x-ray analysis. Asbestos fibers were detected in cooling-tower water at 10 of the 18 sites sampled in the study. At all but three sites, the fibers were detected in cooling-tower basin or blowdown samples, with no fibers detected in the makeup water. The fibers were identified as chrysotile at all sites except one. Concentrations were on the order of 10/sup 6/ to 10/sup 8/ fibers/liter of water, with mass concentrations between <0.1 ..mu..g/liter to 37 ..mu..g/liter. The maximum concentrations of asbestos fibers in air near ground due to drift from cooling towers were estimated (using models) to be on the order of asbestos concentrations reported for ambient air up to distances of 4 km downwind of the towers. The human health hazard due to abestos in drinking-water supplies is not clear. Based on current information, the concentrations of asbestos in natural waters after mixing with cooling-tower blowdown containing 10/sup 6/ to 10/sup 8/ fibers/liter will pose little health risk. These conclusions may need to be revised if future epidemiological studies so indicate.

  9. Consequences in the pumps operation during a large loss of coolant accident

    International Nuclear Information System (INIS)

    Santos, G.A. dos; Sabundjian, G.

    1991-08-01

    The event of living on or turning off the operation of the Reactor Cooling Pumps - RCPs, in the case of a Loss of Coolant Accident - LOCA, has been a reason of a lot of studies after the Three Mile Island 2 accident. Thus, it was investigated a large break LOCA in the cold leg of Angra 1, with the RELAP4/MOD5 Code during the blowdown. The attained results indicated that the best performance of the core was in the case where the RCPs had been turned off in the beginning of the transient, when compared with different operation conditions of the RCPs. (author)

  10. The multi-dimensional module of CATHARE 2 description and application

    Energy Technology Data Exchange (ETDEWEB)

    Barre, F.; Dor, I.; Sun, C. [French Atomic Energy Commission (C.E.A.), Grenoble (France)

    1995-09-01

    In this paper, the three-dimensional module of CATHARE 2 is presented. It is based on a two-phase-flow six-equation model. A predictor/corrector multistep method, with an implicit behavior, is used to discretize the equations. Blowdown and boil-of analytical tests are used for an initial validation of the module. UPTF downcomer refill tests simulating the refill phase of a large-break loss-of-coolant accident are calculated. Additional models, including molecular and turbulent diffusion, are added in order to perform containment calculations.

  11. Quarterly technical report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, July--September 1975

    International Nuclear Information System (INIS)

    1976-02-01

    Light water reactor safety activities performed during July through September 1975 are summarized. The isothermal blowdown test series of the Semiscale Mod-1 test program has provided data for evaluation of break flow phenomena and analyses of piping flow regimes and pump performance. In the LOFT Program, measurement uncertainties were evaluated. The Thermal Fuels Behavior Program completed two power-cooling-mismatch tests on PWR-type fuel rods to investigate critical heat flux characteristics. Model development and verification efforts of the Reactor Behavior Program included development of the SPLEN1 computer code, subroutines for the FRAP-T code, verification of RELAP4, and results of the Halden Recycle Plutonium Experiment

  12. On the fine Simpson moduli spaces of 1-dimensional sheaves supported on plane quartics

    Directory of Open Access Journals (Sweden)

    Iena Oleksandr

    2018-02-01

    Full Text Available A parametrization of the fine Simpson moduli spaces of 1-dimensional sheaves supported on plane quartics is given: we describe the gluing of the Brill-Noether loci described by Drézet and Maican, provide a common parameter space for these loci, and show that the Simpson moduli space M = M4m ± 1(ℙ2 is a blow-down of a blow-up of a projective bundle over a smooth moduli space of Kronecker modules. Two different proofs of this statement are given.

  13. Feasibility Study of Coal Gasification/Fuel Cell/Cogeneration Project. Fort Greely, Alaska Site. Project Description,

    Science.gov (United States)

    1985-11-01

    Estimated Water Emissions 147 7-4 Estimated Solid Wastes 148 7-5 Composition of Blowdown from Stretford Process 149 1 I vi J 7862A LIST OF TABLES (Cont’d...due to the high CO2 concentration in the gas (26% Vol). Therefore, a Stretford liquid oxidation process was chosen for this plant. In this process , the...compliance with the sulfur emission levels of tne plant. A liquid phase oxidation Stretford Sulfur Removal Process is used for the removal of H2S to

  14. Feasibility Study of Coal Gasification/Fuel Cell/Cogeneration Project. Washington, DC Site. Project Description

    Science.gov (United States)

    1985-06-01

    Emissions 1:3 7-4 Estimated Solid Wastes 154 7-5 Composition of Blowdown from Stretford Process 155 vi 7398A I@ L1ST OF TABLES (Cont’d) Tab~le..Pae 7-6...alternative for the sulfur recovery process due to the high C02 concentration in the gas (24% Vol). Therefore, a Stretford liquid oxidation process was chosen...separated fromi the solution, which is regenerated by !i.r-Fparging and recycled. Because the Stretford process cannot remove COS, a hydrolysis step is

  15. Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

    2012-01-01

    Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

  16. Thermal-hydraulic considerations for particle bed reactors

    Science.gov (United States)

    Benenati, R.; Araj, K. J.; Horn, F.

    In the design of particle bed reactor (PBR) cores, consideration must be given to the gas coolant channels and their configuration. Neutronics analysis provides the relative volume fractions of the component materials, but these must be arranged in such a manner as to allow proper cooling of all components by the gas flow at relatively low pressure drops. The thermal hydraulic aspects of this problem are addressed. A description of the computer model used in the analysis of the steady state condition is also included. Blowdown tests on hot particle bed fuel elements were carried out and are described.

  17. Stress analysis of the LOFT modular DTT flowmeter for LOCE transients (L1-5 and L2-4)

    International Nuclear Information System (INIS)

    Mosby, W.R.

    1978-01-01

    An analysis is presented of combined stresses in the LOFT Modular DTT for specified temperature gradients. All combined stress intensities are shown to stay within applicable allowable stress intensities. A fatigue analysis is also presented which indicates that the LOFT Modular DTT will withstand 70,000 blowdown cycles. The LOFT Modular DTT is shown to meet the Class 1 stress requirments. A stress analysis of the tab region of the newly designed MDTT tab-type shroud is included. This stress analysis shows that the Class 1 stress requirements are met by the tab-type MDTT shroud design and that this design imposes no fatigue life limitation on the MDTT

  18. Aspects of Industrial Water Treatment.

    Science.gov (United States)

    1978-02-01

    Coagulation /Flocculation AI(OH)3, Fe(OE)3, Mn(OH) 2 , CaC12 !Chlorination ChloroalnesChlorinated organics 4 ’Boiler blowdown Phosphates, carbonates, tannin ...ad miteriag of water treatment operatious imould pro- vide the moesseery data te ea reesurcee to be coeerved sod to lower the cost of pellem aba*tmomt...Noncarbonate hardness as CsC0 3Odor Taste Trace organic defined by carbon chloroform extract (CCE) Boiler-feedvater and boiler water tests also include

  19. Function analysis of steam isolation valves

    International Nuclear Information System (INIS)

    Persson, R.; Kilpi, K.; Noro, H.; Siikonen, T.; Sjoeberg, A.; Wallen, G.; Aakesson, H.

    1981-01-01

    Function analysis of system-medium-operated steam isolation valves has been the objective of the Swedish-Finnish IVLS project, the results of which are presented in this report. Theoretical models were to be verified against available experimental data, to some extent from the HDR blowdown experiments, which are part of a German reactor safety program. Finnish hydraulic measurements on a valve model (scale 1:2.15) have been performed to give complementary data. The analysis work has covered the thermal-hydraulic behaviour of steam isolation valves as well as phenomena related to structural mechanics. Work performed under contract with the Swedish Nuclear Power Inspectorate. (Author)

  20. Function analysis of steam isolation valves

    International Nuclear Information System (INIS)

    Persson, R.; Sjoeberg, A.; Aakesson, H.; Kilpi, K.; Noro, H.; Siikonen, T.; Wallen, G.

    1981-01-01

    Function analysis of system-medium-operated steam isolation valves has been the objective of the Swedish-Finnish IVLS project, the results of which are presented in this report. Theoretical models were to be verified against available experimental data, to some extent from the HDR blowdown experiments, which are part of a German reactor safety program. Finnish hydraulic measurements on a valve model (scale 1:2.15) have been performed to give complementary data. The analysis work has covered the thermal-hydraulic behaviour of steam isolation valves as well as phenomena related to structural mechanics. (Auth.)

  1. Methods of the working processes modelling of an internal combustion engine by an ANSYS IC Engine module

    Science.gov (United States)

    Kurchatkin, I. V.; Gorshkalev, A. A.; Blagin, E. V.

    2017-01-01

    This article deals with developed methods of the working processes modelling in the combustion chamber of an internal combustion engine (ICE). Methods includes description of the preparation of a combustion chamber 3-d model, setting of the finite-element mesh, boundary condition setting and solution customization. Aircraft radial engine M-14 was selected for modelling. The cycle of cold blowdown in the ANSYS IC Engine software was carried out. The obtained data were compared to results of known calculation methods. A method of engine’s induction port improvement was suggested.

  2. Structure-dynamic model verification calculation of PWR 5 tests

    International Nuclear Information System (INIS)

    Engel, R.

    1980-02-01

    Within reactor safety research project RS 16 B of the German Federal Ministry of Research and Technology (BMFT), blowdown experiments are conducted at Battelle Institut e.V. Frankfurt/Main using a model reactor pressure vessel with a height of 11,2 m and internals corresponding to those in a PWR. In the present report the dynamic loading on the pressure vessel internals (upper perforated plate and barrel suspension) during the DWR 5 experiment are calculated by means of a vertical and horizontal dynamic model using the CESHOCK code. The equations of motion are resolved by direct integration. (orig./RW) [de

  3. Two-phase flow dynamics in ECC

    International Nuclear Information System (INIS)

    Albraaten, P.J.

    1981-07-01

    The present report summarizes the achievements within the project ''Two-phase Systems and ECC''. The results during 1978 - 1980 are accounted for in brief as they have been documented in earlier reports. The results during the first half of 1981 are accounted for in greater detail. They contain a new model for the Basset force and test runs with this model using the test code RISQUE. Furthermore, test runs have been performed with TRAC-PD2 MOD 1. This code was implemented on Edwards Pipe Blowdown experiment (a standard test case) and UC-Berkeley Reflooding experiment (a non-standard test case.) (Auth.)

  4. Application of a drift-flux model to flashing in straight pipes

    International Nuclear Information System (INIS)

    Hirt, C.W.; Romero, N.C.

    1975-06-01

    A new computer program, SOLA-OF, has been written to solve the unsteady, two-dimensional equations of motion for a two-phase mixture. The equations solved are based on the drift-flux approximation and include a phase transition model and a general drift velocity calculation. The SOLA-DF code is used for a study of the blowdown of straight pipes initially filled with water at high temperature and pressure. Computed results are presented that show the relative importance of phase transition rates, pipe friction, drift velocity magnitude, and other model variations. The computed results are also compared with experimental data. 7 references. (auth)

  5. Technical manual for COMET

    International Nuclear Information System (INIS)

    Song, Jin Ho; Kwon, Young Min; Kim, Taek Mo; Lee, Sang Jong; Jeong, Hae Yong

    1996-07-01

    The purpose of this report is to provide a description for a COMET computer code which is to be used in the analysis of mass and energy releases during post-blowdown phase of LOCA. The mass and energy data re to be used as input data for the containment functional design. This report contains a brief description of analytical models and guidelines for the usage of the computer code. This computer code is to be used for both cold leg and hot leg break analyses. A verification analyses are performed for Ulchin 3 and 4 cold and hot leg break. 11 figs (Author)

  6. Stress analysis of the LOFT modular DTT flowmeter for LOCE transients (L1-5 and L2-4)

    Energy Technology Data Exchange (ETDEWEB)

    Mosby, W.R.

    1978-08-16

    An analysis is presented of combined stresses in the LOFT Modular DTT for specified temperature gradients. All combined stress intensities are shown to stay within applicable allowable stress intensities. A fatigue analysis is also presented which indicates that the LOFT Modular DTT will withstand 70,000 blowdown cycles. The LOFT Modular DTT is shown to meet the Class 1 stress requirments. A stress analysis of the tab region of the newly designed MDTT tab-type shroud is included. This stress analysis shows that the Class 1 stress requirements are met by the tab-type MDTT shroud design and that this design imposes no fatigue life limitation on the MDTT.

  7. Investigation of fission product retention by the HTR containment building according to the vented confinement concept. Technical report 1.5

    International Nuclear Information System (INIS)

    Kreidler, R.

    1988-10-01

    The physical behaviour of aerosols during the transient primary gas blowdown phase and their long-time behavior through 10 hours after the beginning of a potential accident were investigated. Three scenarios were considered: pressure relief by way of an open failing safety valve; pressure relief by way of a safety valve after water ingress; severance type break of the pressure balancing pipe in the reactor cell. The deposition of aerosols was found to be decisively influenced by the composition of the atmosphere and the atmosphere mixture (pure air, He atmosphere), resp. in the containment building. (DG) [de

  8. Review of the GOTHIC code and trial application

    International Nuclear Information System (INIS)

    Lacroix, M.; Galanis, N.; Millette, J.

    1996-01-01

    A critical review of the performance of the generic computer code GOTHIC for the generation of thermalhydraulic information for containments was conducted. Several analyses were performed with GOTHIC to predict the flow behaviour and distribution of hydrogen concentration within containments whose geometrical complexity ranged from two simple interconnected rooms to a full scale reactor building. Sensitivity analysis studies were carried out to examine the effect of various modeling parameters. The implementation of physics by the code is reviewed and recommendations on its use for performing blowdown/hydrogen release analyses are made.(author) 5 refs., 9 tabs., 105 figs

  9. 1993 RCRA Part B permit renewal application, Savannah River Site: Volume 10, Consolidated Incineration Facility, Section C, Revision 1

    International Nuclear Information System (INIS)

    Molen, G.

    1993-08-01

    This section describes the chemical and physical nature of the RCRA regulated hazardous wastes to be handled, stored, and incinerated at the Consolidated Incineration Facility (CIF) at the Savannah River Site. It is in accordance with requirements of South Carolina Hazardous Waste Management Regulations R.61-79.264.13(a) and(b), and 270.14(b)(2). This application is for permit to store and teat these hazardous wastes as required for the operation of CIF. The permit is to cover the storage of hazardous waste in containers and of waste in six hazardous waste storage tanks. Treatment processes include incineration, solidification of ash, and neutralization of scrubber blowdown

  10. Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

    Energy Technology Data Exchange (ETDEWEB)

    Meservey, R.H.

    1976-04-01

    The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800/sup 0/F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests.

  11. LOCA verification and data bank

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.; Cox, N.D.; Atwood, C.L.; Madden, S.C.; Condie, K.G.

    1979-01-01

    The purpose of this task was to derive local conditions heat transfer parameters and their uncertainties using computer codes and experimentally derived boundary conditions. To accomplish this objective, Semiscale S-02-9 blowdown experiment was used along with the INVERT (an inverse heat conduction code) and RELAP4 (a thermal hydraulic code) codes as the analytical tools. The uncertainties calculated for the local conditions were limited to those introduced by inaccuracies in the experimentally measured boundary conditions. The propagation of the measurement uncertainties through the codes was investigated by varying the code input using statistical methods and a response surface technique

  12. LOCA verification and data bank. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Varacalle, Jr., D. J.; Cox, N. D.; Atwood, C. L.; Madden, S. C.; Condie, K. G.

    1979-01-01

    The purpose of this task was to derive local conditions heat transfer parameters and their uncertainties using computer codes and experimentally derived boundary conditions. To accomplish this objective, Semiscale S-02-9 blowdown experiment was used along with the INVERT (an inverse heat conduction code) and RELAP4 (a thermal hydraulic code) codes as the analytical tools. The uncertainties calculated for the local conditions were limited to those introduced by inaccuracies in the experimentally measured boundary conditions. The propagation of the measurement uncertainties through the codes was investigated by varying the code input using statistical methods and a response surface technique.

  13. Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1976-04-01

    The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800 0 F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests

  14. Bistable flow occurrence in the 2D model of a steam turbine valve

    Science.gov (United States)

    Pavel, Procházka; Václav, Uruba

    2017-09-01

    The internal flow inside a steam turbine valve was investigated experimentally using PIV measurement. The valve model was proposed to be two-dimensional. The model was connected to the blow-down wind tunnel. The flow conditions were set by the different position of the valve plug. Several angles of the diffuser by diverse radii were investigated concerning flow separation and flow dynamics. It was found that the flow takes one of two possible bistable modes. The first regime is characterized by a massive flow separation just at the beginning of the diffuser section on the one side. The second regime is axisymmetric and the flow separation is not detected at all.

  15. Supplmental testimony of the AEC Regulatory Staff. Public rulemaking hearing on: interim acceptance criteria for emergency core cooling systems for light-water cooled power reactors

    International Nuclear Information System (INIS)

    1972-01-01

    Information is presented concerning sensitivity analysis, loop codes, two-phase pressure drop, critical flow model, pump modeling, PWR core flow distribution, accumulator bypass, fuel densification, gap thermal conductance and UO 2 thermal conductivity, transition boiling heat transfer, clad-to-fluid heat transfer, heat transfer at low pressure, reflood rate analyses, containment back pressure during reflood, BWR FLECHT, PWR reflooding heat transfer FLECHT data, embrittlement and post-blowdown loads, fuel rod physico-chemical reactions, flow blockage, small break analysis, and decay heat. (U.S.)

  16. Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

    A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

  17. Relations between must clarification and organoleptic attributes of wine varietes

    Directory of Open Access Journals (Sweden)

    Vladimír Vietoris

    2014-02-01

    Full Text Available Blowdown musts is important operation performed in winemaking, which can have a major impact on the future quality of the wine. Blowdown of the wine removes components that may carry elements that negatively affect the hygienic and sensory quality of the wine. Fining of musts and wines is carried either by a static method or using different fining preparations. The aim of this work was to evaluate the effect of different methods of decanting on the wine quality varieties of Sauvignon. The overall sensory quality was evaluated (100 - points system, and semantic differential and the aromatic profile (profile method. All sensory evaluations were practiced by skilled sensory panel in controled conditions of Faculty sensory lab. Wine samples were clarified by static manner or with the assistance of the preparation applied to the clarification of wine in two different doses. By the results and their visualization of flavour and smell profile by spider plots we could conclude that pure cultures have positive effect on processed wine. Based on the results we found a beneficial effect of clearing by the clarification of the preparation based on cellulose, polyvinylpolypyrrolidone, gelatin and mineral adsorbents at 100 g.100 L-1  of the sensory quality of the wine.

  18. Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

    International Nuclear Information System (INIS)

    French, R.T.

    1975-08-01

    Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

  19. Reactor safety issues resolved by the 2D/3D program

    International Nuclear Information System (INIS)

    1995-09-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author)

  20. Reactor safety issues resolved by the 2D/3D Program

    International Nuclear Information System (INIS)

    Damerell, P.S.; Simons, J.W.

    1993-07-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

  1. Response of native and exotic bark beetles to high-energy wind event in the Tian Shan Mountains, Kazakhstan

    Science.gov (United States)

    Mukhamadiev, N.; Lynch, A.; O'Connor, C.; Sagitov, A.; Panyushkina, I. P.

    2012-12-01

    On May 17, 2011, the spruce forest of Yile-Alatausky and Medeo National Parks in southeast Kazakhstan was surged by a high-energy cyclonic storm. Severe blowdown damaged several thousand hectare of Tian Shan spruce forest (Picea schrenkiana), with over 90% of trees killed in extensive areas. Bark beetle populations are increasing rapidly, particularly Ips hauseri, I. typographis, I. sexdentatus, and Pityogenes perfossus (all Coleoptera: Curculionidae). Little is known about the frequency or extent of either large storm events or bark beetle outbreaks in the Tian Shan Mountains, nor about associations between outbreaks of these species and temperature and precipitation regimes. Local managers are concerned that triggering bark beetle outbreaks during current unusually warm, dry conditions will have devastating consequences for the residual forest and forest outside of the blowdown. We characterize the bark beetle population response to the 2011 event to date, and reconstruct the temporal and spatial dynamics of historical disturbance events in the area using dendrochronology. Additionally temperature and precipitation-sensitive tree-ring width chronologies from the Tian Shan Mountains are analyzed to determine high- and low-frequency variability of climate for the past 200 years. Catastrophic windstorm disturbances may play a crucial role in determining forest structure across the mountains. We hypothesize that the Tian Shan spruce forest could be prone to severe storm winds and subsequent bark beetle outbreaks and never reach an old-growth phase between events.

  2. PISCES 3DELK - a coupled Euler/Lagrange program for computing dynamic fluid-structure interactions in three dimensions

    International Nuclear Information System (INIS)

    Chu, H.Y.; Cowler, M.S.; Hancock, H.

    1983-01-01

    This paper describes the main features of PISCES 3DELK, a computer code that is used to solve complex three-dimensional fluid-structure interaction problems in reactor safety. These features include: an Eulerian finite difference scheme for calculating fluid flow and large distortions of solid media; a Langrange finite element scheme for calculating the response of thin structures; coupling of the Euler and Langrange schemes at fluid-structure interfaces. The code has been well validated and applied to a number of reactor safety analyses including blowdown in reactor primary vessels and components, and loadings on the secondary containment caused by a breach in the primary containment. Details of two analyses are presented in this paper. The first analysis is of blowdown in a pressurized water reactor caused by a cold leg break (the HDR experiment). Results of the PISCES 3DELK calculation are compared with results obtained by the K-FIX code. Agreement between the two calculations is good. The second analysis is of the depressurization caused by a feedwater pipe break in a steam generator of the CANDU reactor. Calculations have been performed which show that flexibility of internal components in the heat exchanger mitigate structural loadings. (orig.)

  3. Two-phase flow phenomena in broken recirculation line of BWR

    International Nuclear Information System (INIS)

    Kato, Masami; Arai, Kenji; Narabayashi, Tadashi; Amano, Osamu.

    1986-01-01

    When a primary recirculation line of BWR is ruptured, a primary recirculation pump may be subjected to very high velocity two-phase flow and its speed may be accelerated by this flow. It is important for safety evaluation to estimate the pump behavior during blowdown. There are two problems involved in analyzing this behavior. One problem concerns the pump characteristics under two-phase flow. The other involves the two-phase conditions at the pump inlet. If the rupture occurs at a suction side of the pump, choking is considered to occur at a broken jet pump nozzle. Then, a void fraction becomes larger downstream from the jet pump nozzle and volumetric flow through the pump will be very high. However, there is little experimental data available on two-phase flow downstream from a choking plane. Blowdown tests were performed using a simulated broken recirculation line and measured data were analyzed by TRAC-PlA. Analytical results agreed with measured data. (author)

  4. Validation of Effective Models for Simulation of Thermal Stratification and Mixing Induced by Steam Injection into a Large Pool of Water

    Directory of Open Access Journals (Sweden)

    Hua Li

    2014-01-01

    Full Text Available The Effective Heat Source (EHS and Effective Momentum Source (EMS models have been proposed to predict the development of thermal stratification and mixing during a steam injection into a large pool of water. These effective models are implemented in GOTHIC software and validated against the POOLEX STB-20 and STB-21 tests and the PPOOLEX MIX-01 test. First, the EHS model is validated against STB-20 test which shows the development of thermal stratification. Different numerical schemes and grid resolutions have been tested. A 48×114 grid with second order scheme is sufficient to capture the vertical temperature distribution in the pool. Next, the EHS and EMS models are validated against STB-21 test. Effective momentum is estimated based on the water level oscillations in the blowdown pipe. An effective momentum selected within the experimental measurement uncertainty can reproduce the mixing details. Finally, the EHS-EMS models are validated against MIX-01 test which has improved space and time resolution of temperature measurements inside the blowdown pipe. Excellent agreement in averaged pool temperature and water level in the pool between the experiment and simulation has been achieved. The development of thermal stratification in the pool is also well captured in the simulation as well as the thermal behavior of the pool during the mixing phase.

  5. Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

    International Nuclear Information System (INIS)

    Malang, S.

    1975-11-01

    An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

  6. A through calculation of 1,100 MWe PWR large break LOCA by THYDE-P1 EM model

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

    THYDE-P1 is a code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs). Up to now, THYDE-P1 has been applied to various experiment analyses, which show its high capability to analyze LOCAs as a best estimate (BE) calculation code. In this report, evaluation model (EM) calculation method, especialy in the blowdown and refill phases, is established equivalently to WREM/J2 which is regarded as appropriate for an EM calculation code, and the results of them are compared and discussed. The present calculation was the first executed by THYDE-P1-EM, and was performed as Sample Calculation Run 80 which was a part of a series of THYDE-P sample calculations. The calculation was carried out from the LOCA initiation till 400 seconds for a guillotine break at the cold leg of a commercial 1,100 MWe PWR plant. The calculated results agreed well to that of the WREM/J2 code. (author)

  7. Thermohydraulic behavior in a primary cooling system during a loss-of-coolant accident of a light-water reactor

    International Nuclear Information System (INIS)

    Shimamune, Hiroji; Shiba, Masayoshi; Adachi, Hiromichi; Suzuki, Norio; Okubo, Kaoru

    1975-12-01

    With ROSA-I (Rig of Safety Assessment - I), 61 runs of the LWR blowdown experiment have been carried out under the conditions: model reactor type, BWR and PWR; reactor core, none, no-heating and heating; rupture position, upper and lower pressure vessel nozzle; initial discharge pressure, 40, 70 and 100 kg/cm 2 G; and rupture diameter, 25, 50, 70, 100 and 125 mm. The purpose was to obtain the data of thermal and hydrodynamic behavior in the reactor pressure vessel during a blowdown, including in-vessel pressure, coolant temperature, discharge flow rate, model fuel rod surface temperature and shock wave. Analysis was also made with the codes RELAP-2 and -3 developed by NRTS of the United States, to verify the calculation model used. In addition, the results of calculation with the shockwave analysis code DEPCO developed in JAERI were compared with those by experiment. The experimental facility ROSA-I and the results obtained with it and also the analyses made in this connection, are described in detail. (auth.)

  8. Reactor safety issues resolved by the 2D/3D Program. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Damerell, P.S.; Simons, J.W. [eds.] [MPR Associates, Inc., Washington, DC (United States)

    1993-07-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated.

  9. Test results employed by General Electric for boiling water reactor containment and vertical vent loads

    International Nuclear Information System (INIS)

    Fukushima, T.Y.; Singh, A.; James, A.J.; Winkler, W.D.; Walenciak, M.R.; Rosa, J.M.

    1975-10-01

    During a safety relief valve blowdown, air contained in the relief line discharges into the suppression pool with the resulting oscillations of the air bubble causing dynamic loading on the containment. The magnitude and characteristics of such loading depend upon the geometry of the discharge device at the end of the safety relief line. Extensive small scale and large scale testing was performed to evaluate the performance of a four-arm quencher discharge device. Results of these tests, description of test facility, instrumentation and test procedures are described. During a loss-of-coolant accident, steam flows through vertical vent pipes such as employed in Mark I and II Containments and condenses in the suppression pool at the vent exit. During this condensation process, a steam bubble which forms at the vent exit will collapse irregularly leading to water impingement on the vent pipe. The water impingement phenomenon causes lateral loading on the vertical vents. The loading phenomena and series of tests performed to evaluate the load magnitudes are described. During a later part of the safety relief valve blowdown, steam discharges into the suppression pool through the safety relief line end discharge device. Extensive tests were carried out to investigate the high temperature condensation phenomenon and the temperature threshold limits for the occurrence of condensation vibrations for various configurations including the quencher configuration, of the relief line and discharge device. Results of these tests including a description of the test facility, instrumentation and test procedures have been included

  10. Coupled calculation of the radiological release and the thermal-hydraulic behavior of a 3-loop PWR after a SGTR by means of the code RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Van Hove, W.; Van Laeken, K.; Bartsoen, L. [Belgatom, Brussels (Belgium)] [and others

    1995-09-01

    To enable a more realistic and accurate calculation of the radiological consequences of a SGTR, a fission product transport model was developed. As the radiological releases strongly depend on the thermal-hydraulic transient, the model was included in the RELAP5 input decks of the Belgian NPPs. This enables the coupled calculation of the thermal-hydraulic transient and the radiological release. The fission product transport model tracks the concentration of the fission products in the primary circuit, in each of the SGs as well as in the condenser. This leads to a system of 6 coupled, first order ordinary differential equations with time dependent coefficients. Flashing, scrubbing, atomisation and dry out of the break flow are accounted for. Coupling with the thermal-hydraulic calculation and correct modelling of the break position enables an accurate calculation of the mixture level above the break. Pre- and post-accident spiking in the primary circuit are introduced. The transport times in the FW-system and the SG blowdown system are also taken into account, as is the decontaminating effect of the primary make-up system and of the SG blowdown system. Physical input parameters such as the partition coefficients, half life times and spiking coefficients are explicitly introduced so that the same model can be used for iodine, caesium and noble gases.

  11. Experimental investigation on the behaviour of pressure suppression containment systems by the SOPRE-1 facility

    International Nuclear Information System (INIS)

    Cerullo, N.; Delli Gatti, A.; Marinelli, M.; Mazzini, M.; Mazzoni, A.; Sbrana, A.; Todisco, P.

    1977-01-01

    The SOPRE-1 test facility is an integral model (scale 1:13) of a MARK II pressure suppression containment system. It was set up at the University of Pisa in order to study the pressure-temperature transient in pressure suppression containment systems during LOCAs. Knowledge of this transient is necessary to perform a correct structural analysis of reactor containment. The containment system behaviour is studied by changing the principal parameters which affect the transient (blow-down mass and energy release, suppression pool water temperature, vent pipe number and submergence heat transfer coefficients). The first series of tests involved: A) 13 tests with break area of 1.8 cm 2 , B) 8 tests with break area of 20.0 cm 2 . The following experimental conditions were changed: - position of the simulated break (from liquid or steam zone), - water pressure (20-85 Kgsub(p)/cm 2 ) and mass (45-70Kg) in the vessel model. Tests A): the CONTEMPT codes correctly forecast the pressure-temperature history, both in dry- and in wet-well. Tests B): the experimental runs have shown that increasing of blow-down flowrate produces dry-well pressure spatial differences and anomalous vent pipe behaviour. This results in damped oscillations of dry- and wet-well pressure, probably due to alterbating air bubble over-expansion and collapse, and in vent pipe opening and reclosing. (Auth.)

  12. Sensitivity of break-flow-partition on the containment pressure and temperature

    International Nuclear Information System (INIS)

    Kwon, Young Min; Song, Jin Ho; Lee, Sang Yong

    1994-01-01

    For the case of RCS blowdown into the vapor region of a containment at low pressure, the blowdown mixture will start to boil at the containment pressure and liquid will separate from the flow near the break location. The flashed steam is added to the containment atmosphere and liquid is falled to the sump. Analytically, the break flow can be divided into steam and liquid in a number of ways. Discussed in this study is three partition models and Instantaneous Mixing(IM) Model employed in different containment analysis computer codes. IM model is employed in the CONTRANS code developed by ABB-CE for containment thermodynamic analysis. The various partition models were applied to the double ended discharge leg slot break (DEDLS) LOCA which is containment design base accident (CDBA) for Ulchin 3 and 4 PSAR. It was shown that IM model is the most conservative for containment design pressure analysis. Results of the CONTRANS analyses are compared with those of UCN PSAR for which CONTEMPT-LT code was used

  13. NRC Information No. 90-18: Potential problems with Crosby safety relief valves used on diesel generator air start receiver tanks

    International Nuclear Information System (INIS)

    Rossi, C.E.

    1992-01-01

    On March 31, 1989, Cooper Industries was made aware of circumstances at Perry Unit 1 that led to the Division I EDG being declared inoperable. A Crosby safety relief valve on one of the two EDG starting air receiving tanks was inadvertently hit during maintenance activities. The force of the impact caused the valve to open and blow down both air receiving tanks. The safety relief valve did not reseat until approximately 30 psig below the EDG automatic start lockout signal. On January 12, 1990, Cooper Industries learned that a similar event had occurred at Comanche Peak. On January 17, 1990, Cooper Industries submitted a 10 CFR Part 21 report on the affected safety relief valves (Crosby style JMBU and JRU safety relief valves). Although Crosby-style JMBU and JRU safety relief valves were designed to meet the requirements of Section VIII of the ASME Boiler and Pressure Vessel Code, they were not seismically qualified. In addition, the blowdown characteristics of the valves were not consistent with the functional requirements of the system in which they were installed. Cooper Industries has recommended replacing these valves with seismically qualified valves that have the proper blowdown reseat characteristics

  14. Piping hydrodynamic loads for a PWR power up-rate with steam generator replacement

    International Nuclear Information System (INIS)

    Julie M Jarvis; Allen T Vieira; James M Gilmer

    2005-01-01

    Full text of publication follows: Pipe break hydrodynamic loads are calculated for various systems in a PWR for a Power Up-rate (PUR) with a Steam Generator Replacement (SGR). PUR with SGR can change the system pressures, mass flowrates and pipe routing/configuration. These changes can alter the steam generator piping steam/water hammer loads. This paper discusses the need to benchmark against the original design basis, the use of different modeling techniques, and lessons learned. Benchmarking for licensing in the United States is vital in consideration of 10CFR50.59 and other licensing and safety issues. RELAP5 and its force post-processor R5FORCE are used to model the transient loads for various piping systems such as main feedwater and blowdown systems. Other modeling applications, including the Bechtel GAFT program, are used to evaluate loadings in the main steam piping. Forces are calculated for main steam turbine stop valve closure, feedwater pipe breaks and subsequent check valve slam, and blowdown isolation valve closure. These PUR/SGR forces are compared with the original design basis forces. Modeling techniques discussed include proper valve closure modeling, sonic velocity changes due to pipe material changes, and two phase flow effects. Lessons learned based on analyses done for several PWR PUR with SGR are presented. Lessons learned from these analyses include choosing the optimal replacement piping size and routing to improve system performance without resulting in excessive piping loads. (authors)

  15. Application of Pulse Spark Discharges for Scale Prevention and Continuous Filtration Methods in Coal-Fired Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young; Fridman, Alexander

    2012-06-30

    The overall objective of the present work was to develop a new scale-prevention technology by continuously precipitating and removing dissolved mineral ions (such as calcium and magnesium) in cooling water while the COC could be doubled from the present standard value of 3.5. The hypothesis of the present study was that if we could successfully precipitate and remove the excess calcium ions in cooling water, we could prevent condenser-tube fouling and at the same time double the COC. The approach in the study was to utilize pulse spark discharges directly in water to precipitate dissolved mineral ions in recirculating cooling water into relatively large suspended particles, which could be removed by a self-cleaning filter. The present study began with a basic scientific research to better understand the mechanism of pulse spark discharges in water and conducted a series of validation experiments using hard water in a laboratory cooling tower. Task 1 of the present work was to demonstrate if the spark discharge could precipitate the mineral ions in water. Task 2 was to demonstrate if the selfcleaning filter could continuously remove these precipitated calcium particles such that the blowdown could be eliminated or significantly reduced. Task 3 was to demonstrate if the scale could be prevented or minimized at condenser tubes with a COC of 8 or (almost) zero blowdown. In Task 1, we successfully completed the validation study that confirmed the precipitation of dissolved calcium ions in cooling water with the supporting data of calcium hardness over time as measured by a calcium ion probe. In Task 2, we confirmed through experimental tests that the self-cleaning filter could continuously remove precipitated calcium particles in a simulated laboratory cooling tower such that the blowdown could be eliminated or significantly reduced. In addition, chemical water analysis data were obtained which were used to confirm the COC calculation. In Task 3, we conducted a series

  16. Development and validation of effective models for simulation of stratification and mixing phenomena in a pool of water

    International Nuclear Information System (INIS)

    Li, H.; Kudinov, P.; Villanueva, W.

    2011-06-01

    This work pertains to the research program on Containment Thermal-Hydraulics at KTH. The objective is to evaluate and improve performance of methods, which are used to analyze thermal-hydraulics of steam suppression pools in a BWR plant under different abnormal transient and accident conditions. The pressure suppression pool was designed to have the capability as a heat sink to cool and condense steam released from the core vessel and/or main steam line during loss of coolant accident (LOCA) or opening of safety relief valve in normal operation of BWRs. For the case of small flow rates of steam influx, thermal stratification could develop on the part above the blowdown pipe exit and significantly impede the pool's pressure suppression capacity. Once steam flow rate increases significantly, momentum introduced by the steam injection and/or periodic expansion and collapse of large steam bubbles due to direct contact condensation can destroy stratified layers and lead to mixing of the pool water. We use CFD-like model of the general purpose thermal-hydraulic code GOTHIC for addressing the issues of stratification and mixing in the pool. In the previous works we have demonstrated that accurate and computationally efficient prediction of the pool thermal-hydraulics in the scenarios with transition between thermal stratification and mixing, presents a computational challenge. The reason is that direct contact condensation phenomena, which drive oscillatory motion of the water in the blowdown pipes, are difficult to simulate with original GOTHIC models because of appearance of artificial oscillations due to numerical disturbances. To resolve this problem we propose to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) model and the Effective Momentum Source (EMS) model. We use POOLEX/PPOOLEX experiment (Lappeenranta University of Technology in Finland), in order to (a) quantify errors due to GOTHIC's physical models

  17. Unusual occurrences during the whole operation of BN-250 NPP

    International Nuclear Information System (INIS)

    Andropenkov, S.

    2000-01-01

    Unusual occurrences during the whole operation BN-350 NPP. 1. Oil ingress in high pressure receiver for the not reveled reason, 12.05.1994. 2. lncrease of water radioactivity of circulating water supply system due to heat exchanger leak of spent fuel assembly washing out system, 17.09.1993. 3. Lack of passableness of sodium drain header of primary circuit reveled during inspection on scheduled preventative maintenance, 28.11.1996. 4. Destruction of the blow-off line of MCP-6 due to corrosion damage of the pipeline while unit was being operated at rated power, 23.04.1993. 5. Lack of passableness of blow-down pipeline connecting reactor gas cover with gas-type pressurizer while unit was being operated at rated power, 17.11.1994. 6. Sodium ingress in blow-down pipeline of loop-5 intermediate heat exchanger while loop-5 was being fed of sodium during scheduled preventative maintenance, 27.06.1994. 7. Resistance deterioration of electro heating zones of loop-4 due to heat exchanger leak and water ingress in air-pipeline of primary circuit boxes recirculating air system, 02.05.1997. 8. Resistance deterioration of electro heating zones of sodium drain header of secondary circuit was sopped in the water for the extinguishing the fire of blowing ventilation oil-strainer, 23.12.1994. 9. Sodium ingress in gas-type pressurizer through pipeline of primary sodium cleanup system and blow-down pipeline of failed MCP-2 while primary sodium cleanup system was being connected to the primary circuit, 17.08.1976. As a rule, the main reactor systems are scrutinized more carefully than the auxiliary reactor systems and the order actions are existed for eliminating and mitigating of consequences of main reactor system fails. Therefore the auxiliary reactor system fails may impact on the main reactor systems through places of its contact in significant measure. The influence of auxiliary reactor system fails on main reactor systems and its possible consequences for behavior of the main

  18. Development and validation of effective models for simulation of stratification and mixing phenomena in a pool of water

    Energy Technology Data Exchange (ETDEWEB)

    Li, H.; Kudinov, P.; Villanueva, W. (Royal Institute of Technology (KTH). Div. of Nuclear Power Safety (Sweden))

    2011-06-15

    This work pertains to the research program on Containment Thermal-Hydraulics at KTH. The objective is to evaluate and improve performance of methods, which are used to analyze thermal-hydraulics of steam suppression pools in a BWR plant under different abnormal transient and accident conditions. The pressure suppression pool was designed to have the capability as a heat sink to cool and condense steam released from the core vessel and/or main steam line during loss of coolant accident (LOCA) or opening of safety relief valve in normal operation of BWRs. For the case of small flow rates of steam influx, thermal stratification could develop on the part above the blowdown pipe exit and significantly impede the pool's pressure suppression capacity. Once steam flow rate increases significantly, momentum introduced by the steam injection and/or periodic expansion and collapse of large steam bubbles due to direct contact condensation can destroy stratified layers and lead to mixing of the pool water. We use CFD-like model of the general purpose thermal-hydraulic code GOTHIC for addressing the issues of stratification and mixing in the pool. In the previous works we have demonstrated that accurate and computationally efficient prediction of the pool thermal-hydraulics in the scenarios with transition between thermal stratification and mixing, presents a computational challenge. The reason is that direct contact condensation phenomena, which drive oscillatory motion of the water in the blowdown pipes, are difficult to simulate with original GOTHIC models because of appearance of artificial oscillations due to numerical disturbances. To resolve this problem we propose to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) model and the Effective Momentum Source (EMS) model. We use POOLEX/PPOOLEX experiment (Lappeenranta University of Technology in Finland), in order to (a) quantify errors due to GOTHIC

  19. ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3.06.6B - transient film boiling in upflow

    International Nuclear Information System (INIS)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

  20. Experiment prediction for Loft Nonnuclear Experiment L1-4

    International Nuclear Information System (INIS)

    White, J.R.; Berta, V.T.; Holmstrom, H.L.O.

    1977-04-01

    A computer analysis, using the WHAM and RELAP4 computer codes, was performed to predict the LOFT system thermal-hydraulic response for Experiment L1-4 of the nonnuclear (isothermal) test series. Experiment L1-4 will simulate a 200 percent double-ended offset shear in the cold leg of a four-loop large pressurized water reactor. A core simulator will be used to provide a reactor vessel pressure drop representative of the LOFT nuclear core. Experiment L1-4 will be initiated with a nominal isothermal primary coolant temperature of 282.2 0 C, a pressurizer pressure of 15.51 MPa, and a primary coolant flow of 270.9 kg/s. In general, the predictions of saturated blowdown for Experiment Ll-4 are consistent with the expected system behavior, and predicted trends agree with results from Semiscale Test S-01-4A, which simulated the Ll-4 experiment conditions

  1. Experimental studies on transient water-steam impinging jet

    International Nuclear Information System (INIS)

    Kitade, Kozo; Nakatogawa, Tetsundo; Nishikawa, Hideo; Kawanishi, Kohei; Tsuruto, Chuichi.

    1980-01-01

    Blowdown experiments were carried out in order to clarify pipe reaction forces and jet forces at hypothetical pipe break accident in PWR. The experiments were carried out at the initial pressure of about 70 and 150 kg/cm 2 .G with subcooling temperature of 13 -- 41 0 C. The reaction force has a maximum value just after the rupture in such a manner to attain abruptly to a peak and gradually decreases after that time in proportion to the inner pressure of the pipe. A plane board was used as a target, on which two-phase flow jet impinged vertically. A distribution of pressure on the target is most wide just after break. On the other hand, the pressure has a maximum value after a short period of time from the rupture. (author)

  2. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

    The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

  3. Out-of-pile safety studies performed for light water reactors

    International Nuclear Information System (INIS)

    Courtaud, M.

    1975-01-01

    The out-of-pile experimental program performed at the Heat Transfer Laboratory of the CEN-Grenoble in view of constructing the physical models and correlations employed in the computer codes used for describing loss-of-coolant accidents in light water reactors is presented. This program is mainly concerned - for the blowdown phase - with studies of heat transfer in rod bundles performed in the OMEGA loop (4,5MW 170 bars) and of critical flow and flashing in the Moby Dick (10 bars, 360kW) and Canon facilities; -for the core reflooding phase- with the study of heat transfer in rod bundles and rewetting of clads performed on the ERSEC loop (450kW, 6 bars) [fr

  4. Thermalhydraulics and safety-related experiments in support of the Canadian SCWR concept development

    International Nuclear Information System (INIS)

    Leung, L.K.H.

    2014-01-01

    A Canadian National Program has been established for R&D in various technology areas in support of the Canadian Super-Critical Water-cooled Reactor (SCWR) concept development. The thermalhydraulics and safety projects in the National Program had led to the establishment of infrastructure and capabilities at universities. Current projects focus on the applied research to provide relevant experimental databases for developing or validating prediction methods and analytical toolsets. Several experimental projects are currently being carried out to obtain heat transfer data with annuli, 3-rod assembly, and 4-rod assembly in refrigerant-134a flow, carbon dioxide flow, and water flow, respectively, and blow-down and natural-circulation data with tubes in water and carbon dioxide flow, respectively. Key scopes of these experimental projects are described. Experimental data obtained from these projects are presented. (author)

  5. ROSA-III 200% double-ended break integral test RUN 901

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Tasaka, Kanji; Koizumi, Yasuo; Anoda, Yoshinari; Kumamaru, Hiroshige; Murata, Hideo; Suzuki, Mitsuhiro; Shiba, Masayoshi

    1984-02-01

    This report presents the experimental data of RUN 901, a 200% double-ended break test at the recirculation pump suction line with the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The channel inlet flows were measured by differential pressure transducers installed at the channel inlet orifices of the fuel assembly No.4. The PCT (Peak Cladding Temperature) was 780 K occured during the blowdown phase in RUN 901. The whole core was quenched after the ECCS actuation and the effectiveness of ECCS has been confirmed. (author)

  6. Posttest REALP4 analysis of LOFT experiment L1-3A

    International Nuclear Information System (INIS)

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

    1977-10-01

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

  7. CHEMCON User's Manual, Version 3.1

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.

    1995-09-01

    CHEMCON is a computer program developed to analyze thermal transients of tokamak fusion reactors. It contains a one dimensional, cylindrical geometry, conduction model that allows a variety of heat transfer modes within nodes and at node boundaries. Solid regions can be grouped into segments that communicate at their boundaries through a radiation enclosure model. CHEMCON includes a single volume, pressurization/condensation model that is used to include the effects of an in-vessel LOCA and the resulting heat transfer between hot surfaces and cold surfaces in contact with this volume. The code includes properties for 11 solid materials and two gases. CHEMCON also contains specialized models for modeling chemical reactions of node boundaries with air and steam including the gases produced from these reactions. In addition, a model treating the collapse of radiation shields within a gap is also included. CHEMCON is used mainly to simulate the thermal transient for post-blowdown loss-of-coolant-accidents

  8. Fluid dynamics and heat transfer methods for the TRAC code

    International Nuclear Information System (INIS)

    Reed, W.H.; Kirchner, W.L.

    1977-01-01

    A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, we have developed new highly implicit difference techniques that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained. (author)

  9. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  10. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  11. PDE Nozzle Optimization Using a Genetic Algorithm

    Science.gov (United States)

    Billings, Dana; Turner, James E. (Technical Monitor)

    2000-01-01

    Genetic algorithms, which simulate evolution in natural systems, have been used to find solutions to optimization problems that seem intractable to standard approaches. In this study, the feasibility of using a GA to find an optimum, fixed profile nozzle for a pulse detonation engine (PDE) is demonstrated. The objective was to maximize impulse during the detonation wave passage and blow-down phases of operation. Impulse of each profile variant was obtained by using the CFD code Mozart/2.0 to simulate the transient flow. After 7 generations, the method has identified a nozzle profile that certainly is a candidate for optimum solution. The constraints on the generality of this possible solution remain to be clarified.

  12. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo O.; Vertullo, Alicia; Schlamp, Miguel A.; Garcia, Alicia E.

    2000-01-01

    The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model. The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator. As a consequence and due to reactor features the core power is also increased. As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided. Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System. In all the sequences the DNBR and CPR remain above the minimum safety values established by design. Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised. (author)

  13. Fluid transients in fluid-structure interaction - 1987; Proceedings of the Third Symposium, Boston, MA, Dec. 13-18, 1987

    Science.gov (United States)

    Dodge, F. T.; Moody, F. J.

    Papers are presented on a three-dimensional analysis of liquid oxygen sloshing in the Space Shuttle external tank, the flow-induced oscillations of a novel double-wing spring-mass system, added mass and damping coefficents for a hexagonal cylinder, and a new hydraulic pressure intensifier using an oil hammer. Other topics include junction losses in pulsating flow, a finite element analysis of a slender fluid-structure system, two-phase blowdown through a short tube, and check valve behavior under transient flow conditions. Also considered are forces in initially empty pipes subject to rapid filling, a modal analysis of vibrations in liquid-filled piping systems, efficient computation of the pipeline break problem, and fluid dynamics associated with ductile pipeline fracture.

  14. Development and applications of the interim direct heating model for the CONTAIN computer code

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Carroll, D.E.; Tills, J.L.; Washington, K.E.; Williams, D.C.

    1987-01-01

    A model for Direct Containment Heating has been developed for the CONTAIN code which takes account of the competing processes of melt ejection, steam blowdown, heat transfer among debris, gas and walls, chemical reactions in the debris droplets, intercell gas and debris transport, and de-entrainment of the suspended debris. Calculations of accident sequences with the model have shown that a qualitatively different understanding of the Direct Heating phenomenon emerges when more detailed multi-cell calculations are performed within a system-level containment code, compared to the relatively simplistic single cell, adiabatic calculations which have been used in the past. Results are presented for the TMLB sequence at the Surry and Sequoyah plants, and a sensitivity study for some of the most important uncertain parameters is described

  15. Method and apparatus for monitoring two-phase flow. [PWR

    Science.gov (United States)

    Sheppard, J.D.; Tong, L.S.

    1975-12-19

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods.

  16. Apparatus for monitoring two-phase flow

    Science.gov (United States)

    Sheppard, John D.; Tong, Long S.

    1977-03-01

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods.

  17. Effect of water treatment on the comparative costs of evaporative and dry cooled power plants

    International Nuclear Information System (INIS)

    Gold, H.; Goldstein, D.J.; Yung, D.

    1976-07-01

    The report presents the results of a study on the relative cost of energy from a nominal 1000 Mwe nuclear steam electric generating plant using either dry or evaporative cooling at four sites in the United States: Rochester, New York; Sheridan, Wyoming; Gallup, New Mexico and Dallas, Texas. Previous studies have shown that because of lower efficiencies the total annual evaluated costs for dry cooling systems exceeds the total annual evaluated costs of evaporative cooling systems, not including the cost of water. The cost of water comprises the cost of supplying the makeup water, the cost of treatment of the makeup and/or the circulating water in the tower, and the cost of treatment and disposal of the blowdown in an environmentally acceptable manner. The purpose of the study is to show the effect of water costs on the comparative costs of dry and evaporative cooled towers

  18. Kinetic—a system code for analyzing nuclear thermal propulsion rocket engine transients

    Science.gov (United States)

    Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans

    1993-01-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel, coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of controls element (drums or rods). The worth of the control element and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.

  19. KINETIC: A system code for analyzing Nuclear thermal propulsion rocket engine transients

    Science.gov (United States)

    Schmidt, E.; Lazareth, O.; Ludewig, H.

    1993-07-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of control elements (drums or rods). The worth of the control clement and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.

  20. Steamgenerators corrosion monitoring and chemical cleanings

    International Nuclear Information System (INIS)

    Otchenashev, G.

    2001-01-01

    One of the most important secondary side water chemistry objectives is optimization of chemistry conditions to reduce materials corrosion and their products transport into steam generators. Corrosion products (mainly iron and copper oxides) can form deposits on the SG's tubes and essentially decrease their operating resource. The transport of corrosion products by the constant flowrate of feed and blowdown water depends only on their content in these streams. All the internal surfaces (walls, collectors, tubes) were covered with the tough deposit firmly connected with the surface. Corrosion under this deposit was not detected. In some places sludge unconnected with the surface was detected. The lower tubes are located the more unconnected sludge was detected. On SG bottom near the hatch the sludge thickness was about 3 cm. (R.P.)

  1. High-repetition-rate PIV investigations on a generic rocket model in sub- and supersonic flows

    Science.gov (United States)

    Bitter, Martin; Scharnowski, Sven; Hain, Rainer; Kähler, Christian J.

    2011-04-01

    High-repetition-rate PIV measurements were performed in the trisonic wind tunnel facility at the Bundeswehr University Munich in order to investigate the boundary layer parameters on a generic rocket model and the recirculation area in the wake of the model at Mach numbers up to Mach = 2.6. The data are required for the validation of unsteady flow simulations. Because of the limited run time of the blow-down wind tunnel, a high-repetition-rate PIV system was applied to obtain the flow statistics with high accuracy. The results demonstrate this method's potential to resolve small-scale flow phenomena over a wide field of view in a large Mach number range but also show its limitations for the investigations of wall-bounded flows.

  2. Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

    International Nuclear Information System (INIS)

    Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

    1987-01-01

    From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 MPa were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. (orig./GL)

  3. Adsorption process to recover hydrogen from feed gas mixtures having low hydrogen concentration

    Science.gov (United States)

    Golden, Timothy Christopher; Weist, Jr., Edward Landis; Hufton, Jeffrey Raymond; Novosat, Paul Anthony

    2010-04-13

    A process for selectively separating hydrogen from at least one more strongly adsorbable component in a plurality of adsorption beds to produce a hydrogen-rich product gas from a low hydrogen concentration feed with a high recovery rate. Each of the plurality of adsorption beds subjected to a repetitive cycle. The process comprises an adsorption step for producing the hydrogen-rich product from a feed gas mixture comprising 5% to 50% hydrogen, at least two pressure equalization by void space gas withdrawal steps, a provide purge step resulting in a first pressure decrease, a blowdown step resulting in a second pressure decrease, a purge step, at least two pressure equalization by void space gas introduction steps, and a repressurization step. The second pressure decrease is at least 2 times greater than the first pressure decrease.

  4. Hydraulic modeling of thermal discharges into shallow, tidal affected streams

    International Nuclear Information System (INIS)

    Copp, H.W.; Shashidhara, N.S.

    1981-01-01

    A two-unit nuclear fired power plant is being constructed in western Washington state. Blowdown water from cooling towers will be discharged into the Chehalis River nearby. The location of a diffuser is some 21 miles upriver from Grays Harbor on the Pacific Ocean. Because the Chehalis River is classified as an excellent stream from the standpoint of water quality, State regulatory agencies required demonstration that thermal discharges would maintain water quality standards within fairly strict limits. A hydraulic model investigation used a 1:12 scale, undistorted model of a 1300-foot river reach in the vicinity of the diffuser. The model scale was selected to insure fully turbulent flows both in the stream and from the diffuser (Reynolds similitude). Model operation followed the densimetric Froude similitude. Thermistors were employed to measure temperatures in the model; measurements were taken by computer command and such measurements at some 250 positions were effected in about 2.5 seconds

  5. Experience in KINS on Best Estimate Calculation with Uncertainty Evaluation

    International Nuclear Information System (INIS)

    Bang, Young Seok; Huh, Byung-Gil; Cheong, Ae-ju; Woo, Sweng-Woong

    2013-01-01

    In the present paper, experience of Korea Institute of Nuclear Safety (KINS) on Best Estimate (BE) calculation and uncertainty evaluation of large break loss-of-coolant accident (LB LOCA) of Korean Pressurized Water Reactor (PWR) with various type of Emergency Core Cooling System (ECCS) design is addressed. Specifically, the current status of BE code, BE calculations and uncertainty parameters and related approaches are discussed. And the specific problem such as how to recover the difficulty in treating the uncertainty related to the phenomena specific to ECCS design (e.g., upper plenum injection phenomena) is discussed. Based on the results and discussion, it is found that the present KINS-REM has been successfully developed and applied to the regulatory auditing calculations. Need of further study includes the improvement of reflood model of MARS code, uncertainty of blowdown quenching, and reconsideration of the unconcerned model and fuel conductivity degradation with burnup. (authors)

  6. Transition phase in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.; Henninger, R.J.; Jackson, J.F.

    1976-01-01

    Mechanistic analyses of transient-under-cooling accidents have led in some cases to a mild initiating phase instead of a direct hydrodynamic disassembly of the core. The fuel is then trapped in the core by the strong mechanical surroundings and blockages formed by refrozen cladding steel and/or fuel. The formation of fuel blockages has been verified experimentally. The bottled-up core will boil on fission and decay heat, with steel as the working fluid. Boil-up in a churn turbulent flow regime may prevent recriticality due to fuel recompaction. Ultimate fuel removal from the core is probably by a two-phase blow-down after permanent leakage paths are opened. However, a vigorous recriticality can not be precluded. Reactors with void coefficients larger than that in CRBR are more likely to disassemble in the initiating phase, so the transition phase may be unique to small cores

  7. Wake Management Strategies for Reduction of Turbomachinery Fan Noise

    Science.gov (United States)

    Waitz, Ian A.

    1998-01-01

    The primary objective of our work was to evaluate and test several wake management schemes for the reduction of turbomachinery fan noise. Throughout the course of this work we relied on several tools. These include 1) Two-dimensional steady boundary-layer and wake analyses using MISES (a thin-shear layer Navier-Stokes code), 2) Two-dimensional unsteady wake-stator interaction simulations using UNSFLO, 3) Three-dimensional, steady Navier-Stokes rotor simulations using NEWT, 4) Internal blade passage design using quasi-one-dimensional passage flow models developed at MIT, 5) Acoustic modeling using LINSUB, 6) Acoustic modeling using VO72, 7) Experiments in a low-speed cascade wind-tunnel, and 8) ADP fan rig tests in the MIT Blowdown Compressor.

  8. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

  9. ANALYSIS OF A STATION BLACKOUT SCENARIO WITH AN ATLAS TEST

    Directory of Open Access Journals (Sweden)

    YEON-SIK KIM

    2013-04-01

    Full Text Available A station blackout experiment called SBO-01 was performed at the ATLAS facility. From the SBO-01 test, the station blackout scenario can be characterized into two typical phases: A first phase characterized by decay heat removal through secondary safety valves until the SG dryouts, and a second phase characterized by an energy release through a blowdown of the primary system after the SG dryouts. During the second phase, some physical phenomena of the change over a pressurizer function, i.e., the pressurizer being full before the POSRV 1st opening and then its function being taken by the RV, and the termination of normal natural circulation flow were identified. Finally, a core heatup occurred at a low core water level, although under a significant amount of PZR inventory, whose drainage seemed to be hindered owing to the pressurizer function by the RV. The transient of SBO-01 is well reproduced in the calculation using the MARS code.

  10. Response of centrifugal blowers to simulated tornado transients, July-September 1981

    International Nuclear Information System (INIS)

    Idar, E.S.; Gregory, W.S.; Martin, R.A.; Littleton, P.E.

    1982-03-01

    During this quarter, quasi-steady and dynamic testing of the 24-in. centrifugal blower was completed using the blowdown facility located at New Mexico State University. The data were obtained using a new digital data-acquisition system. Software was developed at the Los Alamos National Laboratory to reduce the dynamic test data and create computer-generated movies showing the dynamic performance of the blower under simulated tornado transient pressure conditions relative to its quasi-steady-state performance. Currently, quadrant-four (outrunning flow) data have been reduced for the most severe and a less severe tornado pressure transient. The results indicate that both the quasi-steady and dynamic blower performance are very similar. Some hysteresis in the dynamic performance occurs because of rotational inertia effects in the blower rotor and drive system. Currently quadrant-two (backflow) data are being transferred to the LTSS computer system at Los Alamos and will be reduced shortly

  11. Environmental assessment of the projected uses for geopressured waters

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, J.S.; Manning, J.A.; Meriwether, J.

    1977-11-16

    An assessment of possible environmental effects of the use of geopressured water of the Texas and Louisiana Gulf Coast has been made. The uses considered include generation of electric power, production of low pressure steam for process heat and the direct use of the hot water for space heating. Based upon the projected uses, the direct and indirect emissions are estimated and the impact of these emissions upon the environment are discussed. The possible impacts of the production of large volumes of geopressured fluids are also considered in terms of possibility of subsidence and earthquakes. A summary of available analyses of Gulf Coast deep waters is listed as a guide for estimating expected emissions. Primary environmental problems are identified as waste brine disposal, accidental releases of brines, and subsidence. Minor problems such as cooling tower blowdown streams, noncondensable gas emissions, wind drift from exhaust plumes, noise levels, and construction activities are considered.

  12. The impact of plasma induced flow on the boundary layer in a narrow channel

    Directory of Open Access Journals (Sweden)

    Procházka P.

    2015-01-01

    Full Text Available The induced flow generated by dielectric barrier discharge (DBD actuator working in steady and unsteady regime will be used to modify properties of naturally developed boundary layer (BL in short and long rectangular perspex channel which is connected to the blow-down wind tunnel. The actuator is placed in spanwise configuration and the inlet velocities will range between 5 and 20 m•s-1. Previously, mean flow field and statistical quantities were subjugated to investigation. In this paper, there will be presented dynamical features of the BL. Oscillation pattern decomposition (OPD of influenced flow field and frequency analysis will be presented. These results should be taken into account regarding to use in the flow around a bluff body.

  13. Safety evaluation report on Westinghouse Electric Company ECCS evaluation model for plants equipped with upper head injection

    International Nuclear Information System (INIS)

    Lauben, G.N.; Wagner, N.H.; Israel, S.L.; McPherson, G.D.; Hodges, M.W.

    1978-04-01

    For plants which include an ice condenser containment concept, Westinghouse has planned an additional safety system known as the upper head injection (UHI) system to augment the emergency core cooling system. This system is comprised of additional accumulator tanks and piping arranged to supply cooling water to the top of the core during the blowdown period following a postulated large-break loss-of-coolant accident (LOCA). The objective of UHI is to add to the core cooling provided by the conventional emergency core cooling system (ECCS) and so permit operation at linear heat rates comparable to those permitted in plants utilizing the dry containment concept. In this way, plants which include the UHI system would have greater operating flexibility while still meeting the acceptance criteria as defined in paragraph 50.46 of 10 CFR Part 50. This review is concerned with those changes to the Westinghouse ECCS evaluation model that have been proposed for the UHI-LOCA model

  14. TRAC analysis of steam-generator overfill transients for TMI-1

    International Nuclear Information System (INIS)

    Bassett, B.

    1983-01-01

    A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

  15. Water reclamation and reuse

    International Nuclear Information System (INIS)

    Hrudey, S.E.

    1982-01-01

    A literature review of wastewater treatment for recycle is presented. Wastewater and activated sludge from the processing of petroleum, shale oil, and from coal conversion and lignite liquefaction have been successfully treated for use as boiler feedwater, cooling water makeup, and steam generation. Acid mine drainage has been treated with lime for use in revegetation of spoil areas. Use of tailings decant water for use in a mill concentrator was reported. Ionizing radiation was effective in disinfecting wastewater makeup to power plant cooling systems. The zero discharge concept was demonstrated in several power plants. Reverse osmosis is reported to be the most economical technology for treatment of cooling tower blowdown. It has the capability of 44% recovery of boric acid and 55% recovery of water from nuclear power plant radioactive wastewater. Included are 402 references

  16. Implementation of the thermal-hydraulic transient analysis code RELAP4/MOD5 and MOD6 on the FACOM 230/75 computer system

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Ishigai, Takahiro; Kumakura, Toshimasa; Naraoka, Ken-itsu

    1979-03-01

    Development efforts have continued on the extensively used LOCA analysis code RELAP-4, as seen in its history; that is, from the prototype version MOD2 to the latest one MOD6 which is capable of one-through calculations from blowdown to reflood phase of PWR-LOCA. Many improvements and refinements of the models have enlarged the scopes and extents of phenomena to treat. Correspondingly the size of program has increased version to version, and special programming techniques have continuously been introduced to manage the program within limited capacity of core memory. For example, the Dynamic Storage Allocation of MOD5 and the PRELOAD Preprocessor newly incorporated in MOD6 are those designed for the CDC computer with relatively small core size. Described are these programming techniques in detail and experiences on implementation of the codes on FACOM 230/75, together with some results of confirmatory calculations. (author)

  17. Comparison of NUPIPE-II and SAP IV predicted and experimentally determined dynamic structural responses for German Standard Problem 4a. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Dooley, W.T.; Mosby, W.R.

    1983-01-01

    This paper presents comparisons between two computer code predictions and experimental measurements of the structural response of a pipe line/check valve system subjected to loading from a loss of feedwater transient. The piping system that was modeled and instrumented for measurement was the focus of German Standard Problem 4a; part of the Heissdampfreaktor Safety Program being conducted in the Federal Republic of Germany. The availability of their experimental data offered EG and G a unique opportunity to evaluate two structural codes' predictions, to compare them with each other, and to compare their predictions with the actual measured values of acceleration and displacement. A thermal-hydrualic code, SOLA-LOOP, computed the hydraulic behavior of the system. The hydraulic forcing functions were calculated and placed into the structural codes, NUPIPE-II and SAP IV. It was concluded that both computer programs provided comparable, realistic predictions of the piping system dynamic response to a blowdown load.

  18. Structural dynamics and fracture mechanics calculations of the behaviour of a DN 425 test piping system subjected to transient loading by water hammer

    International Nuclear Information System (INIS)

    Kussmaul, K.; Kobes, E.; Diem, H.; Schrammel, D.; Brosi, S.

    1994-01-01

    Within the scope of the German HDR safety programme, several tests were carried out to investigate transient pipe loading initiated by a simulated double-ended guillotine break event, and subsequent closure of a feedwater check valve (water hammer, blow-down). Numerical analyses by means of finite element programmes were performed in parallel to the experiments. Using water hammer tests of a DN 425 piping system with predamaged components, the procedure of such analyses will be demonstrated. The results are presented, beginning with structural dynamic calculations of the undamaged piping; followed by coupling of structural dynamics and fracture mechanics computations with simple flaw elements (line spring); and finishing with costly three-dimensional fracture mechanics analyses. A good description of the real piping behaviour can be made by the numerical methods, even in the case of high plastification processes. ((orig.))

  19. Experimental study on utilization of air-borne jet sound in coolant leak detector

    International Nuclear Information System (INIS)

    Hayamizu, Y.; Kitahara, T.; Hayashi, T.; Nishimura, M.

    1975-10-01

    Studies have been undertaken to develop a new coolant leak detection method by the use of a microphone to pick up jet sound generated when pressurized high temperature water is discharged from a pressure boundary into the atmosphere. Leakage was simulated in three shapes, such as two machine-made circular holes and longitudinal and transverse slits in an inlet tube of a blowdown test facility. The measured power level of the jet sound was in agreement with theoretical values calculated from Lighthill's equation. In the study of utilization, this new method has been confirmed as applicable, and to be calculated theoretically for design on 'signal to noise ratio' evaluation. Detection of a small coolant leakage of 1 kg/sec is possible in a recirculation pump room which has large background noise from the pump if a suitable isolation wall, such as hot boxes, is installed between the monitored pipes and the pump. (auth.)

  20. An experimental investigation of the thermal mixing in a water pool using a simplified I-sparger

    International Nuclear Information System (INIS)

    Kim, Y. S.; Jun, H. G.; Youn, Y. J.; Park, C. K.; Song, C. H.

    2004-01-01

    The SDVS (Safety Depressurization and Vent System) in the APR1400 is designed to cope with some DBEs (Design Bases Events) and beyond-DBEs related to overpressurization of the RCS (Reactor Coolant System). When the POSRV (Power Operated Safety Relief Valve) is actuated, steam from the pressurizer is discharged to the IRWST(In-containment Refueling Water Storage Tank) through I-spargers. When injected steam is condensed in the pool, it induces water motions and temperature variations in the pool, which effects on the steam jet condensation, vice versa. The B and C(Blowdown and Condensation) loop is a test facility for the thermal mixing through a steam sparger in a water pool. Thermal mixing tests provide basic understanding of the physics and some insights related to efficient pool mixing, dynamic load, and the IRWST design improvement etc

  1. Suppression pool in nuclear power plant

    International Nuclear Information System (INIS)

    Hayakumo, Sunao.

    1981-01-01

    Purpose: To improve the efficiency of vapour condensation for the sake of steam-load depression at the time of blowdown, and to prevent the quake of supression pool water at the time of earthquake. Constitution: Double branching plates having a function of a branching vapor stream in two directions when blowing down the vapor and operating the vent safety valve are provided on the central line of the vent tube disposed radially from the center of a reactor housing in a dry well. Further, a vent safety valve exhaust device is provided between the branching plates. When the vapor discharged from the space in the dry well is discharged through the vent tube and the vent safety valve exhaust device into a suppression pool, the stream line is roughly split by the branching plates, and the flows from the adjacent branching plates and the exhaust device collide with one another, thereby improving the condensing action. (Sekiya, K.)

  2. Connection of safety valves of nuclear power plant steam generator secondary circuit

    International Nuclear Information System (INIS)

    Houska, J.; Zach, J.; Karlach, J.

    1992-01-01

    A common collector is formed, connecting all the branch routes fitted with check flap valves and controllable valves and attached to the blowdown routes of each steam generator. The common collector is attached to the inlet of the safety valve system. The outlet of this safety valve system is connected via a connecting branch to the outlet of the pressurizer safety valve system. The outlets of the pressurizer safety valves open to the pressurizer relief tank. The whole is accommodated in the containment, which ensures that no radioactivity leaks into air during a steam generator accidents. The outlets of the separately controlled safety valves at the attachment branches as well as of the safety valve system of the common collector, reach below the water level of the tank; water is thus cooled at the saturation limit whereby formation of steam by expansion inside the containment is prevented. (M.D.). 1 fig

  3. Dawn Spacecraft Reaction Control System Flight Experience

    Science.gov (United States)

    Mizukami, Masashi; Nakazono, Barry

    2014-01-01

    The NASA Dawn spacecraft mission is studying conditions and processes of the solar system's earliest epoch by investigating two protoplanets remaining intact since their formations, Ceres and Vesta. Launch was in 2007. Ion propulsion is used to fly to and enter orbit around Vesta, depart Vesta and fly to Ceres, and enter orbit around Ceres. A conventional blowdown hydrazine reaction control system (RCS) is used to provide external torques for attitude control. Reaction wheel assemblies were intended to provide attitude control in most cases. However, the spacecraft experienced one, then two apparent failures of reaction wheels. Also, similar thrusters experienced degradation in a long life application on another spacecraft. Those factors led to RCS being operated in ways completely different than anticipated prior to launch. Numerous mitigations and developments needed to be implemented. The Vesta mission was fully successful. Even with the compromises necessary due to those anomalies, the Ceres mission is also projected to be feasible.

  4. Development of general-purpose software to analyze the static thermal characteristic of nuclear power plant

    International Nuclear Information System (INIS)

    Nakao, Yoshinobu; Koda, Eiichi; Takahashi, Toru

    2009-01-01

    We have developed the general-purpose software by which static thermal characteristic of the power generation system is analyzed easily. This software has the notable features as follows. It has the new algorithm to solve non-linear simultaneous equations to analyze the static thermal characteristics such as heat and mass balance, efficiencies, etc. of various power generation systems. It has the flexibility for setting calculation conditions. It is able to be executed on the personal computer easily and quickly. We ensured that it is able to construct heat and mass balance diagrams of main steam system of nuclear power plant and calculate the power output and efficiencies of the system. Furthermore, we evaluated various heat recovery measures of steam generator blowdown water and found that this software could be a useful operation aid for planning effective changes in support of power stretch. (author)

  5. Forced circulation steam generators for SAGD applications

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevan, M [Engineered Boiler Systems (United States)

    2011-07-01

    Steam for steam assisted gravity drainage (SAGD) is traditionally supplied by once-through steam generators (OTSG) or drum type boilers. Research by Cleaver Brooks aims to combine the quality features of OTSGs and drum boilers and to address issues encountered with both these types of machinery. The forced-circulation oil sands steam generator (FC-OSSG) resulted from the design of a flow management system which allowed water quality upsets to be handled successfully. Integration of the boiler and the burner design through computational fluid dynamics (CFD) led to a decrease in the environmental footprint of the steam generator. Circuit maintenance was also significantly facilitated by elaborating new piggable circuits. Moreover, the design of a water cooled furnace enabled the generator to work at constant heat. The resulting FC-OSSG produces cleaner steam than either OTSGs or drum boilers, and has a higher capacity than OTSGs. Additionally, turndown, blowdown and feedwater quality are similar to that obtained using drum boilers.

  6. OECD-LOFT large break LOCA experiments: phenomenology and computer code analyses

    International Nuclear Information System (INIS)

    Brittain, I.; Aksan, S.N.

    1990-08-01

    Large break LOCA data from LOFT are a very important part of the world database. This paper describes the two double-ended cold leg break tests LP-02-6 and LP-LB-1 carried out within the OECD-LOFT Programme. Tests in LOFT were the first to show the importance of both bottom-up and top-down quenching during blowdown in removing stored energy from the fuel. These phenomena are discussed in detail, together with the related topics of the thermal performance of nuclear fuel and its simulation by electric fuel rod simulators, and the accuracy of cladding external thermocouples. The LOFT data are particularly important in the validation of integral thermal-hydraulics codes such as TRAC and RELAP5. Several OECD partner countries contributed analyses of the large break tests. Results of these analyses are summarised and some conclusions drawn. 32 figs., 3 tabs., 45 refs

  7. Verification of the HDR-test V44 using the computer program RALOC-MOD1/83

    International Nuclear Information System (INIS)

    Jahn, H.; Pham, T. v.; Weber, G.; Pham, B.T.

    1985-01-01

    RALOC-MOD1/83 was extended by a drainage and sump level modul and several component models to serve as a containment systems code for various LWR types. One such application is to simulate the blowdown in a full pressure containment which is important for the short and long term hydrogen distribution. The post test calculation of the containment standard problem experiment HDR-V44 shows a good agreement, to the test data. The code may be used for short and long term predictions, but it was learned that double containments need the representation of the gap between the inner and outer shell into several zones to achieve a good long-term temperature prediction. The present work completes the development, verification and documentation of RALOC-MOD1. (orig.) [de

  8. Experience with dispersant application: long-path recirculation cleanup trial at Byron Unit 1 during spring 2011 and online addition update

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Marks, C.; Kreider, M.; Morey, D.; Duncanson, I.; Bates, J.; Sawochka, S.

    2012-01-01

    The first nuclear application of PAA dispersant to improve corrosion product removal during LPR (Long-path recirculation) cleanup occurred at Byron Unit 1 in spring 2011. The main conclusions and lessons learned are as follows: -) there were no significant problems with application of PAA during LPR with an initial PAA concentration of about 650 ppb; -) a reasonable estimate of the additional iron mass removed due to the presence of PAA is 5-9 kg. The qualification work, application details and an assessment of the results are the first focus of this paper. The second part of this paper summarizes the online experience to date at the Exelon and STP (South Texas Project) plants on the effects of dispersant on -) blowdown iron removal efficiency, -) steam generator heat transfer efficiency and -) ion exchange resin performance

  9. Study of Head-loss effect for ECCS Strainer Design Variables with Debris Condition

    International Nuclear Information System (INIS)

    Lee, Sung Myung; Lee, Jong Wook; Kim, Won Seok; Lee, Sang Il; Kim, Chang Hyun; Kim, Sang Yeol

    2011-01-01

    LOCA(Loss Of Coolant Accident) due to station damage such as pipe break in NPP generates various debris fragments. Debris moves into recirculation sump at the bottom of NPP with accompanying blow-down, wash-down, pool-fill and recirculation. If strainers at the sump have not enough performance of filtering, it will generate higher pressure drop inside perforated plate of strainer and affect safety issue. Especially as strainers installed do not satisfy design requirements and performance against NPP LOCA, it is necessary to install new strainer with low head-loss, higher safety requirements compared to the existing strainer. In this study, considering the different situations in each NPP station and design parameters of strainers, we study the optimized design of new strainer

  10. Apparatus for monitoring two-phase flow

    International Nuclear Information System (INIS)

    Sheppard, J.D.; Tong, L.S.

    1977-01-01

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods. 3 claims, 9 figures

  11. LOFT fuel module structural response during loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Selcho, H.S.

    1979-01-01

    The structural response of the reactor fuel modules installed in the Loss-of-Fluid Test (LOFT) facility have been analyzed for subcooled blowdown loading conditions associated with loss-of-coolant experiments (LOCE). Three independent analyses using the WHAM, SHOCK, and SAP computer codes have been interfaced to calculate the transient mechanical behavior of the LOFT fuel. Test data from two LOCEs indicate the analysis method is conservative. Structural integrity of the fuel modules has been assessed by monitoring guide tube temperatures and control rod drop times during the LOCEs. The analysis and experimental test data indicate the fuel module structural integrity will be maintained for the duration of the LOFT experimental program

  12. INCAS TRISONIC WIND TUNNEL

    Directory of Open Access Journals (Sweden)

    Florin MUNTEANU

    2009-09-01

    Full Text Available The 1.2 m x 1.2 m Trisonic Blowdown Wind Tunnel is the largest of the experimental facilities at the National Institute for Aerospace Research - I.N.C.A.S. "Elie Carafoli", Bucharest, Romania. The tunnel has been designed by the Canadian company DSMA (now AIOLOS and since its commissioning in 1978 has performed high speed aerodynamic tests for more than 120 projects of aircraft, missiles and other objects among which the twin jet fighter IAR-93, the jet trainer IAR-99, the MIG-21 Lancer, the Polish jet fighter YRYDA and others. In the last years the wind tunnel has been used mostly for experimental research in European projects such as UFAST. The high flow quality parameters and the wide range of testing capabilities ensure the competitivity of the tunnel at an international level.

  13. Effect of Axisymmetric Aft Wall Angle Cavity in Supersonic Flow Field

    Science.gov (United States)

    Jeyakumar, S.; Assis, Shan M.; Jayaraman, K.

    2018-03-01

    Cavity plays a significant role in scramjet combustors to enhance mixing and flame holding of supersonic streams. In this study, the characteristics of axisymmetric cavity with varying aft wall angles in a non-reacting supersonic flow field are experimentally investigated. The experiments are conducted in a blow-down type supersonic flow facility. The facility consists of a supersonic nozzle followed by a circular cross sectional duct. The axisymmetric cavity is incorporated inside the duct. Cavity aft wall is inclined with two consecutive angles. The performance of the aft wall cavities are compared with rectangular cavity. Decreasing aft wall angle reduces the cavity drag due to the stable flow field which is vital for flame holding in supersonic combustor. Uniform mixing and gradual decrease in stagnation pressure loss can be achieved by decreasing the cavity aft wall angle.

  14. Investigation of some green compounds as corrosion and scale inhibitors for cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Quraishi, M.A.; Farooqi, I.H.; Saini, P.A. (Aligarh Muslim Univ. (India))

    1999-05-01

    The performance of an open-recirculating cooling system, an important component in most industries, is affected by corrosion and scale formation. Numerous additives have been used in the past for the control of corrosion and scale formation. Effects of the naturally occurring compounds azadirachta indica (leaves), punica granatum (shell), and momordica charantia (fruits), on corrosion of mild steel in 3% sodium chloride (NaCl) were assessed using weight loss, electrochemical polarization, and impedance techniques. Extracts of the compounds exhibited excellent inhibition efficiencies comparable to that of hydroxyethylidine diphosphonic acid (HEDP), the most preferred cooling water inhibitor. The compounds were found effective under static and flowing conditions. Extracts were quite effective in retarding formation of scales, and the maximum antiscaling efficiency was exhibited by the extract of azadirachta indica (98%). The blowdown of the cooling system possessed color and chemical oxygen demand (COD). Concentrations of these parameters were reduced by an adsorption process using activated carbon as an adsorbent.

  15. Thermohydraulic behaviour of the hot channel in a PWR type reactor under loss-of-coolant accident conditions (LOCA)

    International Nuclear Information System (INIS)

    Costa, J.R.

    1978-12-01

    An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt

  16. Portsmouth gaseous diffusion plant environmental monitoring report for calendar year 1975

    International Nuclear Information System (INIS)

    Martin, W.E.; Netzer, W.D.

    1976-01-01

    At the Portsmouth Gaseous Diffusion Plant the ambient atmosphere and all effluent streams are sampled and analyzed regularly for conformance to applicable environmental standards. Although neither the State of Ohio nor the federal government has established standards for fluorides in the ambient atmosphere or in vegetation, these parameters also are monitored because fluoride compounds are used extensively in the gaseous diffusion process. Radioactivity is measured in air, water, food, soil, and sediments; and radiation doses are calculated for the public. All public radiation doses are well within federal standards. Non-radioactive effluent parameters either comply with federal standards, or there are projects planned to allow compliance. A disposal facility to remove chromium from recirculating cooling water blowdown will begin operation in June 1976. Also, pH adjustment facilities for liquid effluents and electrostatic precipitators for a coal-fired steam plant are planned for the near future

  17. Analytical modelling of hydrogen transport in reactor containments

    International Nuclear Information System (INIS)

    Manno, V.P.

    1983-09-01

    A versatile computational model of hydrogen transport in nuclear plant containment buildings is developed. The background and significance of hydrogen-related nuclear safety issues are discussed. A computer program is constructed that embodies the analytical models. The thermofluid dynamic formulation spans a wide applicability range from rapid two-phase blowdown transients to slow incompressible hydrogen injection. Detailed ancillary models of molecular and turbulent diffusion, mixture transport properties, multi-phase multicomponent thermodynamics and heat sink modelling are addressed. The numerical solution of the continuum equations emphasizes both accuracy and efficiency in the employment of relatively coarse discretization and long time steps. Reducing undesirable numerical diffusion is addressed. Problem geometry options include lumped parameter zones, one dimensional meshs, two dimensional Cartesian or axisymmetric coordinate systems and three dimensional Cartesian or cylindrical regions. An efficient lumped nodal model is included for simulation of events in which spatial resolution is not significant. Several validation calculations are reported

  18. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

  19. Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

    This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

  20. Fluid history computation methods for reactor safeguards problems using MNODE computer program

    International Nuclear Information System (INIS)

    Huang, Y.S.; Savery, C.W.

    1976-10-01

    A method for predicting the pressure-temperature histories of air, water liquid, and vapor flowing in a zoned containment as a result of high energy pipe rupture is described. The computer code, MNODE, has been developed for 12 connected control volumes and 24 inertia flow paths. Predictions by the code are compared with the results of an analytical gas dynamic problem, semiscale blowdown experiments, full scale MARVIKEN test results, Battelle-Frankfurt model PWR containment test data. The MNODE solutions to NRC/AEC subcompartment benchmark problems are also compared with results predicted by other computer codes such as RELAP-3, FLASH-2, CONTEMPT-PS. The analytical consideration is consistent with Section 6.2.1.2 of the Standard Format (Rev. 2) issued by U.S. Nuclear Regulatory Commission in September 1975

  1. CANDU fuel behaviour under LOCA conditions

    International Nuclear Information System (INIS)

    Kohn, E.

    1989-07-01

    This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

  2. Fast reactor primary cover gas system proposals for CDFR

    International Nuclear Information System (INIS)

    Harrison, L.M.T.

    1987-01-01

    A primary sodium gas cover has been designed for CDFR, it comprises plant to maintain and control; cover gas pressure for all reactor operating at fault conditions, cover gas purity by both blowdown and by a special clean-up facility and the clean argon supply for the failed fuel detection system and the primary pump seal purge. The design philosophy is to devise a cover gas system that can be specified for any LMFBR where only features like vessel and pipework size need to be altered to suit different design and operating conditions. The choice of full power and shutdown operating pressures is derived and the method chosen to control these values is described. A part active/part passive system is proposed for this duty, a surge volume of 250 m 3 gives passive control between full power and hot shutdown. Pressure control operation criteria is presented for various reactor operating conditions. A design for a sodium aerosol filter, based on that used on PFR is presented, it is specifically designed so that it can be fitted with an etched disc type particulate filter and maintenance is minimised. Two methods that maintain cover gas purity are described. The first, used during normal reactor operation with a small impurities ingress, utilises the continuous blowdown associated with the inevitable clean argon purge through the various reactor component seals. The second method physically removes the impurities xenon and krypton from the cover gas by their adsorption, at cryogenic temperature, onto a bed of activated carbon. The equipment required for these two duties and their mode of operation is described with the aid of a system flow diagram. The primary pump seals requires a gas purge to suppress aerosol migration. A system where the argon used for this task is recirculated and partially purified is described. (author)

  3. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  4. One-dimensional transient unequal velocity two-phase flow by the method of characteristics

    International Nuclear Information System (INIS)

    Rasouli, F.

    1981-01-01

    An understanding of two-phase flow is important when one is analyzing the accidental loss of coolant or when analyzing industrial processes. If a pipe in the steam generator of a nuclear reactor breaks, the flow will remain critical (or choked) for almost the entire blowdown. For this reason the knowledge of the two-phase maximum (critical) flow rate is important. A six-equation model--consisting of two continuity equations, two energy equations, a mixture momentum equation, and a constitutive relative velocity equation--is solved numerically by the method of characteristics for one-dimensional, transient, two-phase flow systems. The analysis is also extended to the special case of transient critical flow. The six-equation model is used to study the flow of a nonequilibrium sodium-argon system in a horizontal tube in which the nonequilibrium sodium-argon system in a horizontal tube in which the critical flow condition is at the entrance. A four-equation model is used to study the pressure-pulse propagation rate in an isothermal air-water system, and the results that are found are compared with the experimental data. Proper initial and boundary conditions are obtained for the blowdown problem. The energy and mass exchange relations are evaluated by comparing the model predictions with results of void-fraction and heat-transfer experiments. A simplified two-equation model is obtained for the special case of two incompressible phases. This model is used in the preliminary analysis of batch sedimentation. It is also used to predict the shock formation in the gas-solid fluidized bed

  5. Hydrogen distribution analysis for CANDU 6 containment using the GOTHIC containment analysis code

    International Nuclear Information System (INIS)

    Nguyen, T.H.; Collins, W.M.

    1995-01-01

    Hydrogen may be generated in the reactor core by the zircaloy-steam reaction for a postulated loss of coolant accident (LOCA) scenario with loss of emergency core cooling (ECC). It is important to predict hydrogen distribution within containment in order to determine if flammable mixtures exist. This information is required to determine the best locations in containment for the placement of mitigation devices such as igniters and recombiners. For large break loss coolant accidents, hydrogen is released after the break flow has subsided. Following this period of high discharge the flow in the containment building undergoes transition from forced flow to a buoyancy driven flow (particularly when local air coolers (LACS) are not credited). One-dimensional computer codes (lumped parameter) are applicable during the initial period when a high degree of mixing occurs due to the forced flow generated by the break. However, during the post-blowdown phase the assumption of homogeneity becomes less accurate, and it is necessary to employ three-dimensional codes to capture local effects. This is particularly important for purely buoyant flows which may exhibit stratification effects. In the present analysis a three-dimensional model of CANDU 6 containment was constructed with the GOTHIC computer code using a relatively coarse mesh adequate enough to capture the salient features of the flow during the blowdown and hydrogen release periods. A 3D grid representation was employed for that portion of containment in which the primary flow (LOCA and post-LOCA) was deemed to occur. The remainder of containment was represented by lumped nodes. The results of the analysis indicate that flammable concentrations exist for several minutes in the vicinity of the break and in the steam generator enclosure. This is due to the fact that the hydrogen released from the break is primarily directed upwards into the steam generator enclosure due to buoyancy effects. Once hydrogen production ends

  6. RELAP5/MOD3 assessment using the Semiscale 50% Feed Line Break test S-FS-11

    Energy Technology Data Exchange (ETDEWEB)

    Lee, E.J.; Chung, B.D.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1993-06-01

    The RELAP5/MOD3 5m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase, (b) stabilization phase, and (c) refill phase. The first objective was to assess the code applicability to 50% FLB situation, the second was to evaluate the FSAR conservatisms regarding SG heat transfer degradation, steam line check valve failure, break flow state, and peak primary system pressure, and the third was to validate the EOP effectiveness. The code was able to simulate the major T/H parameters except for the two-phase break flow and the secondary convective heat transfer rate. The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not either conservative, the measured break flow experienced all types of flow conditions, the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operation performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator.

  7. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  8. Development of Realistic Safety Analysis Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Rho, G. H.

    2010-04-01

    The following 3 research items have been studied to develop and establish the realistic safety analysis and the associated technologies for a CANDU reactor. At the first, WIMS-CANDU which is physics cell code for a CANDU has been improved and validated against the physics criticality experiment data transferred through the international cooperation programs. Also an improved physics model to take into account the pressure tube creep was developed and utilized to assess the effects of the pressure tube creep of 0%, 2.5% and 5% diametral increase of pressure tube on core physics parameters. Secondly, the interfacing module between physics and thermal-hydraulics codes has been developed to provide the enhancement of reliability and convenience of the calculation results of the physics parameters such as power coefficient which was calculated by independent code systems. Finally, the important parameters related to the complex heat transfer mechanisms in the crept pressure tubes were identified to find how to improve the existing fuel channel models. One of the important parameters such as the oxidation model of Zr-steam reaction was identified, implemented and verified with the experimental data of the high pressure and temperature fuel channel and its model was utilized for CFD analysis of the crept pressure tube effect on the reactor safety. The results were also provided to validate the CATNENA models of the crept pressure tube and the effects of the pressure tube creep on the blowdown and post-blowdown phase during LOCA was assessed. The results of this study can be used to assess the uncertainty analysis of coolant void reactivity and the effects of the creep deformed pressure tubes on physics/TH/safety issues. Also, those results will be used to improve the current design and operational safety analysis codes, and to technically support the related issues to resolve their problems

  9. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  10. State waste discharge permit application for the 200 Area Effluent Treatment Facility and the State-Approved Land Disposal Site

    International Nuclear Information System (INIS)

    1993-08-01

    Application is being made for a permit pursuant to Chapter 173--216 of the Washington Administrative Code (WAC), to discharge treated waste water and cooling tower blowdown from the 200 Area Effluent Treatment Facility (ETF) to land at the State-Approved Land Disposal Site (SALDS). The ETF is located in the 200 East Area and the SALDS is located north of the 200 West Area. The ETF is an industrial waste water treatment plant that will initially receive waste water from the following two sources, both located in the 200 Area on the Hanford Site: (1) the Liquid Effluent Retention Facility (LERF) and (2) the 242-A Evaporator. The waste water discharged from these two facilities is process condensate (PC), a by-product of the concentration of waste from DSTs that is performed in the 242-A Evaporator. Because the ETF is designed as a flexible treatment system, other aqueous waste streams generated at the Hanford Site may be considered for treatment at the ETF. The origin of the waste currently contained in the DSTs is explained in Section 2.0. An overview of the concentration of these waste in the 242-A Evaporator is provided in Section 3.0. Section 4.0 describes the LERF, a storage facility for process condensate. Attachment A responds to Section B of the permit application and provides an overview of the processes that generated the wastes, storage of the wastes in double-shell tanks (DST), preliminary treatment in the 242-A Evaporator, and storage at the LERF. Attachment B addresses waste water treatment at the ETF (under construction) and the addition of cooling tower blowdown to the treated waste water prior to disposal at SALDS. Attachment C describes treated waste water disposal at the proposed SALDS

  11. The Savannah River Plant Consolidated Incineration Facility

    International Nuclear Information System (INIS)

    Weber, D.A.

    1987-01-01

    A full scale incinerator is proposed for construction at the Savannah River Plant (SRP) beginning in August 1989 for detoxifiction and volume reduction of liquid and solid low-level radioactive, mixed and RCRA hazardous waste. Wastes to be burned include drummed liquids, sludges and solids, liquid process wastes, and low-level boxed job control waste. The facility will consist of a rotary kiln primary combustion chamber followed by a tangentially fired cylindrical secondary combustion chamber (SCC) and be designed to process up to 12 tons per day of solid and liquid waste. Solid waste packaged in combustible containers will be fed to the rotary kiln incinerator using a ram feed system and liquid wastes will be introduced to the rotary kiln through a burner nozzle. Liquid waste will also be fed through a high intensity vortex burner in the SCC. Combustion gases will exit the SCC and be cooled to saturation in a spray quench. Particulate and acid gas are removed in a free jet scrubber. The off-gas will then pass through a cyclone separator, mist eliminator, reheater high efficiency particulate air (HEPA) filtration and induced draft blowers before release to the atmosphere. Incinerator ash and scrubber blowdown will be immobilized in a cement matrix and disposed of in an onsite RCRA permitted facility. The Consolidated Incineration Facility (CIF) will provide detoxification and volume reduction for up to 560,000 CUFT/yr of solid waste and up to 35,700 CUFT/yr of liquid waste. Up to 50,500 CUFT/yr of cement stabilized ash and blowdown will beproduced for an average overall volume reduction fator of 22:1. 3 figs., 2 tabs

  12. Accident tolerant high-pressure helium injection system concept for light water reactors

    International Nuclear Information System (INIS)

    Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

    2016-01-01

    Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

  13. RELAP5/MOD3 assessment using the Semiscale 50% Feed Line Break test S-FS-11

    International Nuclear Information System (INIS)

    Lee, E.J.; Chung, B.D.; Kim, H.J.

    1993-06-01

    The RELAP5/MOD3 5m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase, (b) stabilization phase, and (c) refill phase. The first objective was to assess the code applicability to 50% FLB situation, the second was to evaluate the FSAR conservatisms regarding SG heat transfer degradation, steam line check valve failure, break flow state, and peak primary system pressure, and the third was to validate the EOP effectiveness. The code was able to simulate the major T/H parameters except for the two-phase break flow and the secondary convective heat transfer rate. The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not either conservative, the measured break flow experienced all types of flow conditions, the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operation performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator

  14. Calculation of relative tube/tube support plate displacements in steam generators under accident condition loads using non-linear dynamic analysis methodologies

    International Nuclear Information System (INIS)

    Smith, R.E.; Waisman, R.; Hu, M.H.; Frick, T.M.

    1995-01-01

    A non-linear analysis has been performed to determine relative motions between tubes and tube support plates (TSP) during a steam line break (SLB) event for steam generators. The SLB event results in blowdown of steam and water out of the steam generator. The fluid blowdown generates pressure drops across the TSPS, resulting in out-of-plane motion. The SLB induced pressure loads are calculated with a computer program that uses a drift-flux modeling of the two-phase flow. In order to determine the relative tube/TSP motions, a nonlinear dynamic time-history analysis is performed using a structural model that considers all of the significant component members relative to the tube support system. The dynamic response of the structure to the pressure loads is calculated using a special purpose computer program. This program links the various substructures at common degrees of freedom into a combined mass and stiffness matrix. The program accounts for structural non-linearities, including potential tube and TSP interaction at any given tube position. The program also accounts for structural damping as part of the dynamic response. Incorporating all of the above effects, the equations of motion are solved to give TSP displacements at the reduced set of DOF. Using the displacement results from the dynamic analysis, plate stresses are then calculated using the detailed component models. Displacements form the dynamic analysis are imposed as boundary conditions at the DOF locations, and the finite element program then solves for the overall distorted geometry. Calculations are also performed to assure that assumptions regarding elastic response of the various structural members and support points are valid

  15. Generalization of experimental data on amplitude and frequency of oscillations induced by steam injection into a subcooled pool

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva, Walter; Li, Hua [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Puustinen, Markku [Nuclear Engineering, LUT School of Energy Systems, Lappeenranta University of Technology (LUT), FIN-53851 Lappeenranta (Finland); Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2015-12-15

    Highlights: • Available data on steam injection into subcooled pool is generalized. • Scaling approach is proposed on amplitude and frequency of chugging oscillations. • The scaled amplitude has a maximum at Froude number Fr ≈ 2.8. • The scaled frequency has a minimum at Fr ≈ 6. • Both amplitude and frequency has a strong dependence on pool bulk temperature. - Abstract: Steam venting and condensation into a subcooled pool of water through a blowdown pipe can undergo a phenomenon called chugging, which is an oscillation of the steam–water interface inside the blowdown pipe. The momentum that is generated by the oscillations is directly proportional to the oscillations’ amplitude and frequency, according to the synthetic jet theory. Higher momentum can enhance pool mixing and positively affect the pool's pressure suppression capacity by reducing thermal stratification. In this paper, we present a generalization of available experimental data on the amplitude and frequency of oscillations during chugging. We use experimental data obtained in different facilities at different scales to suggest a scaling approach for non-dimensional amplitude and frequency of the oscillations. We demonstrate that the Froude number Fr (which relates the inertial forces to gravitational forces) can be used as a scaling criterion in this case. The amplitude has maximum at Fr ≈ 2.8. There is also a strong dependence of the amplitude on temperature; the lower the bulk temperature is the higher the scaled amplitude. A known analytical theory can only capture the decreasing trend in amplitude for Fr > 2.8 and fails to capture the increasing trend and the temperature dependence. Similarly, there is a minimum of the non-dimensional frequency at Fr ≈ 6. A strong dependence on temperature is also observed for Fr > 6; the lower the bulk temperature is the higher the scaled frequency. The known analytical theory is able to capture qualitatively the general trend in

  16. CFD simulation of air discharge tests in the PPOOLEX facility

    International Nuclear Information System (INIS)

    Tanskanen, V.; Puustinen, M.

    2008-07-01

    This report summarizes the CFD simulation results of two air discharge tests of the characterizing test program in 2007 with the scaled down PPOOLEX facility. Air was blown to the dry well compartment and from there through a DN200 blowdown pipe into the condensation pool (wet well). The selected tests were modeled with Fluent CFD code. Test CHAR-09-1 was simulated to 28.92 seconds of real time and test CHAR-09-3 to 17.01 seconds. The VOF model was used as a multiphase model and the standard k ε-model as a turbulence model. Occasional convergence problems, usually at the beginning of bubble formation, required the use of relatively short time stepping. The simulation time costs threatened to become unbearable since weeks or months of wall-clock time with 1-2 processors were needed. Therefore, the simulated time periods were limited from the real duration of the experiments. The results obtained from the CFD simulations are in a relatively good agreement with the experimental results. Simulated pressures correspond well to the measured ones and, in addition, fluctuations due to bubble formations and breakups are also captured. Most of the differences in temperature values and in their behavior seem to depend on the locations of the measurements. In the vicinity of regions occupied by water in the experiments, thermocouples getting wet and drying slowly may have had an effect on the measured temperature values. Generally speaking, most temperatures were simulated satisfyingly and the largest discrepancies could be explained by wetted thermocouples. However, differences in the dry well and blowdown pipe top measurements could not be explained by thermocouples getting wet. Heat losses and dry well / wet well heat transfer due to conduction have neither been estimated in the experiments nor modeled in the simulations. Estimation of heat conduction and heat losses should be carried out in future experiments and they should be modeled in future simulations, too. (au)

  17. DOE mixed waste metals partition in a rotary kiln wet off-gas system

    International Nuclear Information System (INIS)

    Burns, D.B.; Looper, M.G.

    1994-01-01

    In 1996, the Savannah River Site plans to begin operation of the Consolidated Incineration Facility (CIF) to treat solid and liquid RCRA hazardous and mixed wastes. Test burns were conducted using surrogate CIF wastes spiked with hazardous metals and organics. The partition of metals between the kiln bottom ash, scrubber blowdown solution, and stack gas was measured as a function of kiln temperature, waste chloride content, and waste form (liquid or solid). Three waste simulants were used in these tests, a high and low chloride solid waste mix (paper, plastic, latex, PVC), and a liquid waste mix (benzene and chlorobenzene). An aqueous solution containing: antimony, arsenic, barium, cadmium, chromium, lead, mercury, nickel, silver, and thallium was added to the waste to determine metals fate under various combustion conditions. Test results were used to divide the metals into three general groups, volatile, semi-volatile, and nonvolatile metals. Mercury was the only volatile metal. No mercury remained in the kiln bottom ash under any incineration condition. Lead, cadmium, thallium, and silver exhibited semi-volatile behavior. The partition between the kiln ash, blowdown, and stack gas depended on incineration conditions. Chromium, nickel, barium, antimony, and arsenic exhibited nonvolatile behavior, with greater than 90 wt % of the metal remaining in the kiln bottom ash. Incineration temperature had a significant effect on the partition of volatile and semi-volatile metals, and no effect on nonvolatile metal partition. As incineration temperatures were increased, the fraction of metal leaving the kiln increased. Three metals, lead, cadmium, and mercury showed a relationship between chloride concentration in the waste and metals partition. Increasing the concentration of chlorides in the waste or burning liquid waste versus solid waste resulted in a larger fraction of metal exiting the kiln

  18. SPREE: A Successful Seismic Array by a Failed Rift System; Analysis of Seismic Noise in the Seismically Quiet Mid-continent

    Science.gov (United States)

    Wolin, E.; van der Lee, S.; Bollmann, T. A.; Revenaugh, J.; Aleqabi, G. I.; Darbyshire, F. A.; Frederiksen, A. W.; Wiens, D.; Shore, P.

    2014-12-01

    The Superior Province Rifting Earthscope Experiment (SPREE) completed its field recording phase last fall with over 96% data return. While 60% of the stations returned data 100% of the time, only 9 performed below 90% and one station had questionable timing. One station was vandalized, another stolen. One station continued recording after its solar panels were pierced by a bullet, while another two stations survived a wildfire and a blow-down, respectively. The blow-down was an extreme wind event that felled hundreds of thousands of trees around the station. SPREE stations recorded many hundreds of earthquakes. Two regional earthquakes and over 400 teleseismic earthquakes had magnitudes over 5.5 and three, smaller local earthquakes had magnitudes over 2.5. We have calculated power spectral estimates between 0.1-1000 s period for the ~2.5-year lifespan of all 82 SPREE stations. Vertical channels performed quite well across the entire frequency range, falling well below the high noise model of Peterson (1993) and usually within 10-15 dB of nearby Transportable Array stations. SPREE stations' horizontal components suffer from long-period (> 30 s) noise. This noise is quietest at night and becomes up to 30 dB noisier during the day in the summer months. We explore possible causes of this variation, including thermal and atmospheric pressure effects. One possibility is that stations are insulated by snow during the winter, reducing temperature variations within the vault. Spring snowmelt creates instability at many of the SPREE stations, evidenced by frequent recenterings and enhanced long-period noise. For all channels, power in the microseismic band (4-16 s) is strongest in the winter, corresponding to storm season in the Northern Hemisphere, and approximately 20 dB weaker during the summer. The power spectrum and temporal variation of microseismic energy is consistent across the entire SPREE array.

  19. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Robbe, M.F.

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  20. Assessment of the CATHARE 3-D module for LBLOCA simulation

    International Nuclear Information System (INIS)

    Pascal Bazin; Isabelle Dor; Christophe Morel

    2005-01-01

    Full text of publication follows: CATHARE is a best-estimate system code developed by CEA, EDF, FRAMATOME-ANP and IRSN for PWR safety analysis, accident management, definition of plant operating procedure and for research and development. It is also used to quantify conservative margins and for licensing. In the framework of Pressurized Water Reactor (PWR) safety studies, Large Break Loss-Of-Coolant Accident (LB LOCA) prediction is still one of the most important and one of the most difficult problem. The three main phases of a LB LOCA are respectively the blowdown, the refilling and the reflooding phases. During the blowdown, the lower plenum voiding results in water entrainment towards the break by steam flowing from the core. Because of the core radial profile, critical heat flux occur but a nonuniform quenching may take place, which results in a 3-D repartition of the energy stored in the core at the beginning of the reflooding. The refilling phase which starts at the accumulator discharge encounters very complex thermalhydraulic phenomena: very strong condensation which induces instabilities, presence of nitrogen degassing from accumulator water which may have an important effect on the transient, countercurrent flow limitation which occurs in the complex geometry of the annular downcomer. The reflooding phase initial conditions in the core are therefore very non-uniform. The presence of buoyancy driven transverse flows below the quench front assures a very efficient mixing between the fuel assemblies. The quench front progression in the hot assemblies is accelerated by pre-cooling due to water cross-flows just above the quench front. Therefore the clad temperature excursion is moderated in the hot assemblies by an increased water carry-over coming partially from colder assemblies. All these multi-dimensional aspects create a very challenging problem for the CATHARE 3-D module. A good prediction of the lower plenum voiding altogether with the amount of

  1. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2016-04-15

    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

  2. PPOOLEX experiments on thermal stratification and mixing

    Energy Technology Data Exchange (ETDEWEB)

    Puustinen, M.; Laine, J.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2009-08-15

    The results of the thermal stratification experiments in 2008 with the PPOOLEX test facility are presented. PPOOLEX is a closed vessel divided into two compartments, dry well and wet well. Extra temperature measurements for capturing different aspects of the investigated phenomena were added before the experiments. The main purpose of the experiment series was to generate verification data for evaluating the capability of GOTHIC code to predict stratification and mixing phenomena. Altogether six experiments were carried out. Heat-up periods of several thousand seconds by steam injection into the dry well compartment and from there into the wet well water pool were recorded. The initial water bulk temperature was 20 deg. C. Cooling periods of several days were included in three experiments. A large difference between the pool bottom and top layer temperature was measured when small steam flow rates were used. With higher flow rates the mixing effect of steam discharge delayed the start of stratification until the pool bulk temperature exceeded 50 deg. C. The stratification process was also different in these two cases. With a small flow rate stratification was observed only above and just below the blowdown pipe outlet elevation. With a higher flow rate over a 30 deg. C temperature difference between the pool bottom and pipe outlet elevation was measured. Elevations above the pipe outlet indicated almost linear rise until the end of steam discharge. During the cooling periods the measurements of the bottom third of the pool first had an increasing trend although there was no heat input from outside. This was due to thermal diffusion downwards from the higher elevations. Heat-up in the gas space of the wet well was quite strong, first due to compression by pressure build-up and then by heat conduction from the hot dry well compartment via the intermediate floor and test vessel walls and by convection from the upper layers of the hot pool water. The gas space

  3. PPOOLEX experiments on stratification and mixing in the wet well pool

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.; Raesaenen, A.; Tanskanen, V.

    2011-03-01

    This report summarizes the results of the thermal stratification and mixing experiments carried out in 2010 with the scaled down, two compartment PPOOLEX test facility designed and constructed at LUT. Steam was blown into the thermally insulated dry well compartment and from there through the DN200 vertical blowdown pipe to the condensation pool filled with sub-cooled water. The main purpose of the experiment series was to generate verification data for evaluating the capability of GOTHIC and APROS codes to predict stratification and mixing phenomena. Another objective was to test the sound velocity measurement system. Altogether five experiments were carried out. The experiments consisted of a small steam flow rate stratification period and of a mixing period with continuously or stepwise increasing flow rate. The dry well structures were heated up to the level of approximately 90 deg. C before the actual experiments. The initial water bulk temperature was 20 deg. C. When the steam flow rate was low enough (typically ∼100-150 g/s) temperatures below the blowdown pipe outlet remained constant while increasing heat-up occurred towards the pool surface layers indicating strong thermal stratification of the wet well pool water. During the stratification period the highest measured temperature difference between pool bottom and surface was approximately 40 deg. C. During the mixing period total mixing of the pool volume was not achieved in any of the experiments. The bottom layers heated up significantly but never reached the same temperature as the topmost layers. The lowest measured temperature difference between the pool bottom and surface was 7-8 deg. C. According to the test results, it seems that a small void fraction doesn't have an effect on the speed of sound in water and that the acquired sound velocity measurement system cannot be used for the estimation of void fraction in the wet well water pool. However, more tests on this issue have to be executed

  4. PPOOLEX experiments on stratification and mixing in the wet well pool

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M.; Raesaenen, A.; Tanskanen, V. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2011-03-15

    This report summarizes the results of the thermal stratification and mixing experiments carried out in 2010 with the scaled down, two compartment PPOOLEX test facility designed and constructed at LUT. Steam was blown into the thermally insulated dry well compartment and from there through the DN200 vertical blowdown pipe to the condensation pool filled with sub-cooled water. The main purpose of the experiment series was to generate verification data for evaluating the capability of GOTHIC and APROS codes to predict stratification and mixing phenomena. Another objective was to test the sound velocity measurement system. Altogether five experiments were carried out. The experiments consisted of a small steam flow rate stratification period and of a mixing period with continuously or stepwise increasing flow rate. The dry well structures were heated up to the level of approximately 90 deg. C before the actual experiments. The initial water bulk temperature was 20 deg. C. When the steam flow rate was low enough (typically approx100-150 g/s) temperatures below the blowdown pipe outlet remained constant while increasing heat-up occurred towards the pool surface layers indicating strong thermal stratification of the wet well pool water. During the stratification period the highest measured temperature difference between pool bottom and surface was approximately 40 deg. C. During the mixing period total mixing of the pool volume was not achieved in any of the experiments. The bottom layers heated up significantly but never reached the same temperature as the topmost layers. The lowest measured temperature difference between the pool bottom and surface was 7-8 deg. C. According to the test results, it seems that a small void fraction doesn't have an effect on the speed of sound in water and that the acquired sound velocity measurement system cannot be used for the estimation of void fraction in the wet well water pool. However, more tests on this issue have to be

  5. PPOOLEX experiments on thermal stratification and mixing

    International Nuclear Information System (INIS)

    Puustinen, M.; Laine, J.; Raesaenen, A.

    2009-08-01

    The results of the thermal stratification experiments in 2008 with the PPOOLEX test facility are presented. PPOOLEX is a closed vessel divided into two compartments, dry well and wet well. Extra temperature measurements for capturing different aspects of the investigated phenomena were added before the experiments. The main purpose of the experiment series was to generate verification data for evaluating the capability of GOTHIC code to predict stratification and mixing phenomena. Altogether six experiments were carried out. Heat-up periods of several thousand seconds by steam injection into the dry well compartment and from there into the wet well water pool were recorded. The initial water bulk temperature was 20 deg. C. Cooling periods of several days were included in three experiments. A large difference between the pool bottom and top layer temperature was measured when small steam flow rates were used. With higher flow rates the mixing effect of steam discharge delayed the start of stratification until the pool bulk temperature exceeded 50 deg. C. The stratification process was also different in these two cases. With a small flow rate stratification was observed only above and just below the blowdown pipe outlet elevation. With a higher flow rate over a 30 deg. C temperature difference between the pool bottom and pipe outlet elevation was measured. Elevations above the pipe outlet indicated almost linear rise until the end of steam discharge. During the cooling periods the measurements of the bottom third of the pool first had an increasing trend although there was no heat input from outside. This was due to thermal diffusion downwards from the higher elevations. Heat-up in the gas space of the wet well was quite strong, first due to compression by pressure build-up and then by heat conduction from the hot dry well compartment via the intermediate floor and test vessel walls and by convection from the upper layers of the hot pool water. The gas space

  6. Nutrient limitation in tropical secondary forests following different management practices.

    Science.gov (United States)

    Nagy, R Chelsea; Rastetter, Edward B; Neill, Christopher; Porder, Stephen

    2017-04-01

    Secondary forests now make up more than one-half of all tropical forests, and constraints on their biomass accumulation will influence the strength of the terrestrial carbon (C) sink in the coming decades. However the variance in secondary tropical forest biomass for a given stand age and climate is high and our understanding of why is limited. We constructed a model of terrestrial C, nitrogen (N), and phosphorus (P) cycling to examine the influence of disturbance and management practices on nutrient limitation and biomass recovery in secondary tropical forests. The model predicted that N limited the rate of forest recovery in the first few decades following harvest, but that this limitation switched to P approximately 30-40 yr after abandonment, consistent with field data on N and P cycling from secondary tropical forest chronosequences. Simulated biomass recovery agreed well with field data of biomass accumulation following harvest (R 2  = 0.80). Model results showed that if all biomass remained on site following a severe disturbance such as blowdown, regrowth approached pre-disturbance biomass in 80-90 yr, and recovery was faster following smaller disturbances such as selective logging. Field data from regrowth on abandoned pastures were consistent with simulated losses of nutrients in soil organic matter, particularly P. Following any forest disturbance that involved the removal of nutrients (i.e., except blowdown), forest regrowth produced reduced biomass relative to the initial state as a result of nutrient loss through harvest, leaching and/or sequestration by secondary minerals. Differences in nutrient availability accounted for 49-94% of the variance in secondary forest biomass C at a given stand age. Management lessons from this study are the importance of strategies that help retain nutrients on site, recognizing the role of coarse woody debris in immobilization and subsequent release of nutrients, and the potential for nutrient additions to enhance

  7. Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon

    2011-01-01

    Research highlights: → The Chinshan Mark I containment pressure-temperature responses are analyzed. → GOTHIC is used to calculate the containment responses under three pipe break events. → This study is used to support the Chinshan Stretch Power Uprate (SPU) program. → The calculated peak pressure and temperature are still below the design values. → The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 o C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 o C). Additionally, the peak drywell temperature of 155.3 o C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 o C, which is below the pool temperature used for evaluating the

  8. Development of automated lance systems for removing deposited sludge around heat transfer tubes with a trianglar pattern in a steam

    International Nuclear Information System (INIS)

    Hwang, K. S.; Sung, H. J.; Jeong, W. T.; Hong, S. Y.; Park, Y. S.

    2003-01-01

    Automated lance systems have been developed to remove sludge deposits filed up around the heat transfer tubes of a triangular pattern in a steam generator. The accessible ways of the lance systems inside the steam generator are the annulus region which occupies the space between the outermost tubes and the inner wall of the steam generator, and the Blowdown Lane region (BdL) without tubes along the centerline of the steam generator. The lance system along the annulus employes a slidable guide support rail and a lance body. The guide support rail, which is composed of two parallel circular rods with a vertical distance, is tightly fixed inside the hand holes. The guide support rail extends from a hand hole at 0 degree to the other hand hole at 180 degree. The lance body is slideably held on the guide support rail by means of supporting holders which are attached on both the bottom and the upper plates of the lance body. The lance body is comprised of a nozzle block with a nozzle cylinder and a first drive means which makes sweeping motion of the nozzle cylinder, a second drive means which aligns the direction of nozzle jets from the nozzle cylinder toward the desired tube lanes by rotating the nozzle block in the horizontal plane, and two side wall supporting wheel assemblies attached to the outer surface of the lance body, rolling along the inner wall of the steam generator. For the transportation of the lance, two control cables which extend outward through the hand holes are attached to both ends of the lance body and are driven by a drive means with a powered drum. The lance system along the blowdown lane adopts a horizontal guide support rail and a lance body which can convey three nozzle blocks for emitting high pressure water in the 30, 90 and 150 degree directions. By utilizing the above two lance systems, the shadow zones around the tubes where the high pressure water does not reach are highly reduced

  9. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Gi Hong [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-12-15

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  10. Applicability of small-scale integral test data to the 4500 MWt ESBWR loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Saha, Pradip; Gamble, Robert E.; Shiralkar, Bharat S.; Fitch, James R.

    2009-01-01

    This paper discusses the scaling methodology used by GE Hitachi Nuclear Energy (GEH) to show that the data obtained from the small-scale integral test facilities, namely, GIST and GIRAFFE-SIT, are relevant to the postulated loss-of-coolant accident (LOCA) of the 4500 MWt ESBWR. The conservation of mass and energy equations for the steam-water mixture in the reactor pressure vessel (RPV) are transformed to the equations for the rates of pressure change and water mass or inventory change. These equations are non-dimensionalized based on the most dominant physical processes of the individual stages of a LOCA, namely, the late blowdown stage, the GDCS (gravity driven cooling system) transition stage and the full GDCS stage. The magnitudes of the non-dimensional Pi-groups, obtained from these equations, for the 4500 MWt ESBWR are compared with those obtained for the small-scale integral tests mentioned earlier. In addition, simplified analyses were conducted for the first two stages by integrating the non-dimensional RPV depressurization rate and the water inventory change rate equations. The results of the 4500 MWt ESBWR are very similar to the test data obtained from the GIST and the GIRAFFE-SIT test facilities. Therefore, based on both the Pi-group magnitudes and the simplified analyses, it is concluded that the small-scale integral test data mentioned above are applicable to the 4500 MWt ESBWR LOCA applications.

  11. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

    1993-03-01

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  12. Through analysis of LOFT L2-2 by THYDE-P code, (1)

    International Nuclear Information System (INIS)

    Hirano, Masashi; Asahi, Yoshiro

    1981-06-01

    A Through analysis of the Test L2-2 loss-of-coolant experiment (LOCE) in the Loss-of-Fluid Test (LOFT) program was made by the THYDE-P code. LOFT Test L2-2 was the first test in the Power Ascension Test Series (Test Series L2) of nuclear full double-ended cold leg break tests. THYDE-P is a computer code to analyze both blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs) and is now under verification study and modifications. Therefore, the LOFT experimental data play an important role at the present stage of the THYDE-P code. The present analysis was performed by best estimate (BE) options as sample calculation Run 30, which is a portion of a series of THYDE-P sample calculations. In this report, the calculated results are compared with the experimental data and discussed. In the present calculation, the core nodes were completely submerged with subcooled water at 55 sec. after the test initiation. It showed a good agreement with the experimental result. (author)

  13. A study on the steam generator data base and the evaluation of chemical environment

    International Nuclear Information System (INIS)

    Yang, Kyung Rin; Yoo, Je Hyoo; Lee, Eun He; Hong, Kwang Pum

    1990-01-01

    In order to make steam generator data base, the basic plant information and water quality control data on the steam generators of the PWR nuclear power plant operating in the world have been collected by EPRI. In this project, the basic information and water quality control data of the domestic PWR nuclear power plants were collected to make steam generator data base on the basic of the EPRI format table, and the computerization of them was performed. Also, the technical evaluation of chemical environments on steam generator of the Kori 2 plant chemists. Workers and researchers working at the research institute and universities and so on. Especially, it is able to be used as a basic plant information in order to develop an artificial intellegence development system in the field on the technical development of the chemical environment. The scope and content of the project are following. The data base on the basic information data in domestic PWR plant. The steam generator data base on water quality control data. The evaluation on the chemical environment in the steam generators of the Kori 2 plant. From previous data, it is concluded as follows. The basic plant information on the domestic PWR power plant were computerized. The steam generator data base were made on the basis of EPRI format table. The chemical environment of the internal steam generators could be estimated from the analytical evaluation of water quality control data of the steam generator blowdown. (author)

  14. Experiment data report for Semiscale Mod-1 tests S-05-2A and S-05-2B (alternate ECC injection tests)

    Energy Technology Data Exchange (ETDEWEB)

    Patton, Jr., M. L.; Collins, B. L.; Sackett, K. E.

    1977-04-01

    Recorded test data are presented for Tests S-05-2A and S-05-2B of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-05-2A and S-05-2B were conducted from initial conditions of 2263 psia and 543/sup 0/F and 2272 psia and 542/sup 0/F, respectively, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the tests, cooling water was injected into the intact loop pump suction and broken loop cold leg to simulate emergency core coolant injection in a PWR with flow rates based on system volume scaling. For Test S-05-2A the intact loop pump speed was held constant throughout the test at the initial blowdown value. During Test S-05-2B the pump speed was reduced and stopped according to a predetermined coastdown schedule.

  15. RCGVS design improvement and depressurization capability tests for Ulchin nuclear power plant units 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Kang Sik; Seong, Ho Je; Jeong, Won Sang; Seo, Jong Tae; Lee, Sang Keun [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Lim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [Korea Electric Power Cooperation, Taejon (Korea, Republic of)

    1998-12-31

    The Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) has been improved from the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 and 4) based on the evaluation results for depressurization capability tests performed at YGN 3 and 4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown phenomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifics size for UCN 3 and 4 has been reduced to 9/32 inch from the 11/32 inch for YGN 3 and 4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3 and 4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation. 6 refs., 5 figs., 1 tab. (Author)

  16. Strategic elements of steam cycle chemistry control practices at TXU's Comanche Peak steam electric station

    International Nuclear Information System (INIS)

    Fellers, B.; Stevens, J.; Nichols, G.

    2002-01-01

    Early industry experience defined the critical importance of Chemistry Control Practices to maintaining long-term performance of PWR steam generators. These lessons provided the impetus for a number of innovations and alternate practices at Comanche Peak. For example, advanced amine investigations and implementation of results provided record low iron transport and deposition. The benefits of the surface-active properties of dimethyl-amine exceeded initial expectations. Operation of pre-coat polishers and steam generator blowdown demineralizers in the amine cycle enabled optimization of amine concentrations and stable pH control. The strategy for coordinated control of oxygen and hydrazine dosing complemented the advanced amine program for protective oxide stabilization. Additionally, a proactive chemical cleaning was performed on Unit 1 to prevent degradations from general fouling of steam generator tube-tube support plate (TSP) and top-of-tubesheet (TTS) crevices. This paper shares the results of these innovations and practices. Also, the bases, theory, and philosophy supporting the strategic elements of program will be presented. (authors)

  17. Supersonic Cavity-Based Flow Control Using a Quasi-DC Discharge

    Science.gov (United States)

    Houpt, A.; Leonov, S.; Hedlund, B.; Ombrello, T.; Carter, C.

    2017-10-01

    The Quasi-DC (Q-DC) discharge is studied as an active flow control authority on a rear-facing cavity in a supersonic duct by creating an oblique shockwave that impinges the cavity. This geometry simulates the geometry of a typical scramjet flameholding scheme. The tests were performed at the University of Notre Dame in the SBR-50 supersonic blowdown rig with dried air at M=2. Schlieren imaging is used to view the flow field with and without the Q-DC discharge in operation. A significant change in the flow field structure is observed. Pressure sensors detect a pressure increase throughout the entire rear-facing cavity while the Q-DC discharge is operating. This reveals that the cavity redistributes the pressure increase from the shockwave as a result of the flow within the cavity being subsonic. As a result of this pressure absorption and redistribution, the impinging shockwave created by the Q-DC is almost completely absorbed. This absorption is confirmed by the schlieren images. The data reveal that the discharge power is the dominating influence, as compared to electrode/discharge geometry, on the pressure increase produced in the cavity. There is a nearly linear correlation between the power of the discharge and the pressure increase produced directly behind the discharge, in the cavity, and on the ramp of the cavity (to varying magnitudes). It is suggested that the 11 electrode system may be slightly more effective than the 7 electrode system.

  18. Better understanding flow-restricted environments from hideout return analyses

    International Nuclear Information System (INIS)

    Ollar, Ph.; Viricel-Honorez, L.

    1998-01-01

    Understanding and controlling the chemical phenomena in flow-restricted areas of the secondary side of steam generators (SG) is a key point in fighting against corrosion (IGA) in these areas. Hence, more than 150 hideout returns from French plants are used to assess the chemical environment in flow-restricted areas. The influence of several parameters such as tube support plates design, SG blowdown flowrate and temperature at the end of hideout returns, on returned weights and on flow-restricted environments predicted by modelling are examined. A correlation between different operating and returned chemistry indexes (molar ratios), and IGA is sought: no link is found between IGA and Na/Cl molar ratio at the beginning of hideout returns. However, as IGA extends, the returned weights of sodium and chloride seem to decrease, suggesting a greater hideout and a lower hideout return efficiency, perhaps due to SG fouling, and/or a smaller hideout/hideout return thanks to an improved operating chemistry. Using the pH computed from hideout return data as an index for the presence of IGA is also studied: no correlation is found between the two. Improving plants chemistry is finally proposed, based on an optimization of hideout returns. (authors)

  19. Clean Firetube Boiler Waterside Heat Transfer Surfaces, Energy Tips: STEAM, Steam Tip Sheet #7 (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    2012-04-01

    A steam energy tip sheet for the Advanced Manufacturing Office (AMO). The prevention of scale formation in firetube boilers can result in substantial energy savings. Scale deposits occur when calcium, magnesium, and silica, commonly found in most water supplies, react to form a continuous layer of material on the waterside of the boiler heat exchange tubes. Scale creates a problem because it typically possesses a thermal conductivity, an order of magnitude less than the corresponding value for bare steel. Even thin layers of scale serve as an effective insulator and retard heat transfer. The result is overheating of boiler tube metal, tube failures, and loss of energy efficiency. Fuel consumption may increase by up to 5% in firetube boilers because of scale. The boilers steam production may be reduced if the firing rate cannot be increased to compensate for the decrease in combustion efficiency. Energy losses as a function of scale thickness and composition are given. Any scale in a boiler is undesirable. The best way to deal with scale is not to let it form in the first place. Prevent scale formation by: (1) Pretreating of boiler makeup water (using water softeners, demineralizers, and reverse osmosis to remove scale-forming minerals); (2) Injecting chemicals into the boiler feedwater; and (3) Adopting proper boiler blowdown practices.

  20. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  1. Analysis of loss of coolant accident and emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

    1977-01-01

    In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

  2. Test Capabilities and Recent Experiences in the NASA Langley 8-Foot High Temperature Tunnel

    Science.gov (United States)

    Hodge, Jeffrey S.; Harvin, Stephen F.

    2000-01-01

    The NASA Langley 8-Foot High Temperature Tunnel is a combustion-heated hypersonic blowdown-to-atmosphere wind tunnel that provides flight enthalpy simulation for Mach numbers of 4, 5, and 7 through an altitude range from 50,000 to 120,000 feet. The open-.jet test section is 8-ft. in diameter and 12-ft. long. The test section will accommodate large air-breathing hypersonic propulsion systems as well as structural and thermal protection system components. Stable wind tunnel test conditions can be provided for 60 seconds. Additional test capabilities are provided by a radiant heater system used to simulate ascent or entry heating profiles. The test medium is the combustion products of air and methane that are burned in a pressurized combustion chamber. Oxygen is added to the test medium for air-breathing propulsion tests so that the test gas contains 21 percent molar oxygen. The facility was modified extensively in the late 1980's to provide airbreathing propulsion testing capability. In this paper, a brief history and general description of the facility are presented along with a discussion of the types of supported testing. Recently completed tests are discussed to explain the capabilities this facility provides and to demonstrate the experience of the staff.

  3. Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

    International Nuclear Information System (INIS)

    Strong, B.R. Jr.; Baschiere, R.J.

    1978-01-01

    The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the 'force equivalent area' (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques. (Auth.)

  4. COMPARE: a computer program for the transient calculation of a system of volumes connected by flowing vents

    International Nuclear Information System (INIS)

    Gido, R.G.; Grimes, C.I.; Lawton, R.G.; Kudrick, J.A.

    1976-09-01

    A description is given of the COMPARE computer program developed for performing transient subcompartment pressure response analyses of nuclear power plants. The subcompartments are represented as volumes (less than or equal to 100) which are connected by junctions (less than or equal to 200) and may have blowdown (less than or equal to 5 sets). The volume thermodynamics and flow equations are for a homogeneous mixture, assumed to be in thermodynamic equilibrium consisting of any one, or any combination, of the following: (a) steam, (b) two-phase water, and (c) any three perfect gases such as air, helium, etc. Flow between volumes is based on (a) the Moody equation, with an arbitrary multiplier, when the flow is critical, (b) compressible, polytropic, orifice flow of an ideal gas-like mixture when the flow is subcritical, and (c) an incompressible subelement method when inertia effects exist. A quasi-static explicit numerical solution technique is used. The program requires 40,000 words on the LASL CDC-7600 and 124,000 10 bytes on an IBM 360/370 computer. A two-volume, one-junction problem requires 0.002 s per time step on the CDC-7600 and 0.012 s on the IBM 360/370

  5. Risk from a pressurized toxic gas system: Part 2, Dispersal consequences

    International Nuclear Information System (INIS)

    Brereton, S.J.; Altenbach, T.J.; Lane, S.G.; Martin, D.W.

    1995-02-01

    During the preparation of a Safety Analysis Report at the Lawrence Livermore National Laboratory. we studied the release of chlorine from a compressed gas experimental apparatus. This paper presents the second pan in a series of two papers on this topic. The first paper focuses on the frequency of an unmitigated release from the system; paper focuses the consequences of the release. The release of chlorine from the experimental apparatus was modeled as an unmitigated blowdown through a 0.25 inch (0.006.4 m) outside diameter tube. The physical properties of chlorine were considered as were the dynamics of the fluid flow problem. The calculated release rate was used as input for the consequence assessment. Downwind concentrations as a function of time were evaluated and then compared to suggested guidelines. For comparison purposes, a typical water treatment plant was briefly studied. The lower hazard presented by the LLNL operation becomes evident when its release is compared to the release of material from a water treatment plant, a hazard which is generally accepted by the public

  6. Safety research for CANDU reactors

    International Nuclear Information System (INIS)

    Hancox, W.T.

    1982-10-01

    Continuing research to develop and verify computer models of CANDU-PHW reactor process and safety systems is described. It is focussed on loss-of-coolant accidents (LOCAs) because they are the precursors of more serious accidents. Research topics include: (i) fluid-dynamic and heat-transfer processes in the heat transport system during the blowdown and refilling phases of LOCAs; (ii) thermal and mechanical behaviour of fuel elements; (iii) thermal and mechanical behaviour of the fuel and the fuel-channel assembly in situations where the heavy-water moderator is the sink for decay heat produced in the fuel; (iv) chemical behaviour of fission gases that might be released into the reactor coolant and transported to the containment system; and (v) combustion of hydrogen-air-steam mixtures that would be produced if fuel temperatures were sufficiently high to initiate the zirconium-water reaction. The current status of the research on each of these topics is highlighted with particular emphasis on the conclusions reached to date and their impact on the continuing program

  7. Analysis of transient flow boiling: application of the method of characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Hancox, W.T. (Atomic Energy of Canada, Ltd., Manitoba); Mathers, W.G.; Kawa, D.

    1978-01-01

    An analysis, based on the method of characteristics, was deveoped for transient flow boiling; particular emphasis was placed on blowdown from subcooled liquid conditions. The governing equations are quasilinear partial differential equations of the hyperbolic type which are derived from the conservation laws assuming one-dimensional homogeneous thermal equilibrium flow. Using a wave tracing procedure, the solution is advanced at grid points in the spacetime plane which are the intersections of the characteristic curves. The contribution of the present paper is in the application of this technique, used extensively in gas dynamics, to flow boiling dynamics. Special procedures were developed to handle discontinuities and complex wave interactions which arise owing to phase changes. Wave diagrams obtained using the wave tracing technique are discussed and compared with experimental data. The more complex forms of the one-dimensional homogeneous conservation laws and required constitutive equations, which incorporate departures from thermal equilibrium are also presented. The extension of the method of characteristics to these systems of equations is discussed.

  8. Environmental evaluation of a nuclear power plant on Lake Erie: some aquatic impacts

    International Nuclear Information System (INIS)

    Reutter, J.M.

    1976-01-01

    The Toledo Edison Company and The Cleveland Electric Illuminating Company are currently building the first unit of the Davis--Besse Nuclear Power Station on the south shore of Lake Erie at Locust Point. This plant will utilize water from Lake Erie for cooling purposes and to replenish the cooling tower blowdown which will be returned to the lake at the maximum of 11.1 0 C above ambient. Laboratory experiments were conducted on Lake Erie fish to determine their seasonal final temperature preferenda and the effects of sudden temperature changes. These results were correlated with the existing fish community at Locust Point to predict the effects of the thermal discharge from the Davis--Besse Nuclear Power Station on the surrounding fish community and to predict the effects of thermal discharges, in general, on the Lake Erie fishery resource. Phytoplankton populations, zooplankton populations and several water quality parameters were measured at Locust Point in 1974 and 1975. Correlation coefficients and multiple regressions were computed using these plankton populations and the water quality parameters to determine the extent to which the existing plankton populations are effected by the water quality. Over 2000 fish representing 24 species were tested to determine their final temperature preferenda

  9. Aerodynamics, heat and mass transfer in steam-aerosol turbulent flows in containment

    Energy Technology Data Exchange (ETDEWEB)

    Nigmatulin, B.I.; Pershukov, V.A.; Ris, V.V. [Research & Engineering Centre of Nuclear Plants Safety, Moscow (Russian Federation)] [and others

    1995-09-01

    In this report an analysis of aerodynamic and heat transfer processes at the blowdown of gas-dispersed mixture into the containment volume is presented. A few models for description of the volume averaged and local characteristics are analyzed. The mathematical model for description of the local characteristics of the turbulent gas-dispersed flows was developed. The calculation of aerodynamic, heat and mass transfer characteristics was based on the Navier-Stokes, energy and gas mass fractions conservation equations. For calculation of dynamics and deposition of the aerosols the original diffusion-inertia model is developed. The pulsating characteristics of the gaseous phase were calculated on the base (k-{xi}) model of turbulence with modification to account thermogravitational force action and influence of particle mass loading. The appropriate boundary conditions using the {open_quotes}near-wall function{close_quotes} approach was obtained. Testing of the mathematical models and boundary conditions has shown a good agreement between computation and data of comparison. The described mathematical models were applied to two- and three dimensional calculations of the turbulent flow in containment at the various stages of the accident.

  10. Perkins Nuclear Station, Units 1, 2, and 3: Final environmental statement (Docket Nos. STN 50-488, STN 50-489, and STN 50-490

    International Nuclear Information System (INIS)

    1975-10-01

    The proposed action is the issuance of a construction permit to the Duke Power Company for the construction of the Perkins Nuclear Station (PNS) Units 1, 2, and 3 located in Davie County, North Carolina. A total of 2402 acres will be used for the PNS site; another 1401 acres will be used for the Carter Creek Impoundment. Construction-related activities on the primary site will disturb about 617 acres. Approximately 631 acres of land will be required for transmission line right-of-way, and a railroad spur will affect 77 acres. This constitutes a minor local impact. The heat dissipation system will require a maximum water makeup of 55,816 gpm, of which 50,514 gpm will be consumed due to drift and evaporative losses. This amount represents 4% of the mean monthly flow of the Yadkin River. The cooling tower blowdown and chemical effluents from the station will increase the dissolved solids concentration in the Yadkin River by a maximum of 18 ppm. The thermal alterations and increases in total dissolved solids concentration will not significantly affect the aquatic productivity of the Yadkin River. 26 figs., 51 tabs

  11. The perturbation theory in the fundamental mode. Its application to the analysis of neutronic experiments involving small amounts of materials in fast neutron multiplying media

    International Nuclear Information System (INIS)

    Remsak, Stanislav.

    1975-01-01

    The formalism of the perturbation theory at the first order, is developed in its simplest form: diffusion theory in the fundamental mode and then the more complex formalism of the transport theory in the fundamental mode. A comparison shows the effect of the angular correlation between the fine structures of the flux and its adjoint function, the difference in the treatment of neutron leakage phenomena, and the existence of new terms in the perturbation formula, entailing a reactivity representation in the diffusion theory that is not quite exact. Problems of using the formalism developed are considered: application of the multigroup formalism, transients of the flux and its adjoint function, validity of the first order approximation etc. A detailed analysis allows the formulation of a criterion specifying the validity range. Transients occuring in the reference medium are also treated. A set of numerical tests for determining a method of elimination of transient effects is presented. Some differential experiments are then discussed: sodium blowdown in enriched uranium or plutonium cores, experiments utilizing some structural materials (iron and oxygen) and plutonium sample oscillations. The Cadarache version II program was systematically used but the analysis of the experiments of plutonium sample oscillation in Ermine required the Cadarache version III program [fr

  12. Applicability of small-scale integral test data to the 4500 MWt ESBWR loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Saha, Pradip [GE Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, NC 28401 (United States)], E-mail: pradip.saha@ge.com; Gamble, Robert E.; Shiralkar, Bharat S.; Fitch, James R. [GE Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, NC 28401 (United States)

    2009-05-15

    This paper discusses the scaling methodology used by GE Hitachi Nuclear Energy (GEH) to show that the data obtained from the small-scale integral test facilities, namely, GIST and GIRAFFE-SIT, are relevant to the postulated loss-of-coolant accident (LOCA) of the 4500 MWt ESBWR. The conservation of mass and energy equations for the steam-water mixture in the reactor pressure vessel (RPV) are transformed to the equations for the rates of pressure change and water mass or inventory change. These equations are non-dimensionalized based on the most dominant physical processes of the individual stages of a LOCA, namely, the late blowdown stage, the GDCS (gravity driven cooling system) transition stage and the full GDCS stage. The magnitudes of the non-dimensional Pi-groups, obtained from these equations, for the 4500 MWt ESBWR are compared with those obtained for the small-scale integral tests mentioned earlier. In addition, simplified analyses were conducted for the first two stages by integrating the non-dimensional RPV depressurization rate and the water inventory change rate equations. The results of the 4500 MWt ESBWR are very similar to the test data obtained from the GIST and the GIRAFFE-SIT test facilities. Therefore, based on both the Pi-group magnitudes and the simplified analyses, it is concluded that the small-scale integral test data mentioned above are applicable to the 4500 MWt ESBWR LOCA applications.

  13. Evaporative processes for desalination of produced water; Processos evaporativos para dessalinizacao de agua produzida a fins de reuso

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Vivian T.; Dezotti, Marcia W. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Quimica; Schuhli, Juliana B.; Gomes, Marcia T.; Pereira Junior, Oswaldo A. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

    2008-07-01

    During the productive life of an oil well, it gets the moment when a big quantity of produced water comes together with the oil. It can achieve 99% in the end of its economical life. The thermal desalination of the formation water is one of the most common technologies for achieving its reuse. This way, it was constructed one 'Robert' evaporator. The tests used different sodium chloride concentrations from 2,000 mg/L to 80,000 mg/L simulating concentrations found in the produced water from PETROBRAS wells. The tests were conducted in three different vacuum pressures. It was observed, increasing the vacuum applied to the system, results in reduction of solution boiling point. The salt concentrations of the brine blowdown were influenced by the sodium chloride concentration at the feed flow, by the vacuum applied to the system and, consequently, by the solution boiling point and flow rates. The produced distillate water presented sodium chloride concentration lower than 2 mg/L, indicating that this system can produce water to reuse in irrigation. (author)

  14. Basic investigation on promotion of joint implementation in fiscal 2000. Efficiency improvement project for district heat supplying plants in Dailian City in China; 2000 nendo kyodo jisshi nado suishin kiso chosa hokokusho. Chugoku/Dailian shi chiiki netsu kyokyu plant kokoritsu kaizen project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    Investigations and discussions have been given on energy saving possibilities at two medium-sized heat and power supplying plants in the city of Dailian in China. The project will improve the operation methods of the heat and power plants so that the energy cost can be minimized, and attempt to improve the boiler heat efficiency and save the energy by means of heat recovery and utilization. The draft modification plan for energy conservation has planned operation optimization for energy conservation, control of boiler operation under variable pressure, modification of the external boiler heat converter, use of inverters for the large capacity motors for boilers, and recovery of heat from the boiler blow-down water. In the analysis, models were structured from the operation data, and the effects of applying the energy saving measures were derived from simulation. As a result, the energy saving effect was found to be about 13,000 tons at the Chunhai plant and about 7,000 tons at the Pulandian plant annually (converted to oil). The reduction in greenhouse gas emission was found to be about 40,000 tons at the Chunhai plant and about 20,000 tons at the Pulandian plant annually. The number of years for investment payback is about 4.1 years at the Chunhai plant, and about 4.9 years at the Pulandian plant, wherein good profitability can be estimated. (NEDO)

  15. The influence of non condensible gas on two phase critical flow

    International Nuclear Information System (INIS)

    Celata, G.P.; Cumo, M.; D'Annibale, F.; Farello, G.E.

    1987-01-01

    With reference to Loss-of-Coolant Accidents in Pressurized Water Reactors and in the frame of the wide scientific landscape of blowdown experiments aiming to the improvement of two-phase critical flows knowledge, it is of interest the analysis of non condensible gas influence on the critical flow (radiolytic gases,metal-water reactions products etc.). The present paper deals with an experiment referring to two-phase steam-water critical flows from long tubes, in which known air flowrates are injected in the stagnation region. The aim of the experiment is to detect the influence of non-condensible gas on the two-phase critical flow behaviour (critical mass flow rate, pressure and temperature profiles along the discharge channel etc.) as well as to individuate the limit, in terms of air concentration, beyond which the critical flow is affected by the presence of the gas. The employed test section is a vertical, circular duct channel with an inner diameter of 4.6 mm and a length of 1500 mm (L/D = 325). Results of initially subcooled liquid experiments (together with some data of satured liquid discharges), up to 15 bars are reported with the analysis of non-condensible effects in the different stagnation conditions

  16. Summary of particle bed reactor designs for the Space Nuclear Thermal Propulsion Program

    Science.gov (United States)

    Powell, J. R.; Ludewig, H.; Todosow, M.

    1993-09-01

    A summary report of the Particle Bed Reactor (PBR) designs considered for the space nuclear thermal propulsion program has been prepared. The first chapters outline the methods of analysis, and their validation. Monte Carlo methods are used for the physics analysis, several new algorithms are used for the fluid dynamics heat transfer and engine system analysis, and commercially available codes are used for the stress analysis. A critical experiment, prototypic of the PBR was used for the physics validation, and blowdown experiments using fuel beds of prototypic dimensions were used to validate the power extraction capabilities from particle beds. In all four different PBR rocket reactor designs were studied to varying degrees of detail. They varied in power from 400 MW to 2000 MW. These designs were all characterized by a negative prompt coefficient, due to Doppler feedback, and the feedback due to moderator heat up varied from slightly negative to slightly positive. In all practical cases, the coolant worth was positive, although core configurations with negative coolant worth could be designed. In all practical cases the thrust/weight ratio was greater than 20.

  17. Industrial energy thrift scheme. Report No. 27. Energy use in the rubber, and linoleum and plastics floor-covering sectors

    Energy Technology Data Exchange (ETDEWEB)

    1981-05-01

    A total of 374 establishments manufacturing rubber products and 42 establishments producing linoleum or other coated substrates were invited to take part in the scheme, leading to 77 and 16 visits respectively. Both sectors manufacture products which contain as a main ingredient polymers of high molecular weight. For purposes of comparison, the different types of products have been grouped under five product headings: new tires, remoulded tires, solid rubber based products, latex based products, and coated substrates. Five sources of energy were used at the sites visited. Fuel oil was the main source, supplying 49 percent of the total, followed by gas with 25 percent. Coal, supplying only 10 percent of the total, contributed much less than it did in the past. Process wastes and reject products, used as fuels, contributed only 1.5 percent to the total supply of energy - but could supply much more. Within the 93 sites visited 760 opportunities for saving energy were noted; nineteen different types of opportunity were observed in at least 20 percent of the sites. Opportunities for saving energy, in descending order of significance, lay in the recovery of waste heat from manufacturing processes, in the reduction of heat loss from process equipment, in the improvement of insulation of buildings, in better control of space heating, and in the recovery of heat from the blowdown of boilers.

  18. Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

    International Nuclear Information System (INIS)

    Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

    1987-01-01

    From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in the reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 mPA were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. On account of the safety margins proved in the experiments, potential inaccuracies in theoretical structure analyses are recommended so as to be on the safe side. On the other hand, it appears that designing pipework with reference to elastic stress categories does not adequately take into account the actual reserves of the pipework material

  19. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    International Nuclear Information System (INIS)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6'' cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author)

  20. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  1. Analysis of LOFT (L1-2) experiment by code RELAP-4J

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Soda, Kunihisa; Shiba, Masayoshi; Kaminaga, Humito

    1977-04-01

    An analysis of the results in LOFT L1-2 LOCE (Loss of Coolant Experiment) was made by the computer code RELAP-4J. The L1-2 experiment is a simple isothermal blowdown test with a core simulator and no ECC activation. It provides the basis for future LOCE with a nuclear core and ECC activation. The results of the analysis lead to the following conclusions. (1) The calculated system pressure transient agrees well with experiment. Primary controlling factors for the calculation are (a) flow resistances of the steam generator simulator, pump simulator and discharge nozzle in the broken loop, (b) mixture level in the downcomer and inlet volume of the operating loop steam generator, and (c) stored heat of the downcomer structure. (2) The pressurizer pressure decreases rapidly, compared with experiment, possibly because the flow resistance in the surge line is smaller than the actual one. Further experiment and analysis are necessary in this respect. (3) The calculated density transient in the cold leg agrees well with experiment. Agreement is not good in the hot leg, however. The discrepancy is possibly caused by the non-homogeneous flow of coolant in the hot leg due to low flow rate. (4) Effect of the pump characteristics on analytical result is insignificant in the isothermal test. However, in the future nuclear test, the effect will be significant because of large steam generation in the core, so measurement of the pump characteristics and improvement of the pump model are necessary. (auth.)

  2. RELAP5/MOD2 implementation on various mainframes including the IBM and SX-2 supercomputer

    International Nuclear Information System (INIS)

    DeForest, D.L.; Hassan, Y.A.

    1987-01-01

    The RELAP5/MOD2 (cycle 36.04) code is a one-dimensional, two-fluid, nonequilibrium, nonhomogeneous transient analysis code designed to simulate operational and accident scenarios in pressurized water reactors (PWRs). System models are solved using a semi-implicit finite difference method. The code was developed at EG and G in Idaho Falls under sponsorship of the US Nuclear Regulatory Commission (NRC). The major enhancement from RELAP5/MOD1 is the use of a six-equation, two-fluid nonequilibrium and nonhomogeneous model. Other improvements include the addition of a noncondensible gas component and the revision and addition of drag formulation, wall friction, and wall heat transfer. Several test cases were run to benchmark the IBM and SX-2 installations against the CDC computer and the CRAY-2 and CRAY/XMP. These included the Edward's pipe blow-down and two separate reflood cases developed to simulate the FLECHT-SEASET reflood test 31504 and a postcritical heat flux (CHF) test performed at Lehigh University

  3. Experimental study and modelization of a propane storage tank depressurization

    International Nuclear Information System (INIS)

    Veneau, Tania

    1995-01-01

    The risks associated with the fast depressurization of propane storage tanks reveals the importance of the 'source term' determination. This term is directly linked, among others, to the characteristics of the jet developed downstream of the breach. The first aim of this work was to provide an original data bank concerning drop velocity and diameter distributions in a propane jet. For this purpose, a phase Doppler anemometer bas been implemented on an experimental set-up. Propane blowdowns have been performed with different breach sizes and several initial pressures in the storage tank. Drop diameter and velocity distributions have been investigated at different locations in the jet zone. These measurements exhibited the fragmentation and vaporisation trends in the jet. The second aim of this work concerned the 'source term'. lt required to study the coupling between the fluid behaviour inside the tank and the flow through the breach. This model took into account the phase exchange when flashing occurred in the tank. The flow at the breach was described with an homogeneous relaxation model. This coupled modelization has been successfully and exhaustively validated. lt originality lies on the application to propane flows. (author) [fr

  4. Report to Congress on abnormal occurrences: April--June 1995. Volume 18, Number 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence (AO) as an unscheduled incident or event that the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such occurrences to be made to Congress. This report provides a description of those incidents and events that have been determined to be AOs during the period of April 1 through June 30, 1995. This report addresses five AOs at NRC-licensed facilities. One involved a reactor coolant system blowdown at a pressurized water reactor (PWR) nuclear power plant, one involved a previously unidentified path for the potential release of radioactivity at a PWR nuclear power plant, two involved medical brachytherapy misadministrations, and one involved a medical therapeutic radiopharmaceutical misadministration. Four AOs submitted by the Agreement States are included. One involved a medical teletherapy misadministration, two involved medical brachytherapy misadministrations, and one involved the overexposure of personnel at a medical center. The report also contains an update of one AO previously reported by an NRC licensee, and two AOs previously reported by the Agreement States. No ``Other Events of Interest`` items are being reported.

  5. LOCA assessment experiments in a full-elevation, CANDU-typical test facility

    International Nuclear Information System (INIS)

    Ingham, P.J.; McGee, G.R.; Krishnan, V.S.

    1990-01-01

    The RD-14 thermal-hydraulics test facility, located at the Whiteshell Nuclear Research Establishment, is a full-elevation model representative of a CANDU primary heat transport system. The facility is scaled to accommodate a single, full-scale (5.0 MW, 21 kg/s), electrically heated channel per pass. The steam generators, pumps, headers, feeders and heated channels are arranged in a typical CANDU figure-of-eight geometry. The loop has an emergency coolant injection system (ECI) that may be operated in several modes, including typical features of the various ECI systems found in CANDU reactors. A series of experiments has been performed in RD-14 to investigate the thermal-hydraulic behaviour during the blowdown and injection phases of a loss-of-coolant accident (LOCA). The tests were designed to cover a full range of break sizes from feeder-sized breaks to guillotine breaks in either an inlet or an outlet header. Breaks resulting in channel flow stagnation were also investigated. This paper reviews the results of some of the LOCA tests carried out in RD-14, and discusses some of the behaviour observed. Plans for future experiments in a multiple-channel RD-14 facility, modified to contain multiple flow channels, are outlined. (orig.)

  6. Desalination of Impaired Water Using Geothermal Energy

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, Craig S [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Akar, Sertac [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Cath, Tzahi [Colorado School of Mines; Vanneste, Johan [Colorado School of Mines; Gustafson, Emily [Colorado School of Mines

    2017-10-04

    Membrane distillation (MD) and nanofiltration (NF) are explored as a means to provide high quality water for on-site use at the Tuscarora geothermal power plant in northern Nevada. The plant uses a wet cooling tower, but decreasing flow from the wells providing makeup water necessitates exploration for alternative water or alternative cooling sources. Scenarios are explored to extend cooling water by (1) extracting fresh water from the geothermal brine, (2) upgrading the makeup-water quality to allow for increased cycles of concentration in the cooling tower, or (3) recovering water from the cooling tower blowdown. The preliminary cost analysis indicates that applying NF to extract water from the injection brine is the most attractive option of the scenarios examined. This approach may be useful for other plants as well. The estimated cost for the NF treatment of the injection brine ranges from $0.63/m3 to $0.45/m3 and provides a reduction in the current makeup well flows of 35% to 71%. Savings from the reduction in makeup well pumping and chemical treatment do not fully offset the estimated cost of the proposed treatment systems; the site will have to weigh the cost of these water treatment options versus alternatives in light of the diminishing flows from the existing cooling-water wells. Testing is planned to quantify the performance of the proposed NF and MD technologies and help refine the estimated system costs.

  7. Assessment of diagnostic methods for determining degradation of motor-operated valves

    International Nuclear Information System (INIS)

    Haynes, H.D.; Farmer, W.S.

    1992-01-01

    The Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of motor-operated valves (MOVs) in support of the Nuclear Plant Aging Research (NPAR) program. This paper provides a summary of the ORNL MOV aging assessment with emphasis on the identification, evaluation, and application of MOV monitoring methods and techniques. The diagnostic information available from any MOV measurable parameters was evaluated by ORNL using MOVs that were mounted on test stands. Those tests led to the conclusion that the single most informative MOV measurable parameter was also the one which was most easily acquired, namely the motor current. Motor current signature analysis (MCSA) was found to provide detailed information related to the condition of the motor, motor operator, and valve across a wide range of levels. As part of the MOV aging assessment, several tests were carried out by ORNL on MOVs having implanted defects and degradations. Tests were also performed on many MOVs located within a nuclear power plant. In addition, ORNL participated in the Gate Valve Flow Interruption Blowdown Test program carried out at Wyle Laboratories in Huntsville, Alabama. Results from all of these tests are summarized in this paper and several selected examples are given. Other areas covered in this paper include descriptions of relevant regulatory issues and activities, other related diagnostics research at ORNL, and interactions ORNL has had with outside organizations for the purpose of disseminating research results

  8. Uncertainties in modeling and scaling in the prediction of fuel stored energy and thermal response

    International Nuclear Information System (INIS)

    Wulff, W.

    1987-01-01

    The steady-state temperature distribution and the stored energy in nuclear fuel elements are computed by analytical methods and used to rank, in the order of importance, the effects on stored energy from statistical uncertainties in modeling parameters, in boundary and in operating conditions. An integral technique is used to calculate the transient fuel temperature and to estimate the uncertainties in predicting the fuel thermal response and the peak clad temperature during a large-break loss of coolant accident. The uncertainty analysis presented here is an important part of evaluating the applicability, the uncertainties and the scaling capabilities of computer codes for nuclear reactor safety analyses. The methods employed in this analysis merit general attention because of their simplicity. It is shown that the blowdown peak is dominated by fuel stored energy alone or, equivalently, by linear heating rate. Gap conductance, peaking factors and fuel thermal conductivity are the three most important fuel modeling parameters affecting peak clad temperature uncertainty. 26 refs., 10 figs., 6 tabs

  9. Final environmental statement related to the operation of Callaway Plant, Unit No. 1 (Docket No. 50-483)

    International Nuclear Information System (INIS)

    1982-01-01

    The final environmental statement contains the second assessment of the environmental impact associated with operation of Callaway Plant Unit 1, pursuant to the National Environmental Policy Act of 1969 (NEPA) and 10 CFR Par 51, as amended, of the NRC's regulations. This statement examines: the purpose and need for the Callaway project, alternatives to the project, the affected environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs. No water-use impacts are expected from cooling-tower markup withdrawn from, or blowdown discharged into, the Missouri River. Land-use and terrestrial- and aquatic-ecological impacts will be small. Air-quality impacts from cooling-tower drift and other emissions and dust will also be small. Impacts to historic and prehistoric sites will be negligible with the development and implementation of the applicant's cultural-resources management plan. No significant impacts are anticipated from normal operational releases of radioactivity. The risk associated with accidental radiation exposure is very low. The net socioeconomic effects of the project will be beneficial. The action called for is the issuance of an operating license for Unit 1 of the Callaway Plant. 18 figs., 16 tabs

  10. Critical heat flux experimentation in an annular test section

    International Nuclear Information System (INIS)

    White, J.D.; Levin, A.E.

    1978-01-01

    Steady-state critical heat flux experiments have been performed in the Forced Convection Test Facility (FCTF), an annular test section containing a single electrically heated rod, for the purpose of testing the applicability of existing critical heat flux correlations. Good accuracy has been obtained using the MacBeth-Barnett critical heat flux correlation for annuli, corrected for the ''stepped cosine'' power profile of the heater. The equivalent diameter of the test section, based on the wetted perimeter, is 2.1 cm (0.83 in.); the heated-to-wetted-perimeter ratio is 0.252. The heated length of the heater rod is 366 cm (144 in.). Nominal pressures for the tests have ranged from 7.2 to 15.5 MN/m 2 (1044 to 2250 psia); coolant flow rates have been 0.32 dm 3 /sec (5 gpm), 0.63 dm 3 /sec (10 gpm), and 1.26 dm 3 /sec (20 gpm); and heater powers of 72 kW, 122 kW, and 144 kW have been used. Maximum error in prediction of first observed critical heat flux is 21 percent; rms error is 11.7 percent. Attempts have also been made to predict the occurrence of critical heat flux during blowdowns (depressurization transients) of the FCTF. The results of these predictions are inconclusive at this time

  11. Suppression Pool Mixing and Condensation Tests in PUMA Facility

    International Nuclear Information System (INIS)

    Ling Cheng; Kyoung Suk Woo; Mamoru Ishii; Jaehyok Lim; Han, James

    2006-01-01

    Condensation of steam with non-condensable in the form of jet flow or bubbly flow inside the suppression pool is an important phenomenon on determining the containment pressure of a passively safe boiling water reactor. 32 cases of pool mixing and condensation test have been performed in Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility under the sponsor of the U.S. Nuclear Regulatory Commission to investigate thermal stratification and pool mixing inside the suppression pool during the reactor blowdown period. The test boundary conditions, such as the steam flow rate, the noncondensable gas flow rate, the initial water temperature, the pool initial pressure and the vent opening submergence depth, which covers a wide range of prototype (SBWR-600) conditions during Loss of Coolant Accident (LOCA) were obtained from the RELAP5 calculation. The test results show that steam is quickly condensed at the exit of the vent opening. For pure steam injection or low noncondensable injection cases, only the portion above the vent opening in the suppression pool is heated up by buoyant plumes. The water below the vent opening can be heated up slowly through conduction. The test results also show that the degree of thermal stratification in suppression pool is affected by the vent opening submergence depth, the pool initial pressure and the steam injection rate. And it is slightly affected by the initial water temperature. From these tests it is concluded that the pool mixing is strongly affected by the noncondensable gas flow rate. (authors)

  12. Derecho Hazards in the United States.

    Science.gov (United States)

    Ashley, Walker S.; Mote, Thomas L.

    2005-11-01

    Convectively generated wind-storms occur over broad temporal and spatial scales; however, the more widespread and longer lived of these windstorms have been given the name "derecho." Utilizing an integrated derecho database, including 377 events from 1986 to 2003, this investigation reveals the amount of insured property losses, fatalities, and injuries associated with these windstorms in the United States. Individual derechos have been responsible for up to 8 fatalities, 204 injuries, forest blow-downs affecting over 3,000 km2 of timber, and estimated insured losses of nearly a $500 million. Findings illustrate that derecho fatalities occur more frequently in vehicles or while boating, while injuries are more likely to happen in vehicles or mobile homes. Both fatalities and injuries are most common outside the region with the highest derecho frequency. An underlying synthesis of both physical and social vulnerabilities is suggested as the cause of the unexpected casualty distribution. In addition, casualty statistics and damage estimates from hurricanes and tornadoes are contrasted with those from derechos to emphasize that derechos can be as hazardous as many tornadoes and hurricanes.

  13. The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

    Directory of Open Access Journals (Sweden)

    Jang Hyung-Wook

    2017-01-01

    Full Text Available The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during loss of coolant accident analysis. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. From original LBLOCA input deck file, the nodalization of downcomer and junction connections with 4 cold legs and direct vessel injection lines are modified for reflecting the realistic cross-flow effect and real downcomer structure. The analysis results show that the peak cladding temperature of new input deck decreases more rapidly than previous result and that the drop of peak cladding temperature was advanced by application of momentum flux term in cross-flow. Additionally, the authors developed a new input deck with multi-dimensional downcomer model and ran MARS code with multi-dimensional input deck as well. By using the modified input deck, the Emergency core cooling system by-pass flow phenomena is better characterized and found to be consistent with both experimental report and regulatory guide.

  14. Effluent testing for the Oak Ridge mixed waste incinerator: Emissions test for August 27, 1990

    International Nuclear Information System (INIS)

    Bostick, W.D.; Bunch, D.H.; Gibson, L.V.; Hoffmann, D.P.; Shoemaker, J.L.

    1990-12-01

    On August 27, 1990, a special emissions test was performed at the K-1435 Toxic Substance Control Act Mixed Waste Incinerator. A sampling and analysis plan was implemented to characterize the incinerator waste streams during a 6 hour burn of actual mixed waste. The results of this characterization are summarized in the present report. Significant among the findings is the observation that less than 3% of the uranium fed to the incinerator kiln was discharged as stack emission. This value is consistent with the estimate of 4% or less derived from long-term mass balance of previous operating experience and with the value assumed in the original Environmental Impact Statement. Approximately 1.4% of the total uranium fed to the incinerator kiln appeared in the aqueous scrubber blowdown; about 85% of the total uranium in the aqueous waste was insoluble (i.e., removable by filtration). The majority of the uranium fed to the incinerator kiln appeared in the ash material, apparently associated with phosphorous as a sparingly-soluble species. Many other metals of potential regulatory concern also appeared to concentrate in the ash as sparingly-soluble species, with minimal partition to the aqueous waste. The aqueous waste was discharged to the Central Neutralization Facility where it was effectively treated by coprecipitation with iron. The treated, filtered aqueous effluent met Environmental Protection Agency interim primary drinking water standards for regulated metals

  15. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200% double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding temperature excursion was terminated at 175 s when complete core quenching occurred. Accumulator flows were initiated at 10 s when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching

  16. Hydrogen Mixing Studies (HMS) assessment manual

    International Nuclear Information System (INIS)

    Lam, K.L.; Wilson, T.L.; Travis, J.R.

    1993-06-01

    This report documents some calculations performed to assess the Hydrogen Mixing Studies (HMS) code. Results are presented first for some analytical test problems, including laminar flow and mass diffusion. The von Karman vortex street problem and the Sandia FLAME Facility and Heiss Dampf Reaktor (HDR) containment facility test problems are then discussed. For the analytical problems, the code gave results that agree exceptionally well with the analytical solutions. Calculations for the von Karman vortex street problem were performed at selected Reynolds numbers for several obstacle types. The computed flow patterns agree well with experimental observations-specifically the occurrence of a vortex street (double row of vortices) above a critical Reynolds number. Calculations for the von Karman vortex street problem were performed at selected Reynolds numbers for several obstacle types. The computed flow patterns agree well with experimental observations-specifically the occurrence of a vortex street (double row of vortices) above a critical Reynolds number. The last assessment problem involves modeling the experiment T31.5. The experiment was carried out in the HDR containment building, which is a large, multi-compartment facility (11 300 m 3 free volume in 72 compartments). In the experiment, a steam-water mixture was first injected into the containment to simulate a large-break blowdown of a pressure vessel, and then superheated steam was injected that was followed by a release of helium-hydrogen light gas. The calculated results (pressure, temperature, and gas concentrations) agree reasonably well with the experimental data

  17. Application of KIMERA Methodology to Kori 3 and 4 LBLOCA M/E Release Analysis

    International Nuclear Information System (INIS)

    Song, Jeung Hyo; Hwang, Byung Heon; Kim, Cheol Woo

    2007-01-01

    A new mass and energy (M/E) release analysis methodology called KIMERA (KOPEC Improved Mass and Energy Release Analysis) has been developed. This is a realistic evaluation methodology of the M/E release analysis for the containment design and is applicable to a LOCA and a main steam line break (MSLB) accident. This KIMERA methodology has the same engine as KREM (KEPRI Realistic Evaluation Model) which is the realistic evaluation methodology for LOCA peak clad temperature analysis. This methodology also has several supplementary conservative models for the M/E release such as break spillage model and multiplier on heat transfer coefficient (HTC). For estimating the applicability of the KIMERA methodology to the licensing analysis, the large break LOCA (LBLOCA) M/E analysis was performed for UCN 3 and 4 which is the typical plant of OPR1000 type. The results showed that the peak pressure and temperature occurred earlier and had lower values than those of UCN 3 and 4 FSAR. The KIMERA methodology takes off the over-conservatism from the FSAR results during the post blowdown period for the large break LOCA and provides more margin in containment design. In this study, the LBLOCA M/E analysis using the KIMERA methodology is to be performed for Kori 3 and 4 which is the typical plant of Westinghouse type. The results are compared with those of the Kori Nuclear Unit 3 and 4 FSAR

  18. Homogeneous non-equilibrium two-phase critical flow model

    International Nuclear Information System (INIS)

    Schroeder, J.J.; Vuxuan, N.

    1987-01-01

    An important aspect of nuclear and chemical reactor safety is the ability to predict the maximum or critical mass flow rate from a break or leak in a pipe system. At the beginning of such a blowdown, if the stagnation condition of the fluid is subcooled or slightly saturated thermodynamic non-equilibrium exists in the downstream, e.g. the fluid becomes superheated to a degree determined by the liquid pressure. A simplified non-equilibrium model, explained in this report, is valid for rapidly decreasing pressure along the flow path. It presumes that fluid has to be superheated by an amount governed by physical principles before it starts to flash into steam. The flow is assumed to be homogeneous, i.e. the steam and liquid velocities are equal. An adiabatic flow calculation mode (Fanno lines) is employed to evaluate the critical flow rate for long pipes. The model is found to satisfactorily describe critical flow tests. Good agreement is obtained with the large scale Marviken tests as well as with small scale experiments. (orig.)

  19. Landscape analysis and pattern of hurricane impact and circulation on mangrove forests of the everglades

    Science.gov (United States)

    Doyle, T.W.; Krauss, K.W.; Wells, C.J.

    2009-01-01

    The Everglades ecosystem contains the largest contiguous tract of mangrove forest outside the tropics that were also coincidentally intersected by a major Category 5 hurricane. Airborne videography was flown to capture the landscape pattern and process of forest damage in relation to storm trajectory and circulation. Two aerial video transects, representing different topographic positions, were used to quantify forest damage from video frame analysis in relation to prevailing wind force, treefall direction, and forest height. A hurricane simulation model was applied to reconstruct wind fields corresponding to the ground location of each video frame and to correlate observed treefall and destruction patterns with wind speed and direction. Mangrove forests within the storm's eyepath and in the right-side (forewind) quadrants suffered whole or partial blowdowns, while left-side (backwind) sites south of the eyewall zone incurred moderate canopy reduction and defoliation. Sites along the coastal transect sustained substantially more storm damage than sites along the inland transect which may be attributed to differences in stand exposure and/or stature. Observed treefall directions were shown to be non-random and associated with hurricane trajectory and simulated forewind azimuths. Wide-area sampling using airborne videography provided an efficient adjunct to limited ground observations and improved our spatial understanding of how hurricanes imprint landscape-scale patterns of disturbance. ?? 2009 The Society of Wetland Scientists.

  20. Gentilly 2 steam generators Spring 2000 outage: tubesheet waterlance cleaning and inspection; upper bundle inspection

    International Nuclear Information System (INIS)

    Akeroyd, J.K.; Plante, S.

    2000-01-01

    A review of the secondary side maintenance activities recently completed during the Gentilly 2 Annual Spring 2000 Maintenance Outage. Activities included: 1) Tubesheet intertube waterlance cleaning and visual inspection, 2) First tube support plate, in-bundle visual inspection of the hot leg, and 3) Upper bundle tube support plate visual inspection. A description of the waterlancing and inspection equipment and setup in the RB at Gentilly 2 is provided. Several innovative techniques were successfully employed and yielded savings in critical path duration, labour and personnel radiation dose. These included accessing the SG tubesheet region through one handhole only and sludge removal utilizing the SG blowdown system. Plant personnel judged tubesheet sludge removal successful. Before and after results of the cleaning process along with samples of the visual inspection results are provided. Inspection of the first support plate, which was a repeat of an inspection done in 1997, was conducted along with an in-bundle inspection of the upper tube supports. Results are presented along with a discussion of the implications for future steam generator maintenance. (author)

  1. Risk from a pressurized toxic gas system: Part 2, Dispersal consequences

    International Nuclear Information System (INIS)

    Brereton, S.J.; Martin, D.; Lane, S.G.; Altenbach, T.J.

    1995-04-01

    During the preparation of a Safety Analysis Report at the Lawrence Livermore National Laboratory, we studied the release of chlorine from a compressed gas experimental apparatus. This paper presents the second part in a series of two papers on this topic. The first paper focuses on the frequency of an unmitigated release from the system; this paper discusses the consequences of the release. The release of chlorine from the experimental apparatus was modeled as an unmitigated blowdown through a 0.25 inch (0.0064 m) outside diameter tube. The physical properties of chlorine were considered as were the dynamics of the fluid flow problem. The calculated release rate was used as input for the consequence assessment. Downwind concentrations as a function of time were evaluated and then compared to suggested guidelines. For comparison purposes, a typical water treatment plant was briefly studied. The lower hazard presented by the LLNL operation becomes evident when its release is compared to the release of material from a water treatment plant, a hazard which is generally accepted by the public

  2. Assessment of Effect on LBLOCA PCT for Change in Upper Head Nodalization

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Huh, Byung Gil; Yoo, Seung Hun; Bang, Youngseok; Seul, Kwangwon; Cho, Daehyung

    2014-01-01

    In this study, the best estimate plus uncertainty (BEPU) analysis of LBLOCA for original and modified nodalizations was performed, and the effect on LBLOCA PCT for change in upper head nodalization was assessed. In this study, the best estimate plus uncertainty (BEPU) analysis of LBLOCA for original and modified nodalizations was performed, and the effect on LBLOCA PCT for change in upper head nodalization was assessed. It is confirmed that modification of upper head nodalization influences PCT behavior, especially in the reflood phase. In conclusions, the modification of nodalization to reflect design characteristic of upper head temperature should be done to predict PCT behavior accurately in LBLOCA analysis. In the best estimate (BE) method with the uncertainty evaluation, the system nodalization is determined by the comparative studies of the experimental data. Up to now, it was assumed that the temperature of the upper dome in OPR-1000 was close to that of the cold leg. However, it was found that the temperature of the upper head/dome might be a little lower than or similar to that of the hot leg through the evaluation of the detailed design data. Since the higher upper head temperature affects blowdown quenching and peak cladding temperature in the reflood phase, the nodalization for upper head should be modified

  3. FLOOD 3 code conversion from Apollo to HP

    International Nuclear Information System (INIS)

    Lee, Hae Cho

    1996-01-01

    FLOOD3 is a Fortran program used to analyze LOCA Reflood and Post Reflood transients for purpose of mechanistically generating containment sizing mass/energy source terms. The reflood time frame starts when safety injection water first enters the bottom of the core, following the initial LOCA blowdown phase, and the ends when liquid level in the core is high enough to quench the core. The post reflood phase starts at the end of reflood phase and lasts until the end of LOCA transient period. Longterm core cooling is initiated following post reflood phase. FLOOD3 code dose not contain separate capability for longterm cooling analysis. This report firstly describes detailed work carried out for installation of FLOOD3 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of FLOOD3 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

  4. LOFT/LP-02-6, Loss of Fluid Test, 1. OECD Large Break Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The fourth OECD LOFT experiment was conducted on 3 October 1983. This was the first OECD LOFT large break experiment. The initial and boundary conditions were chosen to be representative of USNRC licensing limits for commercial PWRs. This included loss of off-site power coincident with LOCA initiation. This experiment included the first use in LOFT of pressurized fuel rods in the center bundle. The experiment was initiated by opening the quick-opening blow-down valves in the broken hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  5. LOFT/LP-FP-1B, Loss of Fluid Test, Fission Product Release Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The seventh OECD LOFT experiment was conducted on 19 December 1984. It was the first of the two experiments to be performed in the LOFT facility with intentional release of fission products. Its objectives were to obtain data on fission product release from the fuel-cladding gap into vapor and reflood water and to collect data on transport of these fission products through and out of the reactor coolant system. The experiment was initiated by a reactor scram with one second delayed opening of the quick-opening blowdown valves. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  6. Latest design of gate valves

    Energy Technology Data Exchange (ETDEWEB)

    Kurzhofer, U.; Stolte, J.; Weyand, M.

    1996-12-01

    Babcock Sempell, one of the most important valve manufacturers in Europe, has delivered valves for the nuclear power industry since the beginning of the peaceful application of nuclear power in the 1960s. The latest innovation by Babcock Sempell is a gate valve that meets all recent technical requirements of the nuclear power technology. At the moment in the United States, Germany, Sweden, and many other countries, motor-operated gate and globe valves are judged very critically. Besides the absolute control of the so-called {open_quotes}trip failure,{close_quotes} the integrity of all valve parts submitted to operational forces must be maintained. In case of failure of the limit and torque switches, all valve designs have been tested with respect to the quality of guidance of the gate. The guidances (i.e., guides) shall avoid a tilting of the gate during the closing procedure. The gate valve newly designed by Babcock Sempell fulfills all these characteristic criteria. In addition, the valve has cobalt-free seat hardfacing, the suitability of which has been proven by friction tests as well as full-scale blowdown tests at the GAP of Siemens in Karlstein, West Germany. Babcock Sempell was to deliver more than 30 gate valves of this type for 5 Swedish nuclear power stations by autumn 1995. In the presentation, the author will report on the testing performed, qualifications, and sizing criteria which led to the new technical design.

  7. An exploratory investigation of a wake disruption technique for studying wake reestablishment time

    Science.gov (United States)

    Clark, L. E.; Jones, R. A.

    1974-01-01

    An exploratory investigation was made of a wake disruption technique for studying the hypersonic-wake reestablishment time in a blowdown wind tunnel. In this technique, a highly underexpanded jet issuing from the base of a 10 deg half-angle cone totally disrupts and displaces the conventional wake. The jet was rapidly shut off by an explosively actuated valve and the time for wake reestablishment was measured. The tests were conducted in the Mach 6 high Reynolds number tunnel at a stagnation temperature of 506 K and stagnation pressure of 2.86 MPa. The model base jet stagnation pressure was 3.55 MPa at room temperature. High-speed schlieren motion pictures indicated that disappearance of the disrupting jet and reestablishment of the wake-recompression shock were probably occurring simultaneously and that the time disruptive-jet-air shutoff to wake recompression shock reestablishment was probably between 200 and 450 microseconds (flow lengths from 1.8 to 4.2). The values of flow lengths are about one-thord to one-half the values measured in impulse facilities in a previous study. This shorter time is believed to be largely due to difference in flow conditions between the jet disruption technique and impulse facilities.

  8. ITER Port Interspace Pressure Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan J [ORNL; Van Hove, Walter A [ORNL

    2016-01-01

    The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

  9. An emergency water injection system (EWIS) for future CANDU reactors

    International Nuclear Information System (INIS)

    Marques, Andre L.F.; Todreas, Neil E.; Driscoll, Michael J.

    2000-01-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m 2 : the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  10. Water Hammer Mitigation on Postulated Pipe Break of Feed Water System

    International Nuclear Information System (INIS)

    Seong, Ho Je; Woo, Kab Koo; Cho, Keon Taek

    2008-01-01

    The Feed Water (FW) system supplies feedwater from the deaerator storage tank to the Steam Generators(S/G) at the required pressure, temperature, flow rate, and water chemistry. The part of FW system, from the S/G to Main Steam Valve House just outside the containment building wall, is designed as safety grade because of its safety function. According to design code the safety related system shall be designed to protect against dynamic effects that may results from a pipe break on high energy lines such as FW system. And the FW system should be designed to minimize blowdown volume of S/G secondary side during the postulated pipe break. Also the FW system should be designed to prevent the initiation or to minimize the effects of water hammer transients which may be induced by the pipe break. This paper shows the results of the hydrodynamic loads induced by the pipe break and the optimized design parameters to mitigate water hammer loads of FW system for Shin-Kori Nuclear Power Plant Unit 3 and 4 (SKN 3 and 4)

  11. Implementation of Aerosol Transport and Resuspension Models into the GAMMA+ Code for the Safety Analysis of a VHTR

    International Nuclear Information System (INIS)

    Yoon, Churl; Tak, Nam Il; Lim, Hong Sik

    2010-01-01

    One of the unique features of a Very High Temperature Gas Cooled Reactor (VHTR) is Vented Low Pressure Containment (VLPC) containing two separate vent paths where both have two gravity operated relief valves in a series. Because VLPC strategy allows the release of a relatively small amount of radioactive fission products(FP) into the environment during the blowdown phase, behavior analyses of the fission products circulating in the primary coolant loop and in the containment are major consideration factors for safety evaluation. For thermal-fluid analysis of a Very High Temperature Gas Cooled Reactor (VHTR), the GAMMA(GAs Multicomponent Mixture Analysis)+ code is under development. The MAEROS model is the multicomponent aerosol module of the CONTAIN code, and has been widely used for aerosol behavior analysis. For the first work of FP module development, the MAEROS model had been implemented as an independent module and examined against some analytic solutions and experimental data by Yoo et al. In this study, an aerosol transport model and a turbulent resuspension model were additionally implemented in the FP module of the GAMMA+ code and verified for FP analysis of a VHTR

  12. Fluid-structure interaction in BWR suppression pool systems. Final report

    International Nuclear Information System (INIS)

    Nickell, R.E.

    1979-09-01

    The discharge of safety relief valves or a severe loss-of-coolant event in a boiling-water-cooled reactor steam supply system triggers a complex pressure suppression system that is based upon sub-surface steam condensation in large pools of water. The physical problems fall into two categories. The first is referred to as vent clearing and describes the process of expelling non-condensables from the system prior to steam flow. The second category covers a variety of phenomena related to the transient overexpansion of a condensable volume and the subsequent inertially-driven volume decrease. The dynamic loading of either event, depending upon fluid-structural design parameters, can be of concern in safety analysis. This report describes the development of a method for calculating the loads and the structural response for both types of problems. The method is embedded in a computer code, called PELE-IC, that couples a two-dimensional, incompressible eulerian fluid algorithm to a finite element shell algorithm. The fluid physics is based upon the SOLA algorithm, which provideds a trial velocity field using the Navier-Stokes equations that is subsequently corrected iteratively so that incompressibility, fluid-structure interface compatibility, and boundary conditions are satisfied. These fluid and fluid-structure algorithms have been extensively verified through calculations of known solutions from the classical literature, and by comparison to air and steam blowdown experiments

  13. Minimization of radioactive liquids released from PWR

    International Nuclear Information System (INIS)

    Yoshikawa, Hideo; Kohri, Masaharu.

    1981-01-01

    The quantity of radioactive substances in the liquids released into the environment from a PWR power station in normal operation was determined, following the path from the sources of generation, that is, the equipments in primary and secondary cooling systems, to the release into the environment after the radioactive substances were removed in treatment facilities. The quantity of radioactive substances released from primary and secondary systems was determined for each source of generation in a standard plant, and the results were examined. As the concrete example of reducing the release on the basis of ''As low as reasonably achievable'' concept, the increase of letdown flow rate and the installation of a condensate-desalting column are reported. As the sources of generation, the primary coolant formed by shim bleed and the drain from primary system equipments, the drain from an auxiliary building floor, radiochemistry waste solution and the drain from intermediate cooling system, the waste water of washing and shower bath, the drain from a turbine building floor, and the blow-down waste from steam generators are enumerated. The concentration of radioactive substances in primary and secondary coolants, the decontamination factor of waste treatment equipments and the measures for reducing the release are described. (Kako, I.)

  14. The stringy instanton partition function

    Energy Technology Data Exchange (ETDEWEB)

    Bonelli, Giulio [International School of Advanced Studies (SISSA),via Bonomea 265, 34136 Trieste (Italy); INFN, Sezione di Trieste,Trieste (Italy); I.C.T.P.,Strada Costiera 11, 34014 Trieste (Italy); Sciarappa, Antonio; Tanzini, Alessandro; Vasko, Petr [International School of Advanced Studies (SISSA),via Bonomea 265, 34136 Trieste (Italy); INFN, Sezione di Trieste,Trieste (Italy)

    2014-01-09

    We perform an exact computation of the gauged linear sigma model associated to a D1-D5 brane system on a resolved A{sub 1} singularity. This is accomplished via supersymmetric localization on the blown-up two-sphere. We show that in the blow-down limit ℂ{sup 2}/ℤ{sub 2} the partition function reduces to the Nekrasov partition function evaluating the equivariant volume of the instanton moduli space. For finite radius we obtain a tower of world-sheet instanton corrections, that we identify with the equivariant Gromov-Witten invariants of the ADHM moduli space. We show that these corrections can be encoded in a deformation of the Seiberg-Witten prepotential. From the mathematical viewpoint, the D1-D5 system under study displays a twofold nature: the D1-branes viewpoint captures the equivariant quantum cohomology of the ADHM instanton moduli space in the Givental formalism, and the D5-branes viewpoint is related to higher rank equivariant Donaldson-Thomas invariants of ℙ{sup 1}×ℂ{sup 2}.

  15. SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR

    International Nuclear Information System (INIS)

    Halsall, M.J.; Course, A.F.; Sidell, J.

    1979-09-01

    SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)

  16. UPTF experiment: Effect of full-scale geometry on countercurrent flow behaviour in PWR downcomer

    International Nuclear Information System (INIS)

    Liebert, J.; Weiss, P.

    1989-01-01

    Four separate effects tests (13 runs) have been performed at UPTF - a 1:1 scale test facility - to investigate the thermal-hydraulic phenomena in the full-scale downcomer of a PWR during end-of-blowdown, refill and reflood phases. Special attention has been paid to the effects of geometry - cold leg arrangement - and ECC-water subcooling on downcomer countercurrent flow and ECC bypass behaviour. A synopsis of the most significant events and a comparison of countercurrent flow limitation (CCFL) data from UPTF and 1/5 scale test facility of Creare are given. The CCFL results of UPTF are compared to data predicted by an empirical correlation developed at Creare, based on the modified dimensionless Wallis parameter J * . A significant effect of cold leg arrangement on CCFL was observed leading to strongly heterogeneous flow condition in the downcomer. CCFL in front of cold leg 1 adjacent to the broken loop exists even for very low steam flow rates. Therefore the benefit of strong water subcooling is not as much as expected. The existing flooding correlation of Creare predicts the full-scale downcomer CCFL insufficiently. New flooding correlations are required to describe the CCFL process adequately. (orig.)

  17. Using Crossflow for Flow Measurements and Flow Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gurevich, A.; Chudnovsky, L.; Lopeza, A. [Advanced Measurement and Analysis Group Inc., Ontario (Canada); Park, M. H. [Sungjin Nuclear Engineering Co., Ltd., Gyeongju (Korea, Republic of)

    2016-10-15

    Ultrasonic Cross Correlation Flow Measurements are based on a flow measurement method that is based on measuring the transport time of turbulent structures. The cross correlation flow meter CROSSFLOW is designed and manufactured by Advanced Measurement and Analysis Group Inc. (AMAG), and is used around the world for various flow measurements. Particularly, CROSSFLOW has been used for boiler feedwater flow measurements, including Measurement Uncertainty Recovery (MUR) reactor power uprate in 14 nuclear reactors in the United States and in Europe. More than 100 CROSSFLOW transducers are currently installed in CANDU reactors around the world, including Wolsung NPP in Korea, for flow verification in ShutDown System (SDS) channels. Other CROSSFLOW applications include reactor coolant gross flow measurements, reactor channel flow measurements in all channels in CANDU reactors, boiler blowdown flow measurement, and service water flow measurement. Cross correlation flow measurement is a robust ultrasonic flow measurement tool used in nuclear power plants around the world for various applications. Mathematical modeling of the CROSSFLOW agrees well with laboratory test results and can be used as a tool in determining the effect of flow conditions on CROSSFLOW output and on designing and optimizing laboratory testing, in order to ensure traceability of field flow measurements to laboratory testing within desirable uncertainty.

  18. Final environmental statement related to construction of Skagit Nuclear Power Project Units 1 and 2: (Docket Nos. 50-522 and 50-523)

    International Nuclear Information System (INIS)

    1975-05-01

    The proposed action is the issuance of construction permits to the Pudget Sound Power and Light Company, Pacific Power and Light Company, Washington Water Power Company and the Washington Public Power Supply System, for the construction of Skagit Nuclear Power Projects Units 1 and 2 (Docket Nos. 50-522 and 50-523) in Skagit County, Washington (about 64 miles north of Seattle and 6 miles ENE of Sedro Woolley). These units are scheduled for commercial service in 1982 and 1985, respectively. Each unit will employ a boiling-water nuclear reactor with a maximum expected thermal power level of 4100 MWt, which is considered in the assessments contained in this statement. At the 3800 MWt power level initially to be licensed, the net electrical capacity of each unit will be 1288 MWe. The exhaust steam from the turbine-generators will be cooled in condensers which will utilize one hyperbolic-type natural-draft cooling tower per unit to dissipate heat to the atmosphere. Water (106 cfs max.) for the cooling tower makeup (82.4 cfs) and other plant uses will be withdrawn from the Skagit River through Ranney Collectors embedded in the north bank of the river and pumped to the plant through a pipeline about 35,000 ft. long. Cooling tower blowdown (7 cfs max.) from the project and dilution water (20 cfs max.) will flow through a pipeline back to the river where it will be discharged through a diffuser

  19. Dynamic loads caused by pressure blasts, steam explosions, and earth quakes; Dynamische Belastungen durch Druckstoesse, Dampfexplosionen und Erdbeben

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, H.H. [SDK Ingenieurunternehmen GmbH, Basel (Switzerland)

    1998-11-01

    The paper deals with description of structures and the relevant dynamic loads. As to the structures, gas, fluid, or solid structures are to be considered. They determine the characteristic vibrational behaviour of the structures in the interconnected system. The excitation type determines the component that will be induced to change characteristic vibrational behaviour of the structure, depending on the load increasing time and the period of excitation. Three examples are given to illustrate the processes. (Water tank subject to quasi-seismic conditions; pipeline affected by blow-down; shut-off valve for a pipe). (orig./CB) [Deutsch] In diesem Beitrag soll auf die Erfassung der Strukturen und die Erfassung der dynamischen Belastungen eingegangen werden. Zur Erfassung der Strukturen sind `Gas-, Fluid- und Festkoerper-Strukturen` zu beachten. Sie bestimmen das Eigenschwingverhalten im Verbund. Die Erregung bestimmt nun, welcher Bereich aus dem Eigen-Schwingverhalten der Struktur ueber die Lastanstiegs-Zeit und die Zeitdauer der Erregung anregbar ist. Drei Beispiele sollen die Aufgabenstellung erlaeutern (Wasserbehaelter unter erdbebenaehnlichen Bedingungen; Rohrleitung unter `Blow-down-Belastung`; Absperrklappe fuer eine Rohrleitung). (orig./MM)

  20. Analysis of combustion performance and emission of extended expansion cycle and iEGR for low heat rejection turbocharged direct injection diesel engines

    Directory of Open Access Journals (Sweden)

    Shabir Mohd F.

    2014-01-01

    Full Text Available Increasing thermal efficiency in diesel engines through low heat rejection concept is a feasible technique. In LHR engines the high heat evolution is achieved by insulating the combustion chamber surfaces and coolant side of the cylinder with partially stabilized zirconia of 0.5 mm thickness and the effective utilization of this heat depend on the engine design and operating conditions. To make the LHR engines more suitable for automobile and stationary applications, the extended expansion was introduced by modifying the inlet cam for late closing of intake valve through Miller’s cycle for extended expansion. Through the extended expansion concept the actual work done increases, exhaust blow-down loss reduced and the thermal efficiency of the LHR engine is improved. In LHR engines, the formation of nitric oxide is more, to reduce the nitric oxide emission, the internal EGR is incorporated using modified exhaust cam with secondary lobe. Modifications of gas exchange with internal EGR resulted in decrease in nitric oxide emissions. In this work, the parametric studies were carried out both theoretically and experimentally. The combustion, performance and emission parameters were studied and were found to be satisfactory.

  1. Dynamics of Learning in MLP: Natural Gradient and Singularity Revisited.

    Science.gov (United States)

    Amari, Shun-Ichi; Ozeki, Tomoko; Karakida, Ryo; Yoshida, Yuki; Okada, Masato

    2018-01-01

    The dynamics of supervised learning play a main role in deep learning, which takes place in the parameter space of a multilayer perceptron (MLP). We review the history of supervised stochastic gradient learning, focusing on its singular structure and natural gradient. The parameter space includes singular regions in which parameters are not identifiable. One of our results is a full exploration of the dynamical behaviors of stochastic gradient learning in an elementary singular network. The bad news is its pathological nature, in which part of the singular region becomes an attractor and another part a repulser at the same time, forming a Milnor attractor. A learning trajectory is attracted by the attractor region, staying in it for a long time, before it escapes the singular region through the repulser region. This is typical of plateau phenomena in learning. We demonstrate the strange topology of a singular region by introducing blow-down coordinates, which are useful for analyzing the natural gradient dynamics. We confirm that the natural gradient dynamics are free of critical slowdown. The second main result is the good news: the interactions of elementary singular networks eliminate the attractor part and the Milnor-type attractors disappear. This explains why large-scale networks do not suffer from serious critical slowdowns due to singularities. We finally show that the unit-wise natural gradient is effective for learning in spite of its low computational cost.

  2. Reactor Safety Research Programs Quarterly Report April -June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1980-11-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  3. On the failure modes of alternative containment designs following postulated core meltdown

    International Nuclear Information System (INIS)

    Chan, C.K.; Knee, H.E.; Okrent, D.

    1977-01-01

    The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR respresentative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best esimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded. (Auth.)

  4. LBB in Candu plants

    Energy Technology Data Exchange (ETDEWEB)

    Kozluk, M.J.; Vijay, D.K. [Ontario Hydro Nuclear, Toronto, Ontario (Canada)

    1997-04-01

    Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption that resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.

  5. Environmental study of nylon flocking process.

    Science.gov (United States)

    Burkhart, J; Piacitelli, C; Schwegler-Berry, D; Jones, W

    1999-05-14

    Environmental measurements for a variety of gas, particulate, and microbiological agents have been made in order to characterize exposures associated with the nylon flocking process. Of all agents measured, particulate is the predominant exposure. Levels of total particulate ranged from O.1 to 240 mg/m3 (x = 11.4 mg/m3). Average respirable particulate was 2.2 mg/m3, ranging from 0.5 to 39.9 mg/m3. Highest levels of particulates were found in the flocking room, and direct reading dust measurements indicate that the highest peak exposures are associated with "blowdown" (a cleaning procedure used between flocking runs). The nature of the airborne particles was investigated using polarized light and scanning electron microscopy. Air samples were found to contain flock particles (fibers nominally 10-15 microm in diameter by about 1000 microm in length) and a variety of respirable particles types, several of which were linked directly to the process. Of special interest were elongated respirable particles, which by microscopic analysis, complemented with melting-point determination, were found to be shreds of nylon.

  6. On the Development of Multi-Dimensional RELAP5 with Conservative Convective Terms

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hyung Wook; Lee, Sang Yong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The results of the 2D/3D experiments were summarized as follows; Flow conditions in the downcomer during end-of blowdown were highly multi-dimensional at full-scale. During reflood, the distribution of water in the core was one-dimensional. But flow in the core exhibited multidimensionality. One-dimensional manometer oscillation between the downcomer and core was observed. The water level was higher in front of the broken cold leg nozzle than at other azimuthal positions. Flow phenomena at the tie plate were uniform. The most famous code RELAP5 is still one dimensional even though it has been applied to various licensing applications. Therefore, author developed the multi-dimensional capability and implemented it into RELAP5. In this paper, two aspects concerning the multidimensional codes will be discussed. One of them is the properness of the type of the momentum equations. The other discussion will be the implementation of the conservative momentum flux term in RELAP5.

  7. Experiment data report for LOFT nonnuclear Test L1-4

    International Nuclear Information System (INIS)

    Batt, D.L.

    1977-07-01

    Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279 0 C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response

  8. Limit the effects of secondary circuit water or steam piping breaks in the reactor building

    International Nuclear Information System (INIS)

    Nachev, N.

    2001-01-01

    The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

  9. A fuzzy approach for modelling radionuclide in lake system.

    Science.gov (United States)

    Desai, H K; Christian, R A; Banerjee, J; Patra, A K

    2013-10-01

    Radioactive liquid waste is generated during operation and maintenance of Pressurised Heavy Water Reactors (PHWRs). Generally low level liquid waste is diluted and then discharged into the near by water-body through blowdown water discharge line as per the standard waste management practice. The effluents from nuclear installations are treated adequately and then released in a controlled manner under strict compliance of discharge criteria. An attempt was made to predict the concentration of (3)H released from Kakrapar Atomic Power Station at Ratania Regulator, about 2.5 km away from the discharge point, where human exposure is expected. Scarcity of data and complex geometry of the lake prompted the use of Heuristic approach. Under this condition, Fuzzy rule based approach was adopted to develop a model, which could predict (3)H concentration at Ratania Regulator. Three hundred data were generated for developing the fuzzy rules, in which input parameters were water flow from lake and (3)H concentration at discharge point. The Output was (3)H concentration at Ratania Regulator. These data points were generated by multiple regression analysis of the original data. Again by using same methodology hundred data were generated for the validation of the model, which were compared against the predicted output generated by using Fuzzy Rule based approach. Root Mean Square Error of the model came out to be 1.95, which showed good agreement by Fuzzy model of natural ecosystem. Copyright © 2013 Elsevier Ltd. All rights reserved.

  10. Effects of viscoelasticity on cavitation in drag reducing fluids

    Science.gov (United States)

    Ting, R. Y.

    1974-01-01

    To study cavitation inception in polymer solutions, a blow-down water tunnel with short running times was used. Tests were made using 1/4 and 1/2 inch diameter models of hemispherical-nose cylinders. To accurately detect the inception of cavitation, a reliable technique was developed using a continuously operating He-Ne gas laser. The laser beam was adjusted to grazing incidence with the model at the minimum pressure point where cavitation inception was to be expected. A sensitive photocell was placed at ninety degrees to detect the beam. As incipient cavitation occurred, the bubbles caused scattering of the laser beam which was picked up by the photocell. Static pressure near the model in the working section of the tunnel was measured using a solid-state pressure pick-up. The signals from the photocell and the pressure pick-up were recorded on an oscillograph. Velocity field visualization was achieved using one microsecond duration light pulses scattered by small polystryrene latex spheres in the flow.

  11. Probabilistic consequence assessment of hydrogen sulphide releases from a heavy water plant

    International Nuclear Information System (INIS)

    1983-01-01

    This report is concerned with the evaluation of the consequences to the public of an accidental release of hydrogen sulphide (H 2 S) to the atmosphere following a pipe or pressure envelope failure, or some other process upset, at a heavy water plant. It covers the first stage of a programme in which the nature of the problem was analyzed and recommendations made for the implementation of a computer model. The concepts of risk assessment and consequence assessment are discussed and a methodology proposed for combining the various elements of the problem into an overall consequence model. These elements are identified as the 'Initiating Events', 'Route to Receptor' and 'Receptor Response' and each is studied in detail in the report. Such phenomena as the blowdown of H 2 S from a rupture, the initial gas cloud behaviour, atmospheric dispersion and the toxicity of H 2 S and sulphur dioxide (SO 2 ) are addressed. Critical factors are identified and modelling requirements specified, with special reference to the Bruce heavy water plant. Finally, an overall model is recommended for implementation at the next stage of the programme, together with detailed terms of reference for the remaining work

  12. Biocide usage in cooling towers in the electric power and petroleum refining industries

    Energy Technology Data Exchange (ETDEWEB)

    Veil, J.; Rice, J.K.; Raivel, M.E.S.

    1997-11-01

    Cooling towers users frequently apply biocides to the circulating cooling water to control growth of microorganisms, algae, and macroorganisms. Because of the toxic properties of biocides, there is a potential for the regulatory controls on their use and discharge to become increasingly more stringent. This report examines the types of biocides used in cooling towers by companies in the electric power and petroleum refining industries, and the experiences those companies have had in dealing with agencies that regulate cooling tower blowdown discharges. Results from a sample of 67 electric power plants indicate that the use of oxidizing biocides (particularly chlorine) is favored. Quaternary ammonia salts (quats), a type of nonoxidizing biocide, are also used in many power plant cooling towers. The experience of dealing with regulators to obtain approval to discharge biocides differs significantly between the two industries. In the electric power industry, discharges of any new biocide typically must be approved in writing by the regulatory agency. The approval process for refineries is less formal. In most cases, the refinery must notify the regulatory agency that it is planning to use a new biocide, but the refinery does not need to get written approval before using it. The conclusion of the report is that few of the surveyed facilities are having any difficulty in using and discharging the biocides they want to use.

  13. Methods of vitrifying waste with low melting high lithia glass compositions

    Science.gov (United States)

    Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

    2001-01-01

    The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

  14. Low melting high lithia glass compositions and methods

    Science.gov (United States)

    Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

    2000-01-01

    The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

  15. An optical method for measuring exhaust gas pressure from an internal combustion engine at high speed

    Science.gov (United States)

    Leach, Felix C. P.; Davy, Martin H.; Siskin, Dmitrij; Pechstedt, Ralf; Richardson, David

    2017-12-01

    Measurement of exhaust gas pressure at high speed in an engine is important for engine efficiency, computational fluid dynamics analysis, and turbocharger matching. Currently used piezoresistive sensors are bulky, require cooling, and have limited lifetimes. A new sensor system uses an interferometric technique to measure pressure by measuring the size of an optical cavity, which varies with pressure due to movement of a diaphragm. This pressure measurement system has been used in gas turbine engines where the temperatures and pressures have no significant transients but has never been applied to an internal combustion engine before, an environment where both temperature and pressure can change rapidly. This sensor has been compared with a piezoresistive sensor representing the current state-of-the-art at three engine operating points corresponding to both light load and full load. The results show that the new sensor can match the measurements from the piezoresistive sensor except when there are fast temperature swings, so the latter part of the pressure during exhaust blowdown is only tracked with an offset. A modified sensor designed to compensate for these temperature effects is also tested. The new sensor has shown significant potential as a compact, durable sensor, which does not require external cooling.

  16. Safety aspects of water chemistry in light water reactors

    International Nuclear Information System (INIS)

    1988-12-01

    The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

  17. Flashing of high-pressure saturated water into the pool water

    International Nuclear Information System (INIS)

    Takamasa, Tomoji; Kondo, Koichi; Aya, Izuo.

    1997-01-01

    This paper presents an experimental study on a saturated high-pressure water discharging into a water pool. The purpose of the experiment is to clarify the phenomena that occur by a blow-down of the water from the pressure vessel into the water-filled containment in the case of a wall-crack accident or a LOCA in a passive safety reactor. The results show that a flashing oscillation (FO) occurs when the water discharges into the pool, under specified experimental conditions. The range of the flashing location oscillates between a point very close to and some distance away from the vent hole. The pressures in the vent tube and water pool constantly fluctuate due to the flashing oscillation. The pressure oscillation and alternating flashing location might be caused by the balancing action between the supply of saturated water, flashing at the control volume and steam condensation on the steam-water interface. The frequencies of FO, or frequencies of pressure oscillation and alternating flashing location, increased as water subcooling increased, and as discharging pressure and vent hole diameter decreased. A linear analysis was conducted using a spherical flashing bubble model in which the motion of bubble is controlled by steam condensation. The effects of these parameters on the period of FO in the experiments can be predicted well by the analysis. (author)

  18. Imaging optical probe for pressurized steam-water environment

    International Nuclear Information System (INIS)

    Donaldson, M.R.; Pulfrey, R.E.

    1979-01-01

    An air-cooled imaging optical probe, with an outside diameter of 25.4 mm, has been developed to provide high resolution viewing of flow regimes in a steam-water environment at 343 0 C and 15.2 MPa. The design study considered a 3-m length probe. A 0.3-m length probe prototype was fabricated and tested. The optical probe consists of a 3.5-mm diameter optics train surrounded by two coaxial coolant flow channels and two coaxial insulating dead air spaces. With air flowing through the probe at 5.7 g/s, thermal analysis shows that no part of the optics train will exceed 93 0 C when a 3-m length probe is immersed in a 343 0 C environment. Computer stress analysis plus actual tests show that the probe can operate successfully with conservative safety factors. The imaging optical probe was tested five times in the design environment at the semiscale facility at the INEL. Two-phase flow regimes in the high temperature, high pressure, steam-water blowdown and reflood experiments were recorded on video tape for the first time with the imaging optical probe

  19. A study on optimization of environmental qualification envelope for Kori 3 and 4 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Byun, Choong Sup; Jo, Jong Young [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-07-01

    The purpose of this study is to present the reevaluation of the Main Steam Line Break (MSLB) applied Boron Injection Tank (BIT) removal and to optimize the environmental qualification (EQ) temperature envelope with thermal lag analysis and liquid entrainment method. BIT alleviates the reactor power excursion during Main Steam Line Break (MSLB) accident. Thermal lag analysis methods by NUREG-0588 is using four times condensing heat transfer coefficient on the passive heat sink surface, the forced convection heat transfer coefficient whenever the condensing is not occurring and during blowdown stage. And the entrainment model is that the all of the break regions within the secondary side are represented by non-homogeneous vapor volumes in which the liquid and steam are uniformly mixed throughout. These methods are focused on making higher the surface temperature of the safety equipment. For the analysis, amount of released mass and energy is calculated using the LOFTRAN code and containment temperature is predicted by CONTEMPT-LT 28 code. These two codes are used to for safety analysis. In accordance with the analysis result, a plant specific EQ test envelope was proposed for Kori 3 and 4 NPP.

  20. A study on optimization of environmental qualification envelope for Kori 3 and 4 NPP

    International Nuclear Information System (INIS)

    Song, Dong Soo; Byun, Choong Sup; Jo, Jong Young

    2009-01-01

    The purpose of this study is to present the reevaluation of the Main Steam Line Break (MSLB) applied Boron Injection Tank (BIT) removal and to optimize the environmental qualification (EQ) temperature envelope with thermal lag analysis and liquid entrainment method. BIT alleviates the reactor power excursion during Main Steam Line Break (MSLB) accident. Thermal lag analysis methods by NUREG-0588 is using four times condensing heat transfer coefficient on the passive heat sink surface, the forced convection heat transfer coefficient whenever the condensing is not occurring and during blowdown stage. And the entrainment model is that the all of the break regions within the secondary side are represented by non-homogeneous vapor volumes in which the liquid and steam are uniformly mixed throughout. These methods are focused on making higher the surface temperature of the safety equipment. For the analysis, amount of released mass and energy is calculated using the LOFTRAN code and containment temperature is predicted by CONTEMPT-LT 28 code. These two codes are used to for safety analysis. In accordance with the analysis result, a plant specific EQ test envelope was proposed for Kori 3 and 4 NPP

  1. Analysis on Containment Response Following a LBLOCA of APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [KINS, Daejeon (Korea, Republic of)

    2016-05-15

    The predictions are in good agreements with the final safety analysis report, which implies the containment integrity is maintained during or after an accident like loss of coolant accident. In this study, the CONTEMPT-LT/028 was used to calculate the pressure and temperature, and in the follow-up study, CONTAIN 2.0 will be used for the pressure and temperature predictions in APR1400 reactors. Shin-Hanul Units 1 and 2 may possess different characteristics of peak pressure and temperature in containment following a large break loss-of-coolant-accident. To assess the important performance independently and to compare with prediction results presented in the final safety analysis report (FSAR) of Shin-Hanul Units 1 and 2 might be helpful to regulatory review for identifying validity of the FSAR. The end of blowdown (EOB) time during a LOCA could largely affect the peak pressure and temperature in the containment. This paper provides CONTEMPT-LT/028 prediction of the peak pressure and temperature of Shin-Hanul Units 1 and 2 following a large break loss-of-coolant-accident and compares with licensee's prediction results.

  2. CFD and FEM modeling of PPOOLEX experiments

    Energy Technology Data Exchange (ETDEWEB)

    Paettikangas, T.; Niemi, J.; Timperi, A. (VTT Technical Research Centre of Finland (Finland))

    2011-01-15

    Large-break LOCA experiment performed with the PPOOLEX experimental facility is analysed with CFD calculations. Simulation of the first 100 seconds of the experiment is performed by using the Euler-Euler two-phase model of FLUENT 6.3. In wall condensation, the condensing water forms a film layer on the wall surface, which is modelled by mass transfer from the gas phase to the liquid water phase in the near-wall grid cell. The direct-contact condensation in the wetwell is modelled with simple correlations. The wall condensation and direct-contact condensation models are implemented with user-defined functions in FLUENT. Fluid-Structure Interaction (FSI) calculations of the PPOOLEX experiments and of a realistic BWR containment are also presented. Two-way coupled FSI calculations of the experiments have been numerically unstable with explicit coupling. A linear perturbation method is therefore used for preventing the numerical instability. The method is first validated against numerical data and against the PPOOLEX experiments. Preliminary FSI calculations are then performed for a realistic BWR containment by modeling a sector of the containment and one blowdown pipe. For the BWR containment, one- and two-way coupled calculations as well as calculations with LPM are carried out. (Author)

  3. Influence of shear layers on the structure of shocks formed by rectangular and parabolic blockages placed in a subsonic flow-field

    Science.gov (United States)

    Cheeda, V. K.; Kumar, A.; Ramamurthi, K.

    2014-03-01

    Flow blockages are used to promote the transition of a flame to a detonation. The structure of shock waves formed with different configurations of blockages was experimentally determined for subsonic incoming flow. High speed subsonic flows could develop ahead of a turbulent flame and the interaction of such flows with blockages could lead to the formation of interacting shock waves, slipstreams, and expansion waves. A blow-down test setup was designed to study the interacting shock pattern formed with different configurations of blockages. The flow was found to accelerate to low supersonic velocities during its passage over the blockages. The shock structure downstream of the blockages was found to depend on the shape, size, and number of blockages as well as the spacing between them. While a parabolic-shaped blockage provided shocks of maximum strength, large blockage ratio values did not permit the formation of shocks. The shear layer, formed in the flow downstream of the blockages, reflected the expansion fan as shock waves and was found to be a major feature influencing the formation of the interacting structure of oblique shocks. The structure and strength of the shock waves are analyzed using hodograms. The formation of the interacting family of shock waves using different configurations of blockages and the spacings between them are discussed.

  4. Validation of effective momentum and heat flux models for stratification and mixing in a water pool

    Energy Technology Data Exchange (ETDEWEB)

    Hua Li; Villanueva, W.; Kudinov, P. [Royal Institute of Technology (KTH), Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-06-15

    The pressure suppression pool is the most important feature of the pressure suppression system in a Boiling Water Reactor (BWR) that acts primarily as a passive heat sink during a loss of coolant accident (LOCA) or when the reactor is isolated from the main heat sink. The steam injection into the pool through the blowdown pipes can lead to short term dynamic phenomena and long term thermal transient in the pool. The development of thermal stratification or mixing in the pool is a transient phenomenon that can influence the pool's pressure suppression capacity. Different condensation regimes depending on the pool's bulk temperature and steam flow rates determine the onset of thermal stratification or erosion of stratified layers. Previously, we have proposed to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) and the Effective Momentum Source (EMS) models. The EHS model is used to provide thermal effect of steam injection on the pool, preserving heat and mass balance. The EMS model is used to simulate momentum induced by steam injection in different flow regimes. The EMS model is based on the combination of (i) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (ii) model proposed by Aya and Nariai for prediction of the amplitude and frequency of oscillations at a given pool temperature and steam mass flux. The complete EHS/EMS models only require the steam mass flux, initial pool bulk temperature, and design-specific parameters, to predict thermal stratification and mixing in a pressure suppression pool. In this work we use EHS/EMS models implemented in containment thermal hydraulic code GOTHIC. The PPOOLEX experiments (Lappeenranta University of Technology, Finland) are utilized to (a) quantify errors due to GOTHIC's physical models and numerical schemes, (b) propose necessary improvements in GOTHIC sub-grid scale

  5. Influence of windthrows and tree species on forest soil plant biomass and carbon stocks

    Science.gov (United States)

    Veselinovic, B.; Hager, H.

    2012-04-01

    The role of forests has generally been recognized in climate change mitigation and adaptation strategies and policies (e.g. Kyoto Protocol within articles 3.3 and 3.4, RES-E Directive of EU, Country Biomass Action Plans etc.). Application of mitigation actions, to decrease of CO2-emissions and, as the increase of carbon(C)-stocks and appropriate GHG-accounting has been hampered due to a lack of reliable data and good statistical models for the factors influencing C-sequestration in and its release from these systems (e.g. natural and human induced disturbances). Highest uncertainties are still present for estimation of soil C-stocks, which is at the same time the second biggest C-reservoir on earth. Spruce monocultures have been a widely used management practice in central Europe during the past century. Such stands are in lower altitudes (e.g. submontane to lower montane elevation zone) and on heavy soils unstable and prone to disturbances, especially on blowdown. As the windthrow-areas act as CO2-source, we hypothesize that conversion to natural beech and oak forests will provide sustainable wood supply and higher stability of stands against blowdown, which simultaneously provides the long-term belowground C-sequestration. This work focuses on influence of Norway spruce, Common beech and Oak stands on belowground C-dynamics (mineral soil, humus and belowground biomass) taking into consideration the increased impact of windthrows on spruce monocultures as a result of climate change. For this purpose the 300-700m altitude and pseudogley (planosols/temporally logged) soils were chosen in order to evaluate long-term impacts of the observed tree species on belowground C-dynamics and human induced disturbances on secondary spruce stands. Using the false chronosequence approach, the C-pools have been estimated for different compartments and age classes. The sampling of forest floor and surface vegetation was done using 30x30 (homogenous plots) and 50x50cm (inhomogeneous

  6. Air pressure waves from Mount St. Helens eruptions

    Energy Technology Data Exchange (ETDEWEB)

    Reed, J.W.

    1987-10-20

    Weather station barograph records as well as infrasonic recordings of the pressure wave from the Mount St. Helens eruption of May 18, 1980, have been used to estimate an equivalent explosion airblast yield for this event. Pressure amplitude versus distance patterns in various directions compared with patterns from other large explosions, such as atmospheric nuclear tests, the Krakatoa eruption, and the Tunguska comet impact, indicate that the wave came from an explosion equivalent of a few megatons of TNT. The extent of tree blowdown is considerably greater than could be expected from such an explosion, and the observed forest damage is attributed to outflow of volcanic material. The pressure-time signature obtained at Toledo, Washington, showed a long, 13-min duration negative phase as well as a second, hour-long compression phase, both probably caused by ejacta dynamics rather than standard explosion wave phenomenology. The peculiar audibility pattern, with the blast being heard only at ranges beyond about 100 km, is explicable by finite amplitude propagation effects. Near the source, compression was slow, taking more than a second but probably less than 5 s, so that it went unnoticed by human ears and susceptible buildings were not damaged. There was no damage as Toledo (54 km), where the recorded amplitude would have broken windows with a fast compression. An explanation is that wave emissions at high elevation angles traveled to the upper stratosphere, where low ambient air pressures caused this energetic pressure oscillation to form a shock wave with rapid, nearly instantaneous compression. Atmospheric refraction then returned part of this wave to ground level at long ranges, where the fast compressions were clearly audible. copyright American Geophysical Union 1987

  7. Palisades Nuclear Plant. Semiannual operating report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    The first half of this period was a continuation of Outage 74-4. Retubing of the main condenser was completed and additional Eddy Current Testing in the steam generators was accomplished. On April 2, 1975 the unit returned to service to conduct the Steam Generator Flushing Program. From April 2 through June 20, 1975 the plant operated at power levels up to 90 percent, with two brief outage interruptions. On April 6, the turbine was tripped because of a leak in the flow instrument tap in the drain line from a feedwater heater to the moisture separator drain tank and on April 22, the turbine tripped due to low oil pressure in the electrohydraulic control system caused by a cracked fitting in the piping to a governor valve. In both instances, the plant was returned to service in a matter of a few hours. The plant was removed from service from June 20 until June 30, 1975 for control rod drive seal repair. The plant resumed operation on June 30, 1975. After a few hours' operation, an erratic feedwater control system tripped the unit from service; however, the unit returned to service the same day. An all volatile treatment program for secondary water has achieved reasonable water chemistry stability and continuous blowdown has removed small quantities of phosphate from the steam generators. Operation continued with two low pressure feedwater heaters and drain cooler bypassed in each train, pending retubing of the feedwater heaters. Two rows of blading of a low pressure turbine were replaced. (U.S.)

  8. PELE-IC, 2-D Eulerian Incompressible Hydrodynamic and Bubble Dynamic after LWR LOCA

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.

    1981-01-01

    1 - Description of problem or function: PELE-IC is a two-dimensional semi-implicit Eulerian hydrodynamics program for the solution of incompressible flow coupled to flexible structures. The code was developed to calculate fluid-structure interactions and bubble dynamics of a pressure-suppression system following a loss-of- coolant accident (LOCA). The fluid, structure, and coupling algorithms have been verified by calculation of benchmark problems and air and steam blowdown experiments. The code is written for both plane and cylindrical coordinates. The coupling algorithm is general enough to handle a wide variety of structural shapes. The concepts of void fractions and interface orientation are used to track the movement of free surfaces, allowing great versatility in following fluid-gas interfaces both for bubble definition and water surface motion without the use of marker particles. 2 - Method of solution: The solution strategy is to first solve the Navier-Stokes equations explicitly using values from the previous time-step. Since these values do not necessarily satisfy the continuity equation, the pressure field is iterated upon until the incompressibility condition for each computational cell is satisfied within prescribed limits. The structural motion is computed by a finite element code from the applied pressure at the fluid-structure interface. The shell structure algorithm uses conventional thin-shell theory with transverse shear. The finite-element spatial discretization employs piecewise-linear interpolation functions and one-point quadrature applied to conical frustra. The Newmark implicit time integration method is used as a one-step module. The fluid code then uses the structure's position and velocity as boundary conditions. The fluid pressure field and the structure's response are corrected iteratively until the normal velocities of fluid and structure are equal. The effects of steam condensation and oscillatory chugging on structures are

  9. Experimental and numerical investigation of the cap-shock structure in over expanded thrust-optimized nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Reijasse, P.; Bouvier, F.; Servel, P.

    2002-07-01

    This paper deals with the aerodynamics of an over-expanded nozzle, when the internal parabolic contour of the nozzle extension is highly thrust-optimized in terms of specific impulse-to-weight ratio. This optimization leads to an internal focusing shock issuing from a little downstream from the throat, even when the nozzle is running at nearly vacuum conditions. When such a nozzle is over-expanded, the focusing shock thus interferes with the over-expansion shock, and it forms from this shock interference a particular shock system, named 'cap-shock' because of the cap-like luminous shape seen in the over-expanded plumes of some real engines. Navier-Stokes calcinations performed in Europe had permitted to numerically analyze such a flow pattern, and they have revealed notably a recirculation bubble on the centerline downstream of the Mach disk, which had never been measured yet. A test campaign characterizing the flow separation in over-expanded sub-scale nozzles has been performed in the R2Ch blowdown wind tunnel of the Onera Chalais-Meudon center. Schlieren photographs of the exhaust jet have authorized a detailed description of the cap-shock pattern. Two-components Laser Doppler Velocimetry measurements have confirmed the existence of a recirculation bubble surrounded by an annular supersonic jet and has given its size. In addition to the calculations and the Schlieren interpretative sketches, these first quantitative experimental characterization of the cap-shock structure permit to state a physical description of the cap-shock induced flow field in the thrust-optimized nozzles. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Mueftueoglu, A.K.; Feltus, M.A. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

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

  11. CFX-10 Analysis of the High Temperature Thermal- Chemical Experiment (CS28-2)

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

    2008-02-01

    A Computational Fluid Dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal-chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal-chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (Large Break Loss of Coolant Accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal-chemical behavior of the 28-element fuel channel, the measured temperatures of the Fuel Element Simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions. In spite of some discrepancy between the measurement data and CFX results, it was found that the CFX-10 prediction based on the Urbanic-Heidrick correlation of the zirconium-steam reaction as well as the Discrete Transfer Model for a radiation heat transfer among the FES of three rings and the pressure tube are quite accurate and sound even for the offset a cluster fuel bundle of an aged fuel channel

  12. In vessel core melt progression phenomena

    International Nuclear Information System (INIS)

    Courtaud, M.

    1993-01-01

    For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

  13. Approach-Phase Precision Landing with Hazard Relative Navigation: Terrestrial Test Campaign Results of the Morpheus/ALHAT Project

    Science.gov (United States)

    Crain, Timothy P.; Bishop, Robert H.; Carson, John M., III; Trawny, Nikolas; Hanak, Chad; Sullivan, Jacob; Christian, John; DeMars, Kyle; Campbell, Tom; Getchius, Joel

    2016-01-01

    The Morpheus Project began in late 2009 as an ambitious e ort code-named Project M to integrate three ongoing multi-center NASA technology developments: humanoid robotics, liquid oxygen/liquid methane (LOX/LCH4) propulsion and Autonomous Precision Landing and Hazard Avoidance Technology (ALHAT) into a single engineering demonstration mission to be own to the Moon by 2013. The humanoid robot e ort was redirected to a deploy- ment of Robonaut 2 on the International Space Station in February of 2011 while Morpheus continued as a terrestrial eld test project integrating the existing ALHAT Project's tech- nologies into a sub-orbital ight system using the world's rst LOX/LCH4 main propulsion and reaction control system fed from the same blowdown tanks. A series of 33 tethered tests with the Morpheus 1.0 vehicle and Morpheus 1.5 vehicle were conducted from April 2011 - December 2013 before successful, sustained free ights with the primary Vertical Testbed (VTB) navigation con guration began with Free Flight 3 on December 10, 2013. Over the course of the following 12 free ights and 3 tethered ights, components of the ALHAT navigation system were integrated into the Morpheus vehicle, operations, and ight control loop. The ALHAT navigation system was integrated and run concurrently with the VTB navigation system as a reference and fail-safe option in ight (see touchdown position esti- mate comparisons in Fig. 1). Flight testing completed with Free Flight 15 on December 15, 2014 with a completely autonomous Hazard Detection and Avoidance (HDA), integration of surface relative and Hazard Relative Navigation (HRN) measurements into the onboard dual-state inertial estimator Kalman lter software, and landing within 2 meters of the VTB GPS-based navigation solution at the safe landing site target. This paper describes the Mor- pheus joint VTB/ALHAT navigation architecture, the sensors utilized during the terrestrial ight campaign, issues resolved during testing, and the navigation

  14. Laguna Verde annulus pressurization loads evaluation

    International Nuclear Information System (INIS)

    Castaneda, M. A.; Cruz, M. A.; Cardenas, J. B.; Vargas, A.; Cruz, H. J.; Mercado, J. J.

    2010-10-01

    Annulus pressurization, jet impingement, pipe whip restraint and jet thrust are phenomena related to postulated pipe ruptures. A postulated pipe rupture at the weld between recirculation, or feedwater piping and a reactor nozzle safe end, will lead to a high flow rate of flashing water/steam mixture into the annulus between the reactor pressure vessel and the biological shield wall. The total effect of the vessel and pipe inventory blowdown from the break being postulated must be accounted for in the evaluation. A recirculation line break will give rise to an angular dependent short term pressure differential around the vessel, followed by a longer term pressure buildup in the annulus. A recirculation line postulated rupture may not produce worst case conditions and reference to time intervals for only the recirculation break should be treated superficially. A postulated rupture of the feedwater piping may produce the extreme case for determining: 1) the shield wall and reactor vessel to pedestal interactions, 2) loading on the reactor vessel internals, or 3) responses for the balance of piping attached to the vessel. Recently it was identified a potential issue regarding the criteria used to determine which cases were evaluated for Annulus Pressurization (A P) loads for new loads plants. The original A P loads methodology in the late 1970 and early 1980 years separated the mass/energy release calculation from the structural response calculation based on the implicit assumption that the maximum overall mass/energy release will result in maximizing the structural response and corresponding stresses on the reactor pressure vessel, internals, and containment structures. This process did not consider the dynamic response in the primary and secondary safety related structures, components and equipment. Consequently, the A P loads used as input for design adequacy evaluations of Nuclear Steam Supply System safety related components for new loads plants might have

  15. Breckinridge Project, initial effort

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-01-01

    The project cogeneration plant supplies electric power, process steam and treated boiler feedwater for use by the project plants. The plant consists of multiple turbine generators and steam generators connected to a common main steam header. The major plant systems which are required to produce steam, electrical power and treated feedwater are discussed individually. The systems are: steam, steam generator, steam generator fuel, condensate and feedwater deaeration, condensate and blowdown collection, cooling water, boiler feedwater treatment, coal handling, ash handling (fly ash and bottom ash), electrical, and control system. The plant description is based on the Phase Zero design basis established for Plant 31 in July of 1980 and the steam/condensate balance as presented on Drawing 31-E-B-1. Updating of steam requirements as more refined process information becomes available has generated some changes in the steam balance. Boiler operation with these updated requirements is reflected on Drawing 31-D-B-1A. The major impact of updating has been that less 600 psig steam generated within the process units requires more extraction steam from the turbine generators to close the 600 psig steam balance. Since the 900 psig steam generation from the boilers was fixed at 1,200,000 lb/hr, the additional extraction steam required to close the 600 psig steam balance decreased the quantity of electrical power available from the turbine generators. In the next phase of engineering work, the production of 600 psig steam will be augmented by increasing convection bank steam generation in the Plant 3 fired heaters by 140,000 to 150,000 lb/hr. This modification will allow full rated power generation from the turbine generators.

  16. Ethanolamine experience at Koeberg nuclear power station, South Africa

    International Nuclear Information System (INIS)

    Galt, K.J.; Caris, N.B.

    2002-01-01

    Following testing of ethanolamine as an alternative to ammonia on Unit 2 in 1997, Unit 1 of the Koeberg Nuclear Power Station was converted to ethanolamine in 1998. The Unit has now operated for just over one and a half cycle on ETA. The decision to change to ETA was made to achieve further reductions in feedwater iron transport. Koeberg has always operated ammonia/hydrazine AVT control and ran the feedwater pH at 9.6-9.7 before the changeover. The original pH levels were increased in response to concerns over flow-accelerated corrosion. A by product of reducing the FAC rates is a reduction in iron transport. Although nominally all-ferrous, there are a number of small copper-containing components and the Koeberg Engineering Department would not countenance a further increase in ammonia concentrations in case of copper transport to the SGs. This led to ethanolamine being selected as an alternative to ammonia. The Koeberg condensate polishing plant has been modified, largely to accommodate ETA operation, but is not currently operable in the modified configuration. It is therefore on standby while ETA is implemented. The SG blowdown demineralizers have begun to be operated past ammonia/ETA break, but optimisation is largely dependent on CPP availability in the modified configuration. This paper documents the Koeberg experience to date of operation under an ethanolamine-AVT regime. As one of the few plants outside of the USA to have changed to ethanolamine, it is hoped we can make a valuable contribution for other non-US plants considering such a move. (authors)

  17. Low Temperature Geothermal Resource Assessment for Membrane Distillation Desalination in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Akar, Sertac; Turchi, Craig

    2017-05-01

    Substantial drought and declines in potable groundwater in the United States over the last decade has increased the demand for fresh water. Desalination of saline water such as brackish surface or groundwater, seawater, brines co-produced from oil and gas operations, industrial wastewater, blow-down water from power plant cooling towers, and agriculture drainage water can reduce the volume of water that requires disposal while providing a source of high-quality fresh water for industrial or commercial use. Membrane distillation (MD) is a developing technology that uses low-temperature thermal energy for desalination. Geothermal heat can be an ideal thermal-energy source for MD desalination technology, with a target range of $1/m3 to $2/m3 for desalinated water depending on the cost of heat. Three different cases were analyzed to estimate levelized cost of heat (LCOH) for integration of MD desalination technology with low-grade geothermal heat: (1) residual heat from injection brine at a geothermal power plant, (2) heat from existing underutilized low-temperature wells, and (3) drilling new wells for low-temperature resources. The Central and Western United States have important low-temperature (<90 degrees C) geothermal resource potential with wide geographic distribution, but these resources are highly underutilized because they are inefficient for power production. According to the USGS, there are 1,075 identified low temperature hydrothermal systems, 55 low temperature sedimentary systems and 248 identified medium to high temperature geothermal systems in the United States. The estimated total beneficial heat potential from identified low temperature hydrothermal geothermal systems and residual beneficial heat from medium to high temperature systems is estimated as 36,300 MWth, which could theoretically produce 1.4 to 7 million m3/day of potable water, depending on desalination efficiency.

  18. Final environmental statement related to construction of Cherokee Nuclear Station, Units 1, 2, and 3: (Docket Nos. STN 50-491, STN 50-492, and STN 50-493)

    International Nuclear Information System (INIS)

    1975-10-01

    The proposed action is the issuance of a construction permit to the Duke Power Company for the construction of the Cherokee Nuclear Station (CNS) Units 1, 2, and 3 located in Cherokee County, South Carolina. A total of 2263 acres will be removed from public use for the CNS site. Construction-related activities on the site will disturb about 751 acres. Approximately 654 acres of land will be required for transmission line right-of-way, and a railroad spur will affect 83 acres. This constitutes a minor regional impact. No significant environmental impacts are anticipated from normal operational releases of radioactive materials. The total annual dose to the US population (total body plus thyroid) from operation of the plant is 210 man-rems which is less than the normal fluctuations in the background dose this population would receive. The occupational dose is approximately 1400 man-rems/year. The heat dissipation system will require a maximum water makeup of 55,814 gpm, of which 50,514 gpm will be consumed due to drift and evaporative losses. This amount represents 4.5% of the mean monthly flow and 23.8% of the low flow of the Broad River. The cooling tower blowdown and chemical effluents from the station will increase the dissolved solids concentration in the river by a maximum of 44 ppM. The thermal alterations and increases in total dissolved solids concentration will not significantly affect the aquatic productivity of the river. 114 refs., 25 figs., 46 tabs

  19. The probability of containment failure by direct containment heating in Zion

    Energy Technology Data Exchange (ETDEWEB)

    Pilch, M.M. [Sandia National Labs., Albuquerque, NM (United States); Yan, H.; Theofanous, T.G. [California Univ., Santa Barbara, CA (United States)

    1994-12-01

    This report is the first step in the resolution of the Direct Containment Heating (DCH) issue for the Zion Nuclear Power Plant using the Risk Oriented Accident Analysis Methodology (ROAAM). This report includes the definition of a probabilistic framework that decomposes the DCH problem into three probability density functions that reflect the most uncertain initial conditions (UO{sub 2} mass, zirconium oxidation fraction, and steel mass). Uncertainties in the initial conditions are significant, but our quantification approach is based on establishing reasonable bounds that are not unnecessarily conservative. To this end, we also make use of the ROAAM ideas of enveloping scenarios and ``splintering.`` Two causal relations (CRs) are used in this framework: CR1 is a model that calculates the peak pressure in the containment as a function of the initial conditions, and CR2 is a model that returns the frequency of containment failure as a function of pressure within the containment. Uncertainty in CR1 is accounted for by the use of two independently developed phenomenological models, the Convection Limited Containment Heating (CLCH) model and the Two-Cell Equilibrium (TCE) model, and by probabilistically distributing the key parameter in both, which is the ratio of the melt entrainment time to the system blowdown time constant. The two phenomenological models have been compared with an extensive database including recent integral simulations at two different physical scales. The containment load distributions do not intersect the containment strength (fragility) curve in any significant way, resulting in containment failure probabilities less than 10{sup {minus}3} for all scenarios considered. Sensitivity analyses did not show any areas of large sensitivity.

  20. Fulton Generating Station Units 1 and 2 (Docket Nos. 50-463 and 50-464): Final environmental statement

    International Nuclear Information System (INIS)

    1975-04-01

    The proposed action is the issuance of construction permits to the Philadelphia Electric Company for the construction of the Fulton Generating Station, Units 1 and 2, located in Fulton and Drumore Townships, Lancaster County, Pennsylvania. Makeup water for cooling will be withdrawn form Conowingo Pond at a maximum rate of 43,000 gpm. The dissolved solids content of the blowdown water will be increased by a factor of about two. The remainder of the water will be evaporated to the atmosphere by cooling towers. About 10 acres offsite, some 7 acres of which is woodland, will be used for railroad-spur construction. About 0.25 mile of new transmission-line rights-of-way (9 acres) will be needed, although 49 miles of new transmission line, which will require about 3 miles of selective clearing, will be constructed on existing rights-of-way. An unestablished amount of land will be used for access-road construction, but the applicant will use existing roadway corridors where feasible. A small loss of consumer species will result from loss of habitat. Some loss of benthic and pelagic organisms in Conowingo Pond will be caused by intake and discharge construction. The Station's thermal and chemical discharges will meet the State water-quality standards. The duration of additional ground-level fog caused by Station operation is expected to be less than 3 hr/year. (Sect. 5.3.3). No observable effects are expected from salt deposition from cooling-tower drift. (Sect. 5.3.3). Decomposers, primary producers, and zooplankton will be entrained and killed in the cooling-tower system; they, as well as benthic organisms, will be affected by the heated-water discharge. This loss will have little effect on the pond food web. 30 figs., 76 tabs

  1. Final supplement to the final environmental statement related to construction of Skagit Nuclear Power Project, Units 1 and 2: (Docket Nos. STN 50-522 and STN 50-523)

    International Nuclear Information System (INIS)

    1977-04-01

    The proposed action is the issuance of construction permits to the Puget Sound Power and Light Company, Pacific Power and Light Company, Washington Water Power Company and the Portland General Electric Company, for the construction of Skagit Nuclear Power Projects Units 1 and 2 in Skagit County, Washington (about 64 miles north of Seattle and 6 miles ENE of Sedro Woolley). These units are scheduled for commercial service in 1984 and 1986, respectively. The exhaust steam from the turbine-generators will be cooled in condensers which will utilize one hyperbolic-type natural-draft cooling tower per unit to dissipate heat to the atmosphere. Water (106 cfs max.) for the cooling tower makeup (82.4 cfs) and other plant uses will be withdrawn from the Skagit River through Ranney Collectors embedded in the north bank of the river and pumped to the plant through a pipeline about 35,000 ft. long. Cooling tower blowdown (7 cfs max.) from the project and dilution water (20 cfs max.) will flow through a pipeline back to the river where it will be discharged through a diffuser. Approximately 1750 acres of forested and agricultural land will be removed from harvesting for the life of the power plant; 360 acres of this land will be diverted to industrial use. This will affect less than 0.5% of standing forest in Skagit County and 16 acres in cultivated crops and pasturage. Increased siltation of onsite creeks and the Skagit River from construction work and the small amounts of heat and chemicals discharged to the river during plant operation will have insignificant impacts on water quality and aquatic biota due to erosion control efforts and dilution by the large river flow. (16,200 cfs average; 4740 cfs 7-day, 10-yr low). 18 figs

  2. An improved gate valve for critical applications in nuclear power plants

    International Nuclear Information System (INIS)

    Kalsi, M.S.; Alvarez, P.D.; Wang, J.K.; Somagyi, D.

    1996-01-01

    U.S. Nuclear Regulatory Commission Generic Letters 89-10 for motor-operated valves (MOVs) and 95-07 for all power-operated valves document in detail the problems related to the performance of the safety-related valves in nuclear power plants. The problems relate to lack of reliable operation under design basis conditions including higher than anticipated stem thrust, unpredictable valve behavior, damage to the valve internals under blowdown/high flow conditions, significant degradation of performance when cycled under AP and flow, thermal binding, and pressure locking. This paper describes an improved motor-operated flexible wedge gate valve design, the GE Sentinel Valve, which is the outcome of a comprehensive and systematic development effort undertaken to resolve the issues identified in the NRC Generic Letters 89-10 and 95-07. The new design provides a reliable, long-term, low maintenance cost solution to the nuclear power industry. One of the key features incorporated in the disc permits the disc flexibility to be varied independently of the disc thickness (pressure boundary) dictated by the ASME Section III Pressure Vessel ampersand Piping Code stress criteria. This feature allows the desired flexibility to be incorporated in the disc, thus eliminating thermal binding problems. A matrix of analyses was performed using finite element and computational fluid dynamics approaches to optimize design for stresses, flexibility, leak-tightness, fluid flow, and thermal effects. The design of the entire product line was based upon a consistent set of analyses and design rules which permit scaling to different valve sizes and pressure classes within the product line. The valve meets all of the ASME Section III Code design criteria and the N-Stamp requirements. The performance of the valve was validated by performing extensive separate effects and plant in-situ tests. This paper summarizes the key design features, analyses, and test results

  3. The development and application of overheating failure model of FBR steam generator tubes. 2

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi

    2001-11-01

    The JNC technical report 'The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes' summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. 1. On the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. 2. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. 3. Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure. (author)

  4. The sensitivity analysis by adjoint method for the uncertainty evaluation of the CATHARE-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Barre, F.; de Crecy, A.; Perret, C. [French Atomic Energy Commission (CEA), Grenoble (France)

    1995-09-01

    This paper presents the application of the DASM (Discrete Adjoint Sensitivity Method) to CATHARE 2 thermal-hydraulics code. In a first part, the basis of this method is presented. The mathematical model of the CATHARE 2 code is based on the two fluid six equation model. It is discretized using implicit time discretization and it is relatively easy to implement this method in the code. The DASM is the ASM directly applied to the algebraic system of the discretized code equations which has been demonstrated to be the only solution of the mathematical model. The ASM is an integral part of the new version 1.4 of CATHARE. It acts as a post-processing module. It has been qualified by comparison with the {open_quotes}brute force{close_quotes} technique. In a second part, an application of the DASM in CATHARE 2 is presented. It deals with the determination of the uncertainties of the constitutive relationships, which is a compulsory step for calculating the final uncertainty of a given response. First, the general principles of the method are explained: the constitutive relationship are represented by several parameters and the aim is to calculate the variance-covariance matrix of these parameters. The experimental results of the separate effect tests used to establish the correlation are considered. The variance of the corresponding results calculated by CATHARE are estimated by comparing experiment and calculation. A DASM calculation is carried out to provide the derivatives of the responses. The final covariance matrix is obtained by combination of the variance of the responses and those derivatives. Then, the application of this method to a simple case-the blowdown Canon experiment-is presented. This application has been successfully performed.

  5. Slag melt granulation and factors affecting the quality of Granulated slag

    Directory of Open Access Journals (Sweden)

    Володимир Петрович Кравченко

    2015-10-01

    Full Text Available An analysis of the state of slags recycling in foreign countries was carried out. A modern principle was put forward in the article: blast furnace is an apparatus for manufacturing of two basic types of products : cast iron and slag. Granulation, as the primary recycling of slag melt fixes the structure with certain properties at rapid cooling. An analysis of the existing methods of granulation was carried out and factors influencing the quality of granular slag were determined, as well as the ways of obtaining granular slag with the required physical and mechanical characteristics. The main factors of granulated slags quality, employed for manufacturing of binding materials are chemical composition and the structure of fine granulated particles. All wet methods of granulation are characterized by high humidity of granulated slag, its value reaching 24,5%, due to increase in granules’ porosity. Real options for reducing humidity of granulated slag may include: development of the process of granulation, ensuring manufacturing of products with increased density and low content of fine fractions, dehydration of slag in high bunkers and stacks at sufficient soaking time and slag blowdown with a stream moving downwards. Using mechanical granulators and gaseous energy carriers (air for melt’s dispersion is an efficient way of reducing water consumption for granulation (semi-dry or dry methods of granulation. It also makes it possible to reduce r consumption of water, supplied for granulation from 3,0 to 0,7-1,5m3/min. Application of air blast for melt’s dispersion influences its fractional composition and grain shape in the slag: the content of the fraction less than 1,25mm reduces to 49,1%, as compared to conventional 92,8%. The content of spherical grains is with tough surface is 33%, it promoting reduction of residual humidity of granulated slag. Thus, application of air blast for granulation of slags is an efficient way of obtaining high

  6. Coolant voiding analysis following SGTR for an HLMC reactor

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

    2000-01-01

    Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

  7. LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, Thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCE is expected to closely model a LPWR LOCA. 2 - Description of test: Experiment LP-LB-1 was conducted on 3 February 1984 in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Cooperation and Development. The primary objectives of Experiment LP-LB-1 were to determine system transient characteristics and to assess code predictive capabilities for design basis large-break loss-of-coolant accidents in pressurized water reactors (PWRs). This experiment simulated a double-ended offset shear of one inlet pipe in a four-loop PWR and was initiated from conditions representative of licensing limits in a PWR. Other boundary conditions for the simulation were loss of offsite power, rapid primary coolant pump coast down, and United Kingdom minimum safeguard emergency core coolant injection rates. The nuclear fuel rods were not pressurized. The transient was initiated by opening the quick-opening blowdown valves in the broken loop hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  8. An improved gate valve for critical applications in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kalsi, M.S.; Alvarez, P.D.; Wang, J.K.; Somagyi, D. [Kalsi Engineering, Inc., Sugar Land, TX (United States)] [and others

    1996-12-01

    U.S. Nuclear Regulatory Commission Generic Letters 89-10 for motor-operated valves (MOVs) and 95-07 for all power-operated valves document in detail the problems related to the performance of the safety-related valves in nuclear power plants. The problems relate to lack of reliable operation under design basis conditions including higher than anticipated stem thrust, unpredictable valve behavior, damage to the valve internals under blowdown/high flow conditions, significant degradation of performance when cycled under AP and flow, thermal binding, and pressure locking. This paper describes an improved motor-operated flexible wedge gate valve design, the GE Sentinel Valve, which is the outcome of a comprehensive and systematic development effort undertaken to resolve the issues identified in the NRC Generic Letters 89-10 and 95-07. The new design provides a reliable, long-term, low maintenance cost solution to the nuclear power industry. One of the key features incorporated in the disc permits the disc flexibility to be varied independently of the disc thickness (pressure boundary) dictated by the ASME Section III Pressure Vessel & Piping Code stress criteria. This feature allows the desired flexibility to be incorporated in the disc, thus eliminating thermal binding problems. A matrix of analyses was performed using finite element and computational fluid dynamics approaches to optimize design for stresses, flexibility, leak-tightness, fluid flow, and thermal effects. The design of the entire product line was based upon a consistent set of analyses and design rules which permit scaling to different valve sizes and pressure classes within the product line. The valve meets all of the ASME Section III Code design criteria and the N-Stamp requirements. The performance of the valve was validated by performing extensive separate effects and plant in-situ tests. This paper summarizes the key design features, analyses, and test results.

  9. Final environmental statement: Related to the operation of Davis-Besse Nuclear Power Station, Unit 1 (Docket No. 50-346)

    International Nuclear Information System (INIS)

    1975-10-01

    The proposed action is the issuance of an operating license to the Toledo Edison Company and the Cleveland Electric Illuminating Company for the startup and operation of the Davis-Besse Nuclear Power Station Unit 1 (the station) located near Port Clinton in Ottawa County, Ohio. The total site area is 954 acres of which 160 acres have been removed from production of grain crops and converted to industrial use. Approximately 600 acres of the area is marshland which will be maintained as a wildlife refuge. The disturbance of the lake shore and lake bottom during construction of the station water intake and discharge pipes resulted in temporary turbidity, silting, and destruction of bottom organisms. Since completion of these activities, evidence of improvement in turbidity and transparency measurements, and the reestablishment of the bottom organism has been obtained. The cooling tower blowdown and service water which the station discharges to Lake Erie, via a submerged jet, will be heated no more than 20/degrees/F above the ambient lake water temperature. Although some small fish and plankton in the discharge water plume will be disabled as a result of thermal shock, exposure to chlorine and buffeting, few adult fish will be affected. The thermal plume resulting from the maximum thermal discharge is calculated to have an area of less than one acre within the 3/degrees/F isotherm (above lake ambient). Approximately 101 miles of transmission lines have been constructed, primarily over existing farmland, requiring about 1800 acres of land for the rights-of-way. Land use will essentially be unchanged since only the land required for the base of the towers is removed from production. Herbicides will not be used to maintain the rights-of-way. 14 figs., 34 refs

  10. Low Temperature Geothermal Resource Assessment for Membrane Distillation Desalination in the United States: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Akar, Sertac; Turchi, Craig

    2016-10-01

    Substantial drought and declines in potable groundwater in the United States over the last decade has increased the demand for fresh water. Desalination of saline water such as brackish surface or groundwater, seawater, brines co-produced from oil and gas operations, industrial wastewater, blow-down water from power plant cooling towers, and agriculture drainage water can reduce the volume of water that requires disposal while providing a source of high-quality fresh water for industrial or commercial use. Membrane distillation (MD) is a developing technology that uses low-temperature thermal energy for desalination. Geothermal heat can be an ideal thermal-energy source for MD desalination technology, with a target range of $1/m3 to $2/m3 for desalinated water depending on the cost of heat. Three different cases were analyzed to estimate levelized cost of heat (LCOH) for integration of MD desalination technology with low-grade geothermal heat: (1) residual heat from injection brine at a geothermal power plant, (2) heat from existing underutilized low-temperature wells, and (3) drilling new wells for low-temperature resources. The Central and Western United States have important low-temperature (<90 degrees C) geothermal resource potential with wide geographic distribution, but these resources are highly underutilized because they are inefficient for power production. According to the USGS, there are 1,075 identified low temperature hydrothermal systems, 55 low temperature sedimentary systems and 248 identified medium to high temperature geothermal systems in the United States. The estimated total beneficial heat potential from identified low temperature hydrothermal geothermal systems and residual beneficial heat from medium to high temperature systems is estimated as 36,300 MWth, which could theoretically produce 1.4 to 7 million m3/day of potable water, depending on desalination efficiency.

  11. Mountain pine beetles and emerging issues in the management of woodland caribou in Westcentral British Columbia

    Directory of Open Access Journals (Sweden)

    Deborah Cichowski

    2005-05-01

    Full Text Available The Tweedsmuir—Entiako caribou (Rangifer tarandus caribou herd summers in mountainous terrain in the North Tweedsmuir Park area and winters mainly in low elevation forests in the Entiako area of Westcentral British Columbia. During winter, caribou select mature lodgepole pine (Pinus contorta forests on poor sites and forage primarily by cratering through snow to obtain terrestrial lichens. These forests are subject to frequent large-scale natural disturbance by fire and forest insects. Fire suppression has been effective in reducing large-scale fires in the Entiako area for the last 40—50 years, resulting in a landscape consisting primarily of older lodgepole pine forests, which are susceptible to mountain pine beetle (Dendroctonus ponderosae attack. In 1994, mountain pine beetles were detected in northern Tweedsmuir Park and adjacent managed forests. To date, mountain pine beetles have attacked several hundred thousand hectares of caribou summer and winter range in the vicinity of Tweedsmuir Park, and Entiako Park and Protected Area. Because an attack of this scale is unprecedented on woodland caribou ranges, there is no information available on the effects of mountain pine beetles on caribou movements, habitat use or terrestrial forage lichen abundance. Implications of the mountain pine beetle epidemic to the Tweedsmuir—Entiako woodland caribou population include effects on terrestrial lichen abundance, effects on caribou movement (reduced snow interception, blowdown, and increased forest harvesting outside protected areas for mountain pine beetle salvage. In 2001 we initiated a study to investigate the effects of mountain pine beetles and forest harvesting on terrestrial caribou forage lichens. Preliminary results suggest that the abundance of Cladina spp. has decreased with a corresponding increase in kinnikinnick (Arctostaphylos uva-ursi and other herbaceous plants. Additional studies are required to determine caribou movement and

  12. Relative vegetation profiles in a Neotropical forest: comparison of lidar instrumentation and field-based measurements

    Science.gov (United States)

    Sullivan, F. B.; Palace, M. W.; Ducey, M.; Czarnecki, C.; Zanin Shimbo, J.; Mota e Silva, J.

    2012-12-01

    Tropical forests are considered to be some of the most structurally complex forests in the world. Understanding vegetation height structure in these forests can aid in understanding the spatial temporal components of disturbance, from blowdowns to gap dynamics. Vegetation profiles can be used to better estimate carbon storage and flux across the landscape. Using light detection and ranging (lidar) data collected at La Selva, Costa Rica from four instruments (three airborne, one terrestrial) at four times since 2005, and field data collected in January 2012, we generated relative vegetation profiles for twenty plots in La Selva. Relative vegetation profiles were derived from lidar data by accounting for obscured plant material through a log transformation of the cumulative proportion of observations (percent canopy closure). Profiles were derived from field data using two different sets of allometric equations describing crown shape and tree height. We conducted a cluster analysis on similarity matrices developed in R (version 2.14.1) using three different metrics (sum of squares, Kullback-Leibler divergence, Kolmogorov-Smirnov D statistic) and identified general similarity between lidar profiles. Results were consistent across each of the three similarity metrics. Three distinct clusters were found, with profiles from three airborne lidar instruments, two profiles from a terrestrial lidar instrument, and profiles derived from field data forming the clusters. Our results indicate that although estimating lidar relative vegetation profiles from field data was not possible, terrestrial lidar relative vegetation profiles are generally similar to airborne relative vegetation profiles. Given the rapidity and repeatability of terrestrial lidar measurements, these results show promise for terrestrial lidar instruments to collect plot-specific data on forest structure and vertical distribution of plant material. Furthermore, identifying relationships between terrestrial and

  13. Hydrogen production from municipal solid waste

    Energy Technology Data Exchange (ETDEWEB)

    Wallman, P.H.; Richardson, J.H.; Thorsness, C.B. [and others

    1996-06-28

    We have modified a Municipal Solid Waste (MSW) hydrothermal pretreatment pilot plant for batch operation and blowdown of the treated batch to low pressure. We have also assembled a slurry shearing pilot plant for particle size reduction. Waste paper and a mixture of waste paper/polyethylene plastic have been run in the pilot plant with a treatment temperature of 275{degrees}C. The pilot-plant products have been used for laboratory studies at LLNL. The hydrothermal/shearing pilot plants have produced acceptable slurries for gasification tests from a waste paper feedstock. Work is currently underway with combined paper/plastic feedstocks. When the assembly of the Research Gasification Unit at Texaco (feed capacity approximately 3/4-ton/day) is complete (4th quarter of FY96), gasification test runs will commence. Laboratory work on slurry samples during FY96 has provided correlations between slurry viscosity and hydrothermal treatment temperature, degree of shearing, and the presence of surfactants and admixed plastics. To date, pumpable slurries obtained from an MSW surrogate mixture of treated paper and plastic have shown heating values in the range 13-15 MJ/kg. Our process modeling has quantified the relationship between slurry heating value and hydrogen yield. LLNL has also performed a preliminary cost analysis of the process with the slurry heating value and the MSW tipping fee as parameters. This analysis has shown that the overall process with a 15 MJ/kg slurry gasifier feed can compete with coal-derived hydrogen with the assumption that the tipping fee is of the order $50/ton.

  14. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    2014-02-01

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  15. 2009 Industry update on dispersant use for steam generator fouling control

    Energy Technology Data Exchange (ETDEWEB)

    Fruzzetti, K. [Electric Power Research Inst., Palo Alto, California (United States); Kreider, M.A.; Anderson, C.E.; Marks, C.R. [Dominion Engineering, Inc., Reston, Virginia (United States); Morey, D.O. [Exelon Corporation (United States); Walton, P.R. [Exelon Corporation (United States); TMI Nuclear Generating Station, Middleton, PA (United States)

    2009-07-01

    During the last few decades, utilities have spent considerable resources reducing the quantity and rate of accumulation of corrosion product deposits within pressurized water reactor (PWR) steam generators (SGs). In the past, two basic approaches have been used: Reducing the corrosion product ingress rate (e.g., by replacing secondary components with corrosion-susceptible materials, implementing favorable chemistry changes, etc.); and, Removing corrosion products which have accumulated in the SGs through top-of-tubesheet (TTS) sludge lancing and other chemical and mechanical methods. Despite the success of these methods, there are limitations, including practical lower limits on the feedwater iron concentration and the high cost and effectiveness limits of cleaning techniques (particularly in crevices). A third approach is the addition of a polymeric dispersant to promote suspension of corrosion iron, thereby reducing deposition onto SG surfaces and facilitating more efficient removal via blowdown (for online applications) or SG / secondary system draining (for wet layup or longpath cleanup applications). More than a decade of qualification work and two full-scale plant trials - at ANO-2 in 2000 and at McGuire Unit 2 in 2005-2007 - addressed initial technical concerns, paving the way for routine online long-term use (LTU) in nuclear recirculating SGs. As a result, during the last five years, utilities in both the US and overseas have initiated plant-specific efforts that build on past qualifications and the two plant trials in preparation for routine LTU of dispersants (e.g., installation of injection equipment and plant-specific evaluations of materials compatibility and SG thermal performance). (author)

  16. Improved management of SG BD demineralizer for reduced generation of low-level radioactive spent resin in Korean nuclear power plants

    International Nuclear Information System (INIS)

    Rhee, I.; Cho, D.; Yeon, J.

    2003-01-01

    Most nuclear power plants in Korea have adopted Ethanolamine(ETA) as a secondary pH control agent to increase the pH at the liquid phase, which may reduce the corrosion in steam generator tubes and moisture separator/reheat system. Along with its beneficial effect of SG protection from corrosion and degradation, the replacement of ammonia with ETA causes the increased generation of spent resin and the reduced run time of demineralizer in steam generator blowdown(SG BD) system. The composition ratio of cation- to anion- exchange resin in SG BD mixed bed should be increased in the ETA chemistry environment to meet the ratio of cation to anion in the aqueous solution, which results in the simultaneous exhaustion of cation and anion exchange resins. The utilization rate of mixed bed is greatest at the cation-to-anion ratio of 95:1 on the theoretical equivalent basis in the solution, but practically highest at that of 22:1 due to the possible inhomogeneous distribution of cation and anion exchange resins in SG BD bed. The run time of the bed could be extended by 30% such that, at that much, the purchase cost of new resin is saved and the production rate of spent resin is reduced. The guideline on the replacement of resin in SG BD bed is not necessary to secure the removal of radioactive particles without the leakage of the primary coolant into the secondary side since all the radioactive ions can be eliminated by SG BD bed with the sufficient time. They are retained during more than one month after their ingress into the SG BD bed without leakage. With the reduced replacement, thus, the SG BD spent resin that comprises 65% of low-level radioactive solid waste can be much cut down

  17. Orbital Express fluid transfer demonstration system

    Science.gov (United States)

    Rotenberger, Scott; SooHoo, David; Abraham, Gabriel

    2008-04-01

    Propellant resupply of orbiting spacecraft is no longer in the realm of high risk development. The recently concluded Orbital Express (OE) mission included a fluid transfer demonstration that operated the hardware and control logic in space, bringing the Technology Readiness Level to a solid TRL 7 (demonstration of a system prototype in an operational environment). Orbital Express (funded by the Defense Advanced Research Projects Agency, DARPA) was launched aboard an Atlas-V rocket on March 9th, 2007. The mission had the objective of demonstrating technologies needed for routine servicing of spacecraft, namely autonomous rendezvous and docking, propellant resupply, and orbital replacement unit transfer. The demonstration system used two spacecraft. A servicing vehicle (ASTRO) performed multiple dockings with the client (NextSat) spacecraft, and performed a variety of propellant transfers in addition to exchanges of a battery and computer. The fluid transfer and propulsion system onboard ASTRO, in addition to providing the six degree-of-freedom (6 DOF) thruster system for rendezvous and docking, demonstrated autonomous transfer of monopropellant hydrazine to or from the NextSat spacecraft 15 times while on orbit. The fluid transfer system aboard the NextSat vehicle was designed to simulate a variety of client systems, including both blowdown pressurization and pressure regulated propulsion systems. The fluid transfer demonstrations started with a low level of autonomy, where ground controllers were allowed to review the status of the demonstration at numerous points before authorizing the next steps to be performed. The final transfers were performed at a full autonomy level where the ground authorized the start of a transfer sequence and then monitored data as the transfer proceeded. The major steps of a fluid transfer included the following: mate of the coupling, leak check of the coupling, venting of the coupling, priming of the coupling, fluid transfer, gauging

  18. Last 20 years of gas hydrates in the oil industry : challenges and achievements in predicting pipeline blockage

    Energy Technology Data Exchange (ETDEWEB)

    Estanga, D.A.; Creek, J.; Subramanian, S.; Kini, R.A. [Chevron Energy Technology Co., Houston, TX (United States)

    2008-07-01

    This paper reviewed how the successes of the past 20 years have shaped the new hydrate focus. It also outlined innovative tools for hydrate plugging prediction. Tools such as CSMHyK-OLGA were developed to address the design and operational challenges associated with offshore production regarding flow assurance in the area of gas hydrates. The effort to understand the complex behavior of gas hydrates in multiphase flow has resulted in new hydrate blockage models. Although the hydrate community continues to debate the impact of kinetics, agglomeration, and oil chemistry effects on hydrate blockage formation in pipelines and wellbores, the petroleum industry still relies on thermodynamic strategies that completely prevent hydrates in production systems. However, these complex strategies such as thermal insulation, electric heating, dead oil displacement, and methanol injection are costly, particularly for marginal fields. As such, research continues in developing a comprehensive multiphase flow simulator capable of handling the transient aspects of production operations, notably shut-in, restart, blowdown and blockage prediction. Model predictions are leading to new operating strategies based on risk management approach. This paper discussed the challenges and opportunities that have shifted the focus from prevention of hydrates to prevention of blockage. Some initial successes in the development of a first generation empirical tool for the prediction of hydrate blockages in flow lines were also presented along with new experimental data that explained how hydrate blockages can manifest in the field. It was concluded that additional research is needed to solve the problem of hydrate plugging mechanism. 12 refs., 6 figs.

  19. An analysis of factors causing the occurrence of off-design thermally induced force effects in the zone of weld joint no. 111-1 in a PGV-1000M steam generator and recommendations on excluding them

    Science.gov (United States)

    Bakirov, M. B.; Levchuk, V. I.; Povarov, V. P.; Gromov, A. F.

    2014-08-01

    Inadmissible operational flaws occurring in the critical zones of heat-transfer and mechanical equipment are commonly revealed in all nuclear power plant units both in Russia and abroad. The number of such flaws will only grow in the future because the majority of nuclear power plants have been in operation for a time that is either close to or even exceeds the assigned service life. In this connection, establishing cause-and-effect relations with regard to accelerated incipience and growth of flaws, working out compensating measures aimed at reducing operational damageability, and setting up monitoring of equipment integrity degradation of during operation are becoming the matters of utmost importance. There is a need to introduce new approaches to comprehensive diagnostics of the technical state of important nuclear power plant equipment, including continuous monitoring of its operational damageability and the extent of its loading in the most critical zones. Starting from 2011, such a monitoring system has successfully been used for the Novovoronezh NPP Unit 5 in the zone of weld joint no. 111-1 of steam generator no. 4. Based on the results from operation of this system in 2011-2013, unsteady thermally induced force effects (periodic thermal shocks and temperature abnormalities) were reveled, which had not been considered in the design, and which have an essential influence on the operational loading of this part. Based on an analysis of cause-and-effect relations pertinent to temperature abnormalities connected with technological operations, a set of measures aimed at reducing the thermally induced force loads exerted on pipeline sections was developed, which includes corrections to the process regulations for safe operation and to the operating manuals (involving changes in the algorithms for manipulating with the stop and control valves in the steam generator blowdown system).

  20. Fluiddynamic effects in the fuel element top nozzle area during refilling and reflooding

    International Nuclear Information System (INIS)

    Hawighorst, A.; Kroening, H.; Mewes, D.; Spatz, R.; Mayinger, F.

    1985-01-01

    During the refilling and reflooding phase following a hypothetical loss of coolant accident in lightwater cooled nuclear reactors, there will be countercurrent flow between discharging steam and the feed of emergency core cooling water. It was the objective of this research project to contribute to a better physical understanding of the fluiddynamic processes in the area of the fuel element top nozzle and so to improve emergency core cooling calculations. Therefore, experimental and theoretical investigations about the entrainment and countercurrent behaviour of gas/liquid flows have been implemented within this project. Fluiddynamic processes in the fuel element top nozzle area were simulated during the reflooding and refilling phase. Based on special internals as single and multiple-hole orifices, basic phenomena of fluidynamics were studied first with air-water. Subsequently, investigations of the system steam/water were conducted. The reactor geometry was approximated step by step, until a complete reactor fuel assembly top nozzle was constituted. The system pressure was 4.8 bars (abs), in accordance with the conditions in the reactor pressure vessel at the end of the blowdown phase. The water was initially fed in at saturation temperature, then, as a second step, fed in at subcooled condition relative to the steam temperature, in order to be able to study condensation effects as well. First, investigations on gas/liquid countercurrent flows in the fluid system air/water are presented. Then one studies countercurrent flow in the system steam/water, including the investigation of condensation effects. Finally, a detailed description of the research on droplet size determination is given

  1. PWR steam generator chemical cleaning. Phase I: solvent and process development. Volume II

    International Nuclear Information System (INIS)

    Larrick, A.P.; Paasch, R.A.; Hall, T.M.; Schneidmiller, D.

    1979-01-01

    A program to demonstrate chemical cleaning methods for removing magnetite corrosion products from the annuli between steam generator tubes and the tube support plates in vertical U-tube steam generators is described. These corrosion products have caused steam generator tube ''denting'' and in some cases have caused tube failures and support plate cracking in several PWR generating plants. Laboratory studies were performed to develop a chemical cleaning solvent and application process for demonstration cleaning of the Indian Point Unit 2 steam generators. The chemical cleaning solvent and application process were successfully pilot-tested by cleaning the secondary side of one of the Indian Point Unit 1 steam generators. Although the Indian Point Unit 1 steam generators do not have a tube denting problem, the pilot test provided for testing of the solvent and process using much of the same equipment and facilities that would be used for the Indian Point Unit 2 demonstration cleaning. The chemical solvent selected for the pilot test was an inhibited 3% citric acid-3% ascorbic acid solution. The application process, injection into the steam generator through the boiler blowdown system and agitation by nitrogen sparging, was tested in a nuclear environment and with corrosion products formed during years of steam generator operation at power. The test demonstrated that the magnetite corrosion products in simulated tube-to-tube support plate annuli can be removed by chemical cleaning; that corrosion resulting from the cleaning is not excessive; and that steam generator cleaning can be accomplished with acceptable levels of radiation exposure to personnel

  2. Development Testing of 1-Newton ADN-Based Rocket Engines

    Science.gov (United States)

    Anflo, K.; Gronland, T.-A.; Bergman, G.; Nedar, R.; Thormählen, P.

    2004-10-01

    With the objective to reduce operational hazards and improve specific and density impulse as compared with hydrazine, the Research and Development (R&D) of a new monopropellant for space applications based on AmmoniumDiNitramide (ADN), was first proposed in 1997. This pioneering work has been described in previous papers1,2,3,4 . From the discussion above, it is clear that cost savings as well as risk reduction are the main drivers to develop a new generation of reduced hazard propellants. However, this alone is not enough to convince a spacecraft builder to choose a new technology. Cost, risk and schedule reduction are good incentives, but a spacecraft supplier will ask for evidence that this new propulsion system meets a number of requirements within the following areas: This paper describes the ongoing effort to develop a storable liquid monopropellant blend, based on AND, and its specific rocket engines. After building and testing more than 20 experimental rocket engines, the first Engineering Model (EM-1) has now accumulated more than 1 hour of firing-time. The results from test firings have validated the design. Specific impulse, combustion stability, blow-down capability and short pulse capability are amongst the requirements that have been demonstrated. The LMP-103x propellant candidate has been stored for more than 1 year and initial material compatibility screening and testing has started. 1. Performance &life 2. Impact on spacecraft design &operation 3. Flight heritage Hereafter, the essential requirements for some of these areas are outlined. These issues are discussed in detail in a previous paper1 . The use of "Commercial Of The Shelf" (COTS) propulsion system components as much as possible is essential to minimize the overall cost, risk and schedule. This leads to the conclusion that the Technology Readiness Level (TRL) 5 has been reached for the thruster and propellant. Furthermore, that the concept of ADN-based propulsion is feasible.

  3. A fuzzy approach for modelling radionuclide in lake system

    International Nuclear Information System (INIS)

    Desai, H.K.; Christian, R.A.; Banerjee, J.; Patra, A.K.

    2013-01-01

    Radioactive liquid waste is generated during operation and maintenance of Pressurised Heavy Water Reactors (PHWRs). Generally low level liquid waste is diluted and then discharged into the near by water-body through blowdown water discharge line as per the standard waste management practice. The effluents from nuclear installations are treated adequately and then released in a controlled manner under strict compliance of discharge criteria. An attempt was made to predict the concentration of 3 H released from Kakrapar Atomic Power Station at Ratania Regulator, about 2.5 km away from the discharge point, where human exposure is expected. Scarcity of data and complex geometry of the lake prompted the use of Heuristic approach. Under this condition, Fuzzy rule based approach was adopted to develop a model, which could predict 3 H concentration at Ratania Regulator. Three hundred data were generated for developing the fuzzy rules, in which input parameters were water flow from lake and 3 H concentration at discharge point. The Output was 3 H concentration at Ratania Regulator. These data points were generated by multiple regression analysis of the original data. Again by using same methodology hundred data were generated for the validation of the model, which were compared against the predicted output generated by using Fuzzy Rule based approach. Root Mean Square Error of the model came out to be 1.95, which showed good agreement by Fuzzy model of natural ecosystem. -- Highlights: • Uncommon approach (Fuzzy Rule Base) of modelling radionuclide dispersion in Lake. • Predicts 3 H released from Kakrapar Atomic Power Station at a point of human exposure. • RMSE of fuzzy model is 1.95, which means, it has well imitated natural ecosystem

  4. Safety evaluation of the loss of fluid test facility project No. 394

    International Nuclear Information System (INIS)

    1975-05-01

    Assessment of the safety of the LOFT facility and subsequent recommendations have been based on a comparison of the LOFT facility to requirements for commercial power reactors. In this comparison, the many unique features of the LOFT facility were considered including the low power level, the limited operational use as a test reactor, and the remoteness of the site. Based on this assessment, it is concluded, that while the likelihood of an accidental release of fission products may be greater than for a commercial power reactor, the consequences of such a release are reduced by the lower fission product inventory, the remoteness of the site and the capability of evacuating the Idaho National Engineering Laboratory (INEL) and adjacent areas. There is reasonable assurance that the public health and safety will not be endangered due to operation of this facility, specifically: The INEL site is acceptable with respect to location, land use, population distribution, controlled access, hydrology, meteorology, geology and seismology. Sufficient engineered safety features have been included to assure that the potential offsite doses are well within 10 CFR Part 100 guidelines. The LOFT facility has been designed in general accordance with standards, guides and codes which are comparable to those applied to commercial power reactors and any exceptions to these have been based on the unique features of the LOFT facility. Certain matters including the final safety analyses based on detailed component designs, Technical Specifications, LOCE controls and detailed program plan have not been reviewed but we assume will properly be resolved by ERDA, which has the ultimate responsibility for the safety of this facility. Changes to the facility design or program plan such as removal of the fueled Mobile Test Assembly or blowdowns to the containment vessel also will require additional analyses and review. (U.S.)

  5. US nuclear safety. Review and experience

    International Nuclear Information System (INIS)

    Hanauer, S.H.

    1977-01-01

    The paper deals with the evolution of reactor safety principles, design bases, regulatory requirements, and experience in the United States. Safety concerns have evolved over the years, from reactivity transients and shut-down systems, to blowdowns and containment, to severe design basis accidents and mitigating systems, to the performance of actual materials, systems and humans. The primary safety concerns of one epoch have been superseded in considerable measure by those of later times. Successive plateaus of technical understanding are achieved by solutions being found to earlier problems. Design studies, research, operating experience and regulatory imperatives all contribute to the increased understanding and thus to the safety improvements adopted and accepted. The improvement of safety with time, and the ability of existing reactors to operate safely in the face of new concerns, has confirmed the correctness and usefulness of the defence-in-depth approach and safety margins used in safety design in the United States of America. A regulatory programme such as the one in the United States justifies its great cost by its important contributions to safety. Yet only the designers, constructors and operators of nuclear power plants can actually achieve public safety. The regulatory programme audits, assesses and spot-checks the actual work. Since neither materials nor human beings are flawless, mistakes will be made; that is why defence-in-depth and safety margins are provided. The regulatory programme should enhance safety by decreasing the frequency of uncorrected mistakes. Maintenance of public safety also requires technical and managerial competence and attention in the organizations responsible for nuclear plants as well as regulatory organizations. (author)

  6. Biological assessments for the low energy demonstration accelerator, 1996 and 1997

    Energy Technology Data Exchange (ETDEWEB)

    Cross, S.

    1998-12-31

    The Department of Energy (DOE) plans to build, install, and operate a Low Energy Demonstration Accelerator (LMA) in Technical Area 53 of the Los Alamos National Laboratory (LANL). LEDA will demonstrate the accelerator technology necessary to produce tritium, but is not designed to produce tritium at LANL. USFWS reviewers of the Biological Assessment prepared for LEDA insisted that the main drainage be monitored to measure and document changes to vegetation, soils, wildlife, and habitats due to LEDA effluent discharges. The Biology Team of ESH-20 (LANL`s Ecology Group) has performed these monitoring activities during 1996 and 1997 to document baseline conditions before LEDA released significant effluent discharges. Quarterly monitoring of the outfall which will discharge LEDA blowdown effluent had one exceedance of permitted parameters, a high chlorine discharge that was quickly remedied. Samples from 12 soil pits in the drainage area contained no hydric indicators, such as organic matter in the upper layers, streaking, organic pans, and oxidized rhizospheres. Vegetation transacts in the meadows that LEDA discharges will flow through contained 44 species of herbaceous plants, all upland taxa. Surveys of resident birds, reptiles, and amphibians documented a fauna typical of local dry canyons. No threatened or endangered species inhabit the project area, but increased effluent releases may make the area more attractive to many wildlife species, an endangered raptor, and several other species of concern. Biological best management practices especially designed for LEDA are discussed, including protection of floodplains, erosion control measures, hazards posed by increased usage of the area by deer and elk and revegetation of disturbed areas.

  7. Development and application of a dual RELAP5-3D-based engineering simulator for ABWR

    International Nuclear Information System (INIS)

    Yang, C.-Y.; Liang, Thomas K.S.; Pei, B.S.; Shih, C.K.; Chiang, S.C.; Wang, L.C.

    2009-01-01

    For any innovated plant design, the designed paper plant can be converted into a computer as a digital plant with advanced simulation techniques before being constructed into a real plant. A digital plant, namely engineering simulator, can be applied for: (1) verification of system design and system integration, (2) power test simulation, (3) plant transient and accident analyses, (4) plant abnormal and emergency procedure development and verification, (5) design change verification and analysis, etc. An advanced engineering simulator was successfully developed for the LungMen advanced boiling water reactor (ABWR) plant to support various applications before and after commercial operation. This plant specific engineering simulator was developed based on two separate RELAP5-3D modules synchronized on a commercial simulation platform, namely 3-Key Master. On this advanced LungMen plant simulation (ALPS) platform, major plant dynamics were simulated by two separate RELAP5-3D modules, one for reactor system modeling and the other for balance of plant (BOP) system modeling. Moreover, major control systems as well as emergency core cooling system (ECCS) were all simulated in great detail with built-in tasks of this commercial simulation platform. Different from real time calculation on training simulator, precision of engineering calculation is intentionally kept by synchronizing modules based on the most time-consuming one. During synchronization, each module will check its' own converge criteria in each small time advancement. This plant specific advanced ABWR engineering simulator has been successfully applied on: (1) licensing blowdown analysis of feed water line break (FWLB) for containment design; (2) phenomena investigation of low-pressure ECC injection bypass during FWLB; (3) analysis of FW pump performance during power ascending; (4) verification of plant vendor's pre-test calculations of each start-up test.

  8. Characterization of In-Drum Drying Products

    International Nuclear Information System (INIS)

    Kroselj, V.; Jankovic, M.; Skanata, D.; Medakovic, S.; Harapin, D.; Hertl, B.

    2006-01-01

    A few years ago Krsko NPP decided to introduce In-Drum Drying technology for treatment and conditioning of evaporator concentrates and spent ion resins. The main reason to employ this technology was the need for waste volume reduction and experience with vermiculite-cement solidification that proved inadequate for Krsko NPP. Use of In-Drum Drying technology was encouraged by good experience in the field at some German and Spanish NPP's. In the paper, solidification techniques in vermiculite-cement matrix and In-Drum Drying System are described briefly. The resulting waste forms (so called solidification and dryer products) and containers that are used for interim storage of these wastes are described as well. A comparison of the drying versus solidification technology is performed and advantages as well as disadvantages are underlined. Experience gained during seven years of system operation has shown that crying technology resulted in volume reduction by factor of 20 for evaporator concentrates, and by factor of 5 for spent ion resin. Special consideration is paid to the characterization of dryer products. For evaporator concentrates the resulting waste form is a solid salt block with up to 5% bound water. It is packaged in stainless steel drums (net volume of 200 l) with bolted lids and lifting rings. The fluidized spent ion resins (primary and blow-down) are sluiced into the spent resin drying tank. The resin is dewatered and dried by electrical jacket heaters. The resulting waste (i.e. fine granulates) is directly discharged into a shielded stainless steel drum with bolted lid and lifting rings. Characterization of both waste forms has been performed in accordance with recommendations given in Characterization of Radioactive Waste Forms and Packages issued by International Atomic Energy Agency, 1997. This means that radiological, chemical, physical, mechanical, biological and thermal properties of the waste form has been taken into consideration. In the paper

  9. Assessment of the Polyacrylic Acid for an Ammonia Water Treatment and for Alloy 800NG SG Tube Material in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Lamouroux, Christine; You, Dominique; Plancque, Gabriel; Roy, Marc; Laire, Charles; Schnongs, Philippe

    2012-09-01

    To prevent the Steam Generators (SG) fouling by corrosion products or the Tube Support Plate (TSP) blockage the on-line injection of a dispersant such the Polyacrylic Acid (PAA) could be a relevant water treatment. Long-term trials performed in PWRs have shown that the PAA, injected at the SG inlet, facilitate the evacuation of the iron oxides by the SG blowdown. Given the ammonia treatment of the secondary water of the Belgian PWRs, the R and D program carried out was devoted to: - Verify the innocuousness of the PAA and its degradation products versus Alloy 800NG SCC susceptibility in case of over concentrations and sludge presence, - Assess the potential impact of the PAA and its thermal degradation products on the specific NH 3 water treatment. The main results can be summarized as following: The corrosion tests performed with PAA in case of over concentrations and sludge couldn't point out any negative effect of the dispersant on the SCC susceptibility of tubing materials such as Alloy 800NG. No significant modification of the tube oxide layer has been observed. At the SG operating temperature, the PAA is decomposed and a large spectrum from high to lower molecular weights polymers than the initial PAA arises. The fragmentation of the polymer into low molecular weight polyacrylic acids is obtained within 20 minutes and the average molecular weight is reduced by 50% from the original one. The thermal degradation products, their quantity and their kinetic of appearance, have been determined. The generated acetate concentration during the on-line dispersant application should remain low compared to the current values observed in the SG water. From the numerical simulation based on acetate concentration and on the kinetic law deduced from the experimental work, it can be concluded that in a 2-phase medium, the margin on the water pH compared to the neutral pH remains high. At 180 deg. C, no impact on the water pH is identified, taking into account realistic

  10. Chemical preventive remedies for steam generators fouling and tube support plate blockages

    International Nuclear Information System (INIS)

    Alves Vieira, M.; Mayos, M.; Coquio, N.; Fourcroy, H.; Battesti, P.

    2010-01-01

    In 2006, EDF identified on several PWR units broached hole blockage on the upper Steam Generator (SG) Tube Support Plates (TSP). TSP blockage often occurs in association with secondary fouling. The units with copper alloys materials are more affected due the applied low pH 25 o C (9.20) all volatile treatment (AVT). Carbon steels materials are less protected against flow accelerated corrosion (FAC) and therefore more corrosion products enter the SGs through the final feed water (FFW). In parallel of chemical cleanings to remove oxides deposits in SGs, EDF has defined a strategy to improve operating conditions. It mainly relies on the removal of copper alloys materials to implement a high pH AVT (9.60) as a preventive remedy. However for some plants, copper alloys removal is not straightforward due to environmental constraints. EDF must indeed manage the implementation of a biocide treatment needed in closed loop cooling systems (as copper has a bacteriostatic effect on micro-organisms) and more generally must comply with discharge authorisations for chemical conditioning reagents or biocide reagent. An alternative conditioning was tested on the Dampierre 4 unit in 2007/2008 during 6 months to assess if operating at 9.40 was acceptable regarding the impacts on copper alloys materials. The perspective would be to implement it in the units where no biocide treatment can be applied on a short term. In parallel, other chemical conditionings or additives will be implemented or tested. First of all, EDF will carry out a trial test with APA in order to assess its efficiency on the removal of oxides deposits through SG blowdown. On the other hand, AVT with high pH ethanolamine (ETA) will be implemented as an alternative of ammonia and morpholine conditioning on some chosen plants. Ethanolamine is selected as a way to mitigate FAC kinetics in two-phase flow areas (reheaters or moisture heater separator) or to limit liquid releases. This paper provides the lessons of the

  11. Using Land Surface Phenology as the Basis for a National Early Warning System for Forest Disturbances

    Science.gov (United States)

    Hargrove, W. W.; Spruce, J.; Norman, S. P.; Hoffman, F. M.

    2011-12-01

    The National Early Warning System (EWS) provides an 8-day coast-to-coast snapshot of potentially disturbed forests across the U.S.. A prototype system has produced national maps of potential forest disturbances every eight days since January 2010, identifying locations that may require further investigation. Through phenology, the system shows both early and delayed vegetation development and detects all types of unexpected forest disturbances, including insects, disease, wildfires, frost and ice damage, tornadoes, hurricanes, blowdowns, harvest, urbanization, landslides, drought, flood, and climate change. The USDA Forest Service Eastern Forest Environmental Threat Assessment Center is collaborating with NASA Stennis Space Center and the Western Wildland Environmental Threat Assessment Center to develop the tool. The EWS uses differences in phenological responses between an expectation based on historical data and a current view to strategically identify potential forest disturbances and direct attention to locations where forest behavior seems unusual. Disturbance maps are available via the Forest Change Assessment Viewer (FCAV) (http://ews.forestthreats.org/gis), which allows resource managers and other users to see the most current national disturbance maps as soon as they are available. Phenology-based detections show not only vegetation disturbances in the classical sense, but all departures from normal seasonal vegetation behavior. In 2010, the EWS detected a repeated late-frost event at high elevations in North Carolina, USA, that resulted in delayed seasonal development, contrasting with an early spring development at lower elevations, all within close geographic proximity. Throughout 2011, there was a high degree of correspondence between the National Climatic Data Center's North American Drought Monitor maps and EWS maps of phenological drought disturbance in forests. Urban forests showed earlier and more severe phenological drought disturbance than

  12. Improved estimates of filtered total mercury loadings and total mercury concentrations of solids from potential sources to Sinclair Inlet, Kitsap County, Washington

    Science.gov (United States)

    Paulson, Anthony J.; Conn, Kathleen E.; DeWild, John F.

    2013-01-01

    Previous investigations examined sources and sinks of mercury to Sinclair Inlet based on historic and new data. This included an evaluation of mercury concentrations from various sources and mercury loadings from industrial discharges and groundwater flowing from the Bremerton naval complex to Sinclair Inlet. This report provides new data from four potential sources of mercury to Sinclair Inlet: (1) filtered and particulate total mercury concentrations of creek water during the wet season, (2) filtered and particulate total mercury releases from the Navy steam plant following changes in the water softening process and discharge operations, (3) release of mercury from soils to groundwater in two landfill areas at the Bremerton naval complex, and (4) total mercury concentrations of solids in dry dock sumps that were not affected by bias from sequential sampling. The previous estimate of the loading of filtered total mercury from Sinclair Inlet creeks was based solely on dry season samples. Concentrations of filtered total mercury in creek samples collected during wet weather were significantly higher than dry weather concentrations, which increased the estimated loading of filtered total mercury from creek basins from 27.1 to 78.1 grams per year. Changes in the concentrations and loading of filtered and particulate total mercury in the effluent of the steam plant were investigated after the water softening process was changed from ion-exchange to reverse osmosis and the discharge of stack blow-down wash began to be diverted to the municipal water-treatment plant. These changes reduced the concentrations of filtered and particulate total mercury from the steam plant of the Bremerton naval complex, which resulted in reduced loadings of filtered total mercury from 5.9 to 0.15 grams per year. Previous investigations identified three fill areas on the Bremerton naval complex, of which the western fill area is thought to be the largest source of mercury on the base

  13. Optical Property Retention Methods for the T-170M Space Telescope Mirrors Surface in the Project «Spektr-UF» at the Preflight Preparation Stage

    Directory of Open Access Journals (Sweden)

    F L. Chubarov

    2017-01-01

    ISO 14644-1-2002; contamination tolerance of optical system components being, at most, 0.03% of the total surface area (300 p.p.m. is provided by the rational choice of the blow-down scheme of the telescope with feeding nitrogen through the equipment bay;4    a rational design of the supercharger using a gas analyzer and boost valves with external feedback provides driving pressure and operation autonomy of the system that maintains atmosphere control.

  14. Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Liang-Che, E-mail: lcdai@iner.gov.tw; Chen, Yen-Shu; Yuann, Yng-Ruey

    2014-12-15

    Highlights: • The GOTHIC code is used for the PWR dry containment pressure and temperature analysis. • Boundary conditions are hot standby and 102% power main steam line break accidents. • Containment pressure and temperature responses of GOTHIC are similar with FSAR. • The capability of the developed model to perform licensing calculation is assessed. - Abstract: Units 1 and 2 of the Maanshan nuclear power station are the typical Westinghouse three-loop PWR (pressurized water reactor) with large dry containments. In this study, the containment analysis program GOTHIC is adopted for the dry containment pressure and temperature analysis. Free air space and sump of the PWR dry containment are individually modeled as control volumes. The containment spray system and fan cooler unit are also considered in the GOTHIC model. The blowdown mass and energy data of the main steam line break (hot standby condition and various reactor thermal power levels) are tabulated in the Maanshan Final Safety Analysis Report (FSAR) 6.2 which could be used as the boundary conditions for the containment model. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. In this study, hot standby and 102% reactor thermal power main steam line break accidents are selected. The calculated peak containment pressure is 323.50 kPag (46.92 psig) for hot standby MSLB, which is a little higher than the FSAR value of 311.92 kPag (45.24 psig). But it is still below the design value of 413.69 kPag (60 psig). The calculated peak vapor temperature inside the containment is 187.0 °C (368.59 F) for 102% reactor thermal power MSLB, which is lower than the FSAR result of 194.42 °C (381.95 F). The effects of the containment spray system and fan cooler units could be clearly observed in the GOTHIC analysis. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR

  15. Direct torus venting analysis for Chinshan BWR-4 plant with MARK-I containment

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2017-03-15

    Highlights: • Study the effectiveness of Direct Torus Venting System (DTVS) during extended SBO of 24 h for Chinshan MARK-I plant. • Containment response is analyzed by GOTHIC based on boundary conditions from RETRAN calculation. • Analyses are performed with and without DTVS, respectively. • Suppression pool is sub-divided and thermal stratification is observed. - Abstract: The Chinshan plant, owned by Taiwan Power Company, has twin units of BWR-4 reactor and MARK-I containment. Both units have been operating at rated core thermal power of 1840 MWt. The existing Direct Torus Venting System (DTVS) is the main system used for venting the containment during the extended station blackout event. The purpose of this paper is to study the effects of the DTVS venting on the response of the containment pressure and temperature. The reactor is depressurized by manually opening the safety relief valves (SRVs) during the SBO, which causes the mass and energy to be discharged into and heat up the suppression pool. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The DTVS model is established in the GOTHIC model based on the venting size, venting piping loss, venting initiation time, and venting source. The lumped volume model, 1-D coarse-mesh model, and 3-D coarse-mesh model are considered in the torus volume. The calculation is first done without DTVS venting to establish a reference basis. Then a case with DTVS available is performed. Comparison of the two cases shows that the existing DTVS design is effective in mitigating the severity of the containment pressure and temperature transients. The results also show that the 1-D coarse-mesh model may not be appropriate since a

  16. Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

    International Nuclear Information System (INIS)

    Dai, Liang-Che; Chen, Yen-Shu; Yuann, Yng-Ruey

    2014-01-01

    Highlights: • The GOTHIC code is used for the PWR dry containment pressure and temperature analysis. • Boundary conditions are hot standby and 102% power main steam line break accidents. • Containment pressure and temperature responses of GOTHIC are similar with FSAR. • The capability of the developed model to perform licensing calculation is assessed. - Abstract: Units 1 and 2 of the Maanshan nuclear power station are the typical Westinghouse three-loop PWR (pressurized water reactor) with large dry containments. In this study, the containment analysis program GOTHIC is adopted for the dry containment pressure and temperature analysis. Free air space and sump of the PWR dry containment are individually modeled as control volumes. The containment spray system and fan cooler unit are also considered in the GOTHIC model. The blowdown mass and energy data of the main steam line break (hot standby condition and various reactor thermal power levels) are tabulated in the Maanshan Final Safety Analysis Report (FSAR) 6.2 which could be used as the boundary conditions for the containment model. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. In this study, hot standby and 102% reactor thermal power main steam line break accidents are selected. The calculated peak containment pressure is 323.50 kPag (46.92 psig) for hot standby MSLB, which is a little higher than the FSAR value of 311.92 kPag (45.24 psig). But it is still below the design value of 413.69 kPag (60 psig). The calculated peak vapor temperature inside the containment is 187.0 °C (368.59 F) for 102% reactor thermal power MSLB, which is lower than the FSAR result of 194.42 °C (381.95 F). The effects of the containment spray system and fan cooler units could be clearly observed in the GOTHIC analysis. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR

  17. ForWarn Forest Disturbance Change Detection System Provides a Weekly Snapshot of US Forest Conditions to Aid Forest Managers

    Science.gov (United States)

    Hargrove, W. W.; Spruce, J.; Kumar, J.; Hoffman, F. M.

    2012-12-01

    The Eastern Forest Environmental Threat Assessment Center and Western Wildland Environmental Assessment Center of the USDA Forest Service have collaborated with NASA Stennis Space Center to develop ForWarn, a forest monitoring tool that uses MODIS satellite imagery to produce weekly snapshots of vegetation conditions across the lower 48 United States. Forest and natural resource managers can use ForWarn to rapidly detect, identify, and respond to unexpected changes in the nation's forests caused by insects, diseases, wildfires, severe weather, or other natural or human-caused events. ForWarn detects most types of forest disturbances, including insects, disease, wildfires, frost and ice damage, tornadoes, hurricanes, blowdowns, harvest, urbanization, and landslides. It also detects drought, flood, and temperature effects, and shows early and delayed seasonal vegetation development. Operating continuously since January 2010, results show ForWarn to be a robust and highly capable tool for detecting changes in forest conditions. To help forest and natural resource managers rapidly detect, identify, and respond to unexpected changes in the nation's forests, ForWarn produces sets of national maps showing potential forest disturbances at 231m resolution every 8 days, and posts the results to the web for examination. ForWarn compares current greenness with the "normal," historically seen greenness that would be expected for healthy vegetation for a specific location and time of the year, and then identifies areas appearing less green than expected to provide a strategic national overview of potential forest disturbances that can be used to direct ground and aircraft efforts. In addition to forests, ForWarn also tracks potential disturbances in rangeland vegetation and agriculural crops. ForWarn is the first national-scale system of its kind based on remote sensing developed specifically for forest disturbances. The ForWarn system had an official unveiling and rollout in

  18. Free-flight measurement technique in the free-piston high-enthalpy shock tunnel.

    Science.gov (United States)

    Tanno, H; Komuro, T; Sato, K; Fujita, K; Laurence, S J

    2014-04-01

    A novel multi-component force-measurement technique has been developed and implemented at the impulse facility JAXA-HIEST, in which the test model is completely unrestrained during the test and thus experiences free-flight conditions for a period on the order of milliseconds. Advantages over conventional free-flight techniques include the complete absence of aerodynamic interference from a model support system and less variation in model position and attitude during the test itself. A miniature on-board data recorder, which was a key technology for this technique, was also developed in order to acquire and store the measured data. The technique was demonstrated in a HIEST wind-tunnel test campaign in which three-component aerodynamic force measurement was performed on a blunted cone of length 316 mm, total mass 19.75 kg, and moment of inertia 0.152 kgm(2). During the test campaign, axial force, normal forces, and pitching moment coefficients were obtained at angles of attack from 14° to 32° under two conditions: H0 = 4 MJ/kg, P0 = 14 MPa; and H0 = 16 MJ/kg, P0 = 16 MPa. For the first, low-enthalpy condition, the test flow was considered a perfect gas; measurements were thus directly compared with those obtained in a conventional blow-down wind tunnel (JAXA-HWT2) to evaluate the accuracy of the technique. The second test condition was a high-enthalpy condition in which 85% of the oxygen molecules were expected to be dissociated; high-temperature real-gas effects were therefore evaluated by comparison with results obtained in perfect-gas conditions. The precision of the present measurements was evaluated through an uncertainty analysis, which showed the aerodynamic coefficients in the HIEST low enthalpy test agreeing well with those of JAXA-HWT2. The pitching-moment coefficient, however, showed significant differences between low- and high-enthalpy tests. These differences are thought to result from high-temperature real-gas effects.

  19. Characterization Investigation Study: Volume 3, Radiological survey of surface soils

    Energy Technology Data Exchange (ETDEWEB)

    Solow, A.J.; Phoenix, D.R.

    1987-12-01

    The Feed Materials Production Center was constructed to produce high purity uranium metal for use at various Department of Energy facilities. The waste products from these operations include general uncontaminated scrap and refuse, contaminated and uncontaminated metal scrap, waste oils, low-level radioactive waste, co-contaminated wastes, mixed waste, toxic waste, sludges from water treatment, and fly ash from the steam plant. This material is estimated to total more than 350,000 cubic meters. Other wastes stored in this area include laboratory chemicals and other combustible materials in the burn pit; fine waste stream sediments in the clear well; fly ash and waste oils in the two fly ash areas; lime-alum sludges and boiler plant blowdown in the lime sludge ponds; and nonradioactive sanitary waste, construction rubble, and asbestos in the sanitary landfill. A systematic survey of the surface soils throughout the Waste Storage Area, associated on-site drainages, and the fly ash piles was conducted using a Field Instrument for Detecting Low-Energy Radiation (FIDLER). Uranium is the most prevalent radioactive element in surface soil; U-238 is the principal radionuclide, ranging from 2.2 to 1790 pCi/g in the general Waste Storage Area. The maximum values for the next highest activity concentrations in the same area were 972 pCi/g for Th-230 and 298 pCi/g for U-234. Elevated activity concentrations of Th-230 were found along the K-65 slurry line, the maximum at 3010 pCi/g. U-238 had the highest value of 761 pCi/g in the drainage just south of pit no. 5. The upper fly ash area had the highest radionuclide activity concentrations in the surface soils with the maximum values for U-238 at 8600 pCi/g, U-235 at 2190 pCi/g, U-234 at 11,400 pCi/g, Tc-99 at 594 pCi/g, Ra-226 at 279 pCi/g, and Th-230 at 164 pCi/g.

  20. Application of Pulsed Electrical Fields for Advanced Cooling and Water Recovery in Coal-Fired Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Young Cho; Alexander Fridman

    2009-04-02

    The overall objective of the present work was to develop technologies to reduce freshwater consumption in a cooling tower of coal-based power plant so that one could significantly reduce the need of make-up water. The specific goal was to develop a scale prevention technology based an integrated system of physical water treatment (PWT) and a novel filtration method so that one could reduce the need for the water blowdown, which accounts approximately 30% of water loss in a cooling tower. The present study investigated if a pulsed spark discharge in water could be used to remove deposits from the filter membrane. The test setup included a circulating water loop and a pulsed power system. The present experiments used artificially hardened water with hardness of 1,000 mg/L of CaCO{sub 3} made from a mixture of calcium chloride (CaCl{sub 2}) and sodium carbonate (Na{sub 2}CO{sub 3}) in order to produce calcium carbonate deposits on the filter membrane. Spark discharge in water was found to produce strong shockwaves in water, and the efficiency of the spark discharge in cleaning filter surface was evaluated by measuring the pressure drop across the filter over time. Results showed that the pressure drop could be reduced to the value corresponding to the initial clean state and after that the filter could be maintained at the initial state almost indefinitely, confirming the validity of the present concept of pulsed spark discharge in water to clean dirty filter. The present study also investigated the effect of a plasma-assisted self-cleaning filter on the performance of physical water treatment (PWT) solenoid coil for the mitigation of mineral fouling in a concentric counterflow heat exchanger. The self-cleaning filter utilized shockwaves produced by pulse-spark discharges in water to continuously remove scale deposits from the surface of the filter, thus keeping the pressure drop across the filter at a relatively low value. Artificial hard water was used in the

  1. Twenty-Five Years of Ecological Recovery of East Fork Poplar Creek: Review of Environmental Problems and Remedial Actions

    Science.gov (United States)

    Loar, James M.; Stewart, Arthur J.; Smith, John G.

    2011-06-01

    In May 1985, a National Pollutant Discharge Elimination System permit was issued for the Department of Energy's Y-12 National Security Complex (Y-12 Complex) in Oak Ridge, Tennessee, USA, allowing discharge of effluents to East Fork Poplar Creek (EFPC). The effluents ranged from large volumes of chlorinated once-through cooling water and cooling tower blow-down to smaller discharges of treated and untreated process wastewaters, which contained a mixture of heavy metals, organics, and nutrients, especially nitrates. As a condition of the permit, a Biological Monitoring and Abatement Program (BMAP) was developed to meet two major objectives: demonstrate that the established effluent limitations were protecting the classified uses of EFPC, and document the ecological effects resulting from implementing a Water Pollution Control Program at the Y-12 Complex. The second objective is the primary focus of the other papers in this special series. This paper provides a history of pollution and the remedial actions that were implemented; describes the geographic setting of the study area; and characterizes the physicochemical attributes of the sampling sites, including changes in stream flow and temperature that occurred during implementation of the BMAP. Most of the actions taken under the Water Pollution Control Program were completed between 1986 and 1998, with as many as four years elapsing between some of the most significant actions. The Water Pollution Control Program included constructing nine new wastewater treatment facilities and implementation of several other pollution-reducing measures, such as a best management practices plan; area-source pollution control management; and various spill-prevention projects. Many of the major actions had readily discernable effects on the chemical and physical conditions of EFPC. As controls on effluents entering the stream were implemented, pollutant concentrations generally declined and, at least initially, the volume of water

  2. K-FIX(3D), Transient 2 Phase Flow Hydrodynamic, X-Y-Z and Cylindrical Geometry, Eulerian Method

    International Nuclear Information System (INIS)

    Rivard, W.C.; Torrey, MD.

    1982-01-01

    1 - Description of problem or function: This package consists of two programs K-FIX(3D,FLX) which extend the transient, two-dimensional, two fluid program K-FIX (NESC Abstract 727) to perform three- dimensional calculations. The transient dynamics of three- dimensional, two-phase flow with interfacial exchange are calculated at all flow speeds. Each phase is described in terms of its own density, velocity, and temperature. The application is to flow in the annulus between two cylinders where the inner cylinder moves periodically perpendicular to its axis. K-FIX(3D) is easily adaptable to a variety of two phase flow problems while K-FIX(3D,FLX) combines KFIX(3D), the three- dimensional version of the KFIX code, with the three-dimensional, elastic shell code FLX for application to a very specific class of problems. KFIX(3D,FLX) was developed specifically to calculate the coupled fluid structure dynamics of a light water reactor core support barrel under accident conditions. Motion may be induced by blowdown, prescribed displacement, or seismic action. 2 - Method of solution: In the K-FIX(3D), the six field equations for the two phases couple through mass, momentum, and energy exchange. The equations are solved using an Eulerian finite-difference technique that implicitly couples the rates of phase transitions, momentum, and energy exchange to determination of the pressure, density and velocity fields. The implicit solution is accomplished iteratively without linearizing the equations, thus eliminating the need for numerous derivative terms. With the three-dimensional K-FIX code calculations in Cartesian and cylindrical geometries can be performed. Obstacles built from the computing cells can be specified within the computing volume. In cylindrical geometry, calculations can be performed in the full 360 degrees or any angular segment. To enhance computing efficiency a new cell indexing scheme has been introduced, and computing time is reduced further by deletion of

  3. Development of mass and energy release analysis methodology

    International Nuclear Information System (INIS)

    Kim, Cheol Woo; Song, Jeung Hyo; Park, Seok Jeong; Kim, Tech Mo; Han, Kee Soo; Choi, Han Rim

    2009-01-01

    Recently, new approaches to the accident analysis using the realistic evaluation have been attempted. These new approaches provide more margins to the plant safety, design, operation and maintenance. KREM (KEPRI Realistic Evaluation Methodology) for a large break loss-of-coolant accident (LOCA) is performed using RELAP5/MOD3 computer code including realistic evaluation models. KOPEC has developed KIMERA (KOPEC Improved Mass and Energy Release Analysis methodology) based on the realistic evaluation to improve the analysis method for the mass and energy (M/E) release and to obtain the adequate margin. KIMERA uses a simplified single code system unlike conventional M/E release analysis methodologies. This simple code system reduces the computing efforts especially for LOCA analysis. The computer code systems of this methodology are RELAP5K/CONTEMPT4 (or RELAP5-ME) like KREM methodology which couples RELAP5/MOD3.1/K and CONTEMPT4/MOD5. The new methodology, KIMERA based on the same engine as KREM, adopted conservative approaches for the M/E release such as break spillage model, multiplier on heat transfer coefficient (HTC), and long-term cooling model. KIMERA is developed based on a LOCA and applied to a main steam line break (MSLB) and approved by Korea Government. KIMERA has an ability of one-through calculation of the various transient stages of LOCAs in a single code system and calculate the M/E release analysis during the long term cooling period with the containment pressure and temperature (P/T) response. The containment P/T analysis results are compared with those of the Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) FSAR which is the OPR1000 (Optimized Power Reactor 1000) type nuclear power plant. The results of a large break LOCA and an MSLB are similar to those of FSAR for UCN 3 and 4. However, the containment pressure during the post-blowdown period of a large break LOCA has much lower second peak than the first peak. The resultant containment peak

  4. Analytical studies on optimization of containment design pressure

    International Nuclear Information System (INIS)

    Haware, S.K.; Ghosh, A.K.; Kushwaha, H.S.

    2005-01-01

    The containment of the proposed Advanced Heavy Water Reactor (AHWR) is divided into two main volumes viz. V1 and V2 interconnected by vent system via suppression pool. The arrangement is such that the volume V2 surrounds the volume V1 (see Fig.1). Blow Out Panels (BOPs), installed on volume V1 are designed to rupture at a differential pressure of 50 kPa. The containment was analysed using the in-house developed code CONTRAN, for three different scenario considered viz. (i) Loss of Coolant Accident (LOCA) involving double ended break in the downcomer pipe, (ii) LOCA involving double ended break in the reactor inlet header and (iii) Main Steam Line Break (MSLB) Accident. It was revealed that the accident involving the double-ended break of reactor inlet header results in the maximum value of the containment peak pressure. Results of the analyses indicated that the size of the BOP has bearing on the containment peak pressure. Therefore, five cases were analysed, varying the size of BOP from 0 to 10 m 2 , in order to quantify the influence of the size of BOP on the containment peak pressure. The blowdown mass and energy discharge data calculated using the code RELAP5/MOD3.2 was used in the analysis. It was observed that the vents are cleared in around 0.41 seconds into the accident. The containment peak pressures obtained in various cases are presented in Fig.2. The containment peak pressure varies with the size of BOP and passes through minima for a BOP size of around 5 m 2 . There are two flow processes, competing with each other viz. the steam-air mixture passage through the vent system via suppression pool and direct passage of steam air mixture through BOP bypassing the suppression pool. Though the energy suppression efficiency of the suppression pool decreases with increasing size of BOP, the pressure suppression efficiency was found to be maximum at around 5 m 2 size of BOP. The containment peak pressure passing through minima indicates that there is a scope for

  5. SULFUR POLYMER ENCAPSULATION.

    Energy Technology Data Exchange (ETDEWEB)

    KALB, P.

    2001-08-22

    Sulfur polymer cement (SPC) is a thermoplastic polymer consisting of 95 wt% elemental sulfur and 5 wt% organic modifiers to enhance long-term durability. SPC was originally developed by the U.S. Bureau of Mines as an alternative to hydraulic cement for construction applications. Previous attempts to use elemental sulfur as a construction material in the chemical industry failed due to premature degradation. These failures were caused by the internal stresses that result from changes in crystalline structure upon cooling of the material. By reacting elemental sulfur with organic polymers, the Bureau of Mines developed a product that successfully suppresses the solid phase transition and significantly improves the stability of the product. SPC, originally named modified sulfur cement, is produced from readily available, inexpensive waste sulfur derived from desulfurization of both flue gases and petroleum. The commercial production of SPC is licensed in the United States by Martin Resources (Odessa, Texas) and is marketed under the trade name Chement 2000. It is sold in granular form and is relatively inexpensive ({approx}$0.10 to 0.12/lb). Application of SPC for the treatment of radioactive, hazardous, and mixed wastes was initially developed and patented by Brookhaven National Laboratory (BNL) in the mid-1980s (Kalb and Colombo, 1985; Colombo et al., 1997). The process was subsequently investigated by the Commission of the European Communities (Van Dalen and Rijpkema, 1989), Idaho National Engineering Laboratory (Darnell, 1991), and Oak Ridge National Laboratory (Mattus and Mattus, 1994). SPC has been used primarily in microencapsulation applications but can also be used for macroencapsulation of waste. SPC microencapsulation has been demonstrated to be an effective treatment for a wide variety of wastes, including incinerator hearth and fly ash; aqueous concentrates such as sulfates, borates, and chlorides; blowdown solutions; soils; and sludges. It is not

  6. Holy sludge : Toronto's new bylaw and disposal strategy for biosolids impacts industry and sets a national precedent

    Energy Technology Data Exchange (ETDEWEB)

    Crittenden, G. [ed.

    2001-01-01

    Toronto, Ontario has implemented a new approach to the management of sewage sludge also known as biosolids. The decision was made to shut down its multi-hearth incinerator at the Ash bridges Bay Treatment Plant and increase the beneficial use, also called land application of biosolids, to 100 per cent in the near future. In addition, the disposal of dangerous chemicals, agricultural waste, and other wastes in municipal drains and sewers is being clamped down. It was determined that preventing pollutants from entering municipal wastewater would greatly increase public acceptance of the application of biosolids on agricultural land. All human activities will feel the impact, from organic waste in grocery stores to dental amalgams, from waste oil and solvents at auto repair shops to harsh chemical used in the metal plating industry. The new bylaw adopted by the City of Toronto prevents any individual from discharging or depositing into a storm sewer or watercourse (or a municipal or private sewer connecting with a storm sewer): hazardous waste chemicals, blowdown water, combustible liquids, floating debris, fuel, hauled sewage, hauled waste, hazardous industrial waste or chemicals, as well as an array of other substances. A plan must be submitted by any sector or industry discharging pollutants, and the plan must detail the processes generating the pollutants, as well as the measures required to eliminate the discharges over three and six years. An Industrial Waste Surcharge Agreement or a Sanitary Discharge Agreement might be obtained from the city covering the additional costs involved in treating the discharges. The only substances covered by those agreements are: biochemical oxygen demand, phenolics, total phosphorus and total suspended solids. Water originating from a source other than the city's water supply is covered under the sanitary agreement. Requirements for spills reporting, grease interceptors, motor oil interceptors, lubricating grease and

  7. Water Quality of NPP Secondary Side with Combined Water Chemistry of Ammonia and Ethanolamine

    International Nuclear Information System (INIS)

    Rhee, In-H.; Jung, Hyun-jun; Cho, Daechul; Park, Byunggi

    2012-09-01

    Ammonia (AM) and Ethanolamine (ETA), as pH control additive agents, were injected to the secondary side in a Korean NPP for the even pH in the entire secondary system including the wet region and the condensate. Ammonia and ETA are dominant in the vapor and liquid phases, respectively, since the former and latter are more and less volatile than water in the temperature range of 30 to 300 . pH of 9.5 to 9.7 was maintained in the water-steam cycle at the concentrations of ammonia with ∼1.0 ppm and ETA of ∼1.8 ppm. From the standpoint of corrosion, i.g, concentration of Fe, the water quality of secondary side was improved by the combined water treatment of ammonia and ETA, compared to all volatile treatment of ammonia. The electrical conductivity was increased from 6 to 10 μS/cm due to the presence of organic carboxylates produced by the decomposition of ETA. ETA was broken down by <5% in steam generator and converted into formate, acetate, and glycolate, among which acetate was largely formed. But inorganic ions such as Na + , Cl - , and SO 4 2- are not changed because their ingress was not made and the selectivity of resin over those ions was not fairly altered. The runtime of demineralizer in steam generator blowdown was shortened by a third for a mixture of ammonia and ETA. Most of Fe was originated from the shell side of heat exchangers including the condenser as a result of corrosion. Fe was only eliminated by ion exchange demineralizers, i.e., 46% at CPP and 3% at SG BD and 70% of Fe oxides were accumulated at the steam generator, on the basis of Fe concentration at the final feedwater. In conclusion, ETA is preferable to ammonia for the enhancement of pH in the liquid phase of water-steam mixture such as the shell side of heat exchanger and also the full-flow operation of CPP is more desirable than partial-flow operation for the improved removal of corrosion products, regardless of hydrogen- or amine-type operation. (authors)

  8. ChemANDTM - a system health monitor for plant chemistry

    International Nuclear Information System (INIS)

    Turner, C.W.; Mitchel, G.R.; Balakrishnan, P.V.; Tosello, G.

    1999-07-01

    Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that the plant's system health be continually monitored and managed. AECL has developed a System Health Monitor called ChemAND for CANDU plant chemistry. ChemAND, a Chemistry ANalysis and Diagnostic system, monitors key chemistry parameters in the heat transport system, moderator-cover gas, annulus gas, and the steam cycle during full-power operation and feeds these parameters to models that calculate the effect of current plant operating conditions on the present and future health of the system. Chemistry data from each of the systems are extracted on a regular basis from the plant's Historical Data Server and are sorted according to function, e.g., indicators for condenser in-leakage, air in-leakage, heavy water leakage into the annulus gas, fuel failure, etc. Each parameter is conveniently displayed and is trended along with its alarm limits. ChemAND currently has two analytical models developed for the balance-of-plant. CHEMSOLV calculates crevice chemistry conditions in the steam generator (SG) from either the SG blowdown chemistry conditions or from a simulated condenser leak. This information will be used by operations personnel to evaluate the potential for SG tube corrosion in the crevice region. CHEMSOLV also calculates chemistry conditions throughout the steam-cycle system, as determined by the transport of volatile species such as ammonia, hydrazine, morpholine, and oxygen. A second model, SLUDGE, calculates the deposit loading in the SG as a function of time, based on concentrations of corrosion product in the final feedwater and plant operating conditions. Operations personnel can use this information to predict where to inspect and when to clean. In a future development, SLUDGE will track deposit loading arising from start-up crud bursts and will be used in conjunction with the thermohydraulics code, THIRST, to predict

  9. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [KHNP-CENTERAL RESEARCH INSTITUTE, Daejeon (Korea, Republic of)

    2013-10-15

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic

  10. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    International Nuclear Information System (INIS)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

    2013-01-01

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic state in

  11. Recovery Act: Innovative CO2 Sequestration from Flue Gas Using Industrial Sources and Innovative Concept for Beneficial CO2 Use

    Energy Technology Data Exchange (ETDEWEB)

    Dando, Neal [Alcoa Inc., Pittsburgh, PA (United States); Gershenzon, Mike [Alcoa Inc., Pittsburgh, PA (United States); Ghosh, Rajat [Alcoa Inc., Pittsburgh, PA (United States)

    2012-07-31

    The overall goal of this DOE Phase 2 project was to further develop and conduct pilot-scale and field testing of a biomimetic in-duct scrubbing system for the capture of gaseous CO2 coupled with sequestration of captured carbon by carbonation of alkaline industrial wastes. The Phase 2 project, reported on here, combined efforts in enzyme development, scrubber optimization, and sequestrant evaluations to perform an economic feasibility study of technology deployment. The optimization of carbonic anhydrase (CA) enzyme reactivity and stability are critical steps in deployment of this technology. A variety of CA enzyme variants were evaluated for reactivity and stability in both bench scale and in laboratory pilot scale testing to determine current limits in enzyme performance. Optimization of scrubber design allowed for improved process economics while maintaining desired capture efficiencies. A range of configurations, materials, and operating conditions were examined at the Alcoa Technical Center on a pilot scale scrubber. This work indicated that a cross current flow utilizing a specialized gas-liquid contactor offered the lowest system operating energy. Various industrial waste materials were evaluated as sources of alkalinity for the scrubber feed solution and as sources of calcium for precipitation of carbonate. Solids were mixed with a simulated sodium bicarbonate scrubber blowdown to comparatively examine reactivity. Supernatant solutions and post-test solids were analyzed to quantify and model the sequestration reactions. The best performing solids were found to sequester between 2.3 and 2.9 moles of CO2 per kg of dry solid in 1-4 hours of reaction time. These best performing solids were cement kiln dust, circulating dry scrubber ash, and spray dryer absorber ash. A techno-economic analysis was performed to evaluate the commercial viability of the proposed carbon capture and sequestration process in full-scale at an aluminum smelter and

  12. POLYETHYLENE ENCAPSULATION

    International Nuclear Information System (INIS)

    Kalb, P.

    2001-01-01

    Polyethylene microencapsulation physically homogenizes and incorporates mixed waste particles within a molten polymer matrix, forming a solidified final waste form upon cooling. Each individual particle of waste is embedded within the polymer block and is surrounded by a durable, leach-resistant coating. The process has been successfully applied for the treatment of a broad range of mixed wastes, including evaporator concentrate salts, soil, sludges, incinerator ash, off-gas blowdown solutions, decontamination solutions, molten salt oxidation process residuals, ion exchange resins, granular activated carbon, shredded dry active waste, spill clean-up residuals, depleted uranium powders, and failed grout waste forms. For waste streams containing high concentrations of soluble toxic metal contaminants, additives can be used to further reduce leachability, thus improving waste loadings while meeting or exceeding regulatory disposal criteria. In this configuration, contaminants are both chemically stabilized and physically solidified, making the process a true stabilization/solidification (S/S) technology. Unlike conventional hydraulic cement grouts or thermosetting polymers, thermoplastic polymers such as polyethylene require no chemical. reaction for solidification. Thus, a stable, solid, final waste form product is assured on cooling. Variations in waste chemistry over time do not affect processing parameters and do not require reformulation of the recipe. Incorporation of waste particles within the polymer matrix serves as an aggregate and improves the mechanical strength and integrity of the waste form. The compressive strength of polyethylene microencapsulated waste forms varies based on the type and quantity of waste encapsulated, but is typically between 7 and 17.2 MPa (1000 and 2500 psi), well above the minimum strength of 0.4 MPa (160 psi) recommended by the U.S. Nuclear Regulatory Commission (NRC) for low-level radioactive waste forms in support of 10 CFR 61

  13. Application of Spatial Data Modeling and Geographical Information Systems (GIS) for Identification of Potential Siting Options for Various Electrical Generation Sources

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Gary T [ORNL; Belles, Randy [ORNL; Blevins, Brandon R [ORNL; Hadley, Stanton W [ORNL; Harrison, Thomas J [ORNL; Jochem, Warren C [ORNL; Neish, Bradley S [ORNL; Omitaomu, Olufemi A [ORNL; Rose, Amy N [ORNL

    2012-05-01

    contiguous United States. If a cell meets the requirements of each criterion, the cell is deemed a candidate area for siting a specific power generation form relative to a reference plant for that power type. Some SSEC parameters preclude siting a power plant because of an environmental, regulatory, or land-use constraint. Other SSEC assist in identifying less favorable areas, such as proximity to hazardous operations. All of the selected SSEC tend to recommend against sites. The focus of the ORNL electrical generation source siting study is on identifying candidate areas from which potential sites might be selected, stopping short of performing any detailed site evaluations or comparisons. This approach is designed to quickly screen for and characterize candidate areas. Critical assumptions supporting this work include the supply of cooling water to thermoelectric power generation; a methodology to provide an adequate siting footprint for typical power plant applications; a methodology to estimate thermoelectric plant capacity while accounting for available cooling water; and a methodology to account for future ({approx}2035) siting limitations as population increases and demands on freshwater sources change. OR-SAGE algorithms were built to account for these critical assumptions. Stream flow is the primary thermoelectric plant cooling source evaluated in this study. All cooling was assumed to be provided by a closed-cycle cooling (CCC) system requiring makeup water to account for evaporation and blowdown. Limited evaluations of shoreline cooling and the use of municipal processed water (gray) cooling were performed. Using a representative set of SSEC as input to the OR-SAGE tool and employing the accompanying critical assumptions, independent results for the various power generation sources studied were calculated.

  14. Performance of natural gas distribution networks during the Kocaeli earthquake - 17 august 1999; Comportement des reseaux de distributions de gaz naturel lors du tremblement de terre de Kocaeli 17 aout 1999

    Energy Technology Data Exchange (ETDEWEB)

    Zarea, M.; Adrien, M. [Gaz de France (GDF), 75 - Paris (France)

    2000-07-01

    The Kocaeli (Izmit) earthquake struck recently, on August 17, 1999, a well developed area of Turkey. This earthquake, of a magnitude 7.4 on the open Richter scale, severely damaged numerous buildings, industrial infrastructure, and made a lot of victims. In this context, most attention is given to issues like: seismology (why and how did it happen, what will happen next, etc.), seismic design and construction (why buildings collapsed and how to avoid this in the future). Some other subjects get less attention, because their direct influence in the overall damage is smaller. The behaviour of 'lifelines', designating all the networks which contribute to 'modern' lifestyle: water, energy, communications, etc., belong to this category. Nevertheless, the performance of lifelines during such strong earthquakes is also important, because they can contribute to minimise its impact. This impact has its usual two aspects: integrity and operability. For instance, the integrity requirement means that failures of the considered lifeline due to the earthquake should not directly affect property and life. The operability requirement means that a given subset of the lifeline remains operational, in order to fulfill vital tasks. We propose here a brief analysis of the performance of two relatively recently commissioned gas distribution systems: IZGAZ in Izmit, close the epicenter, and IDGAS in Istanbul. They have the advantage of representing a large sample of a recent implementation of the PE (polyethylene) technique, which has reached maturity. Both are cases of the Gaz de France 4 bar PE technology transferred to a Turkish operator, who completely managed the crisis. The first part describes the two networks, both their high medium pressure steel network, regulators, and the intermediate PE network, finishing with service lines and boxes. Then, the damage reported by the operational teams and their very important shut-down and blowdown actions are summarised

  15. UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

  16. Chemical treatment of deposits of junctions 'collector-tube' of horizontal steam generators

    International Nuclear Information System (INIS)

    Alkassem, S.N.

    2009-01-01

    ring gaps of underexpander can allow under specific conditions, created by lithium hydroxide, to provide a vapor lock and dry salt in the top of a crack. It results are of the density of corrosion current and the hydrogenation rate of coffer-dam perforated part of collector made from steel 08Ch18N10T. From these viewpoints the deposits are the important factor of prolongation the resource of coffer-dam. Complexing combination agents during their simultaneous presence in water, tends to form more than complex, mixed complexes with increased solubility for example, Ttrilon-B (EDTA) and its sodium salts. So if representing Complexing agents of iron represent in the form of [Fe-EDTA] 2 and [Fe-EDTA] - or hydroxocomplexing agents of iron [FeOH EDTA] - [FeOH EDTA] 2 [FeOH EDTA] 3 [Fe-EDTA] 4 a results of research EDTA solutions complexing agents indicate especially to a higher strength FE - EDTA and FEOH - EDTA (a great value of pK). The following conclusions are formulated in the paper: Optimization of water-chemical mode of the second contour remaining, the importance factor, for reducing the corrosion damage of steam generators and should be carried out according to the criteria of increasing the technical resource of metal as junctions, so as a whole of steam generator. Optimization processes of chemical washing should be come with reconstruction facilities of steam generator and its strapping (systems, blow-down, level measuring). Deactivation and chemical washing processes should be completed step of a passive protective film. (author)

  17. Computational studies of reacting flows with applications to zinc selenide nanoparticle synthesis and methane/hydrogen separation

    Science.gov (United States)

    Koutsona, Maria

    a predictive model describing pressure and concentration dynamics during Pressure Swing Adsorption (PSA) of binary (or pseudo-binary) gas mixtures. The separation of metane-hydrogen mixtures over 5A-zeolite was used as an example. The PSA cycle considered in this study includes the following 5 steps: (1) pressurization with product, (2) high-pressure adsorption, (3) cocurrent depressurization, (4) countercurrent blowdown and (5) countercurrent purge with product at low pressure. The PSA mathematical model describes the following processes gas flow in the bed (as axially dispersed plug flow) and the mass balance of the components of the mixture coupled to adsorption/desorption kinetics. The model results in a system of coupled partial differential equations in the axial bed dimension and time. The Galerkin Finite Element Method was used to discretize the equations in the axial direction of the bed. The resulting system of ordinary differential equations (ODE's) in time is solved by using an Euler full-implicit scheme. The model is being used by Chemical Design, Inc., for the initial design of PSA units.

  18. Dispersant application: (1) during steam generator wet layup for removal of existing deposits, and (2) during the long-path recirculation cleanup process of the condensate/feedwater system to reduce startup corrosion product transport to the steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Fruzzetti, K. [Electric Power Research Inst., Palo Alto, California (United States); Anderson, C.; Marks, C.; Kreider, M. [Dominion Engineering, Inc., Reston, Virginia (United States); Walton, B. [Exelon Corp., TMI Nuclear Generating Station, Middleton, Pennsylvania (United States); Reecher, W. [Exelon Corp., Byron Nuclear Generating Station, Byron, Illinois (United States); Morey, D. [Exelon Corp., Warrenville, Illinois (United States)

    2010-07-01

    During the last few decades, utilities have spent considerable resources minimizing corrosion product deposition within steam generators (SGs). In the past, two basic approaches have been used: Reducing the corrosion product ingress rate (e.g., by replacing secondary components containing corrosion-susceptible materials, implementing favorable chemistry changes, etc.); Removing corrosion products which have accumulated in the SGs through top-of-tubesheet (TTS) sludge lancing and other chemical and mechanical methods. Despite the success of these methods, there are limitations, including practical lower limits on the feedwater iron concentration and the high cost and effectiveness limits of cleaning techniques (particularly for crevices). A third approach is the online addition of a polymeric dispersant to promote suspension of corrosion iron, thereby reducing deposition onto SG surfaces and facilitating more efficient removal via blowdown. More than a decade of qualification work and two full-scale plant trials - at ANO-2 in 2000 and at McGuire Unit 2 from 2005 to 2007 - addressed initial technical concerns, paving the way for routine use in nuclear SGs. Online application of dispersant at the four Exelon plants with recirculating SGs is the focus of another paper at this conference. This paper is focused on the additional benefits that could be gained from similar dispersant applications during: normal SG wet layup to remove some of the existing deposit inventory; routine long-path recirculation cleanup of the PWR secondary side prior to startup. The addition of dispersant to the SGs during full wet layup periods could provide additional benefit by dispersing loose sludge powder that has accumulated, thereby facilitating its removal. Routine dispersant-assisted wet layup applications could be performed in conjunction with normal layup protocols without affecting the planned outage schedule, and could potentially reduce the frequency of more costly deposit

  19. Calculating corrections in F-theory from refined BPS invariants and backreacted geometries

    Energy Technology Data Exchange (ETDEWEB)

    Poretschkin, Maximilian

    2015-07-01

    This thesis presents various corrections to F-theory compactifications which rely on the computation of refined Bogomol'nyi-Prasad-Sommerfield (BPS) invariants and the analysis of backreacted geometries. Detailed information about rigid supersymmetric theories in five dimensions is contained in an index counting refined BPS invariants. These BPS states fall into representations of SU(2){sub L} x SU(2){sub R}, the little group in five dimensions, which has an induced action on the cohomology of the moduli space of stable pairs. In the first part of this thesis, we present the computation of refined BPS state multiplicities associated to M-theory compactifications on local Calabi-Yau manifolds whose base is given by a del Pezzo or half K3 surface. For geometries with a toric realization we use an algorithm which is based on the Weierstrass normal form of the mirror geometry. In addition we use the refined holomorphic anomaly equation and the gap condition at the conifold locus in the moduli space in order to perform the direct integration and to fix the holomorphic ambiguity. In a second approach, we use the refined Goettsche formula and the refined modular anomaly equation that govern the (refined) genus expansion of the free energy of the half K3 surface. By this procedure, we compute the refined BPS invariants of the half K3 from which the results of the remaining del Pezzo surfaces are obtained by flop transitions and blow-downs. These calculations also make use of the high symmetry of the del Pezzo surfaces whose homology lattice contains the root lattice of exceptional Lie algebras. In cases where both approaches are applicable, we successfully check the compatibility of these two methods. In the second part of this thesis, we apply the results obtained from the calculation of the refined invariants of the del Pezzo respectively the half K3 surfaces to count non-perturbative objects in F-theory. The first application is given by BPS states of the E

  20. Carbon Dust Filtration in Three Different Nuclear Process Environments: A comparison the challenges Carbon Dust Filtration Presents Under Different Process Conditions

    International Nuclear Information System (INIS)

    Chadwick, Chris

    2014-01-01

    Inits thirty five years of activity as an engineering company in nuclear filtration sector, the Porvair Filtration Group has experienced several demands to remove of Carbon/graphite dust from several nuclear gas streams. Of particular interest among those applications are, and those to be reported upon in this paper, are; • High temperature, high pressure, high DP resistant (high strength) filters operating in the CO2 environment of the UK fleet of AGR (Advanced Gas-Cooled Reactors) • Removing gross quantities of Carbon dust from the exhaust stream of a radioactive, nuclear organics decomposition, waste process • High pressure Helium filtration to remove Carbon dust for a gas flow associated with the Fuel Handling System in the High Temperature Reactor programme Each process is different from the other and presents its own unique problems. The paper will present to this conference the very different properties Carbon dust appears to exhibit in each of these very different applications, and to discuss the effects those significant differences had/have on Porvair’s responses to each application. An interesting comparison will be made of the substantial difference between the performance of the UK AGR filters and those used in the US for the removal of decomposed organics, and the significantly different properties the Carbon appears to exhibit in each unique set of conditions Two UK AGR stations which are described are taken out of service when their bypass blowdown filters reach an operating DP of about 700mB DP (starting at a clean DP of around 100mB) to enable their replacement. The used filter assemblies are lifted from their housings and placed in an active storage area. Analysis of the used filter assemblies has shown that, where they are observable, they appear to be pristine with no apparent surface discolouration. It is only when examined under magnification that it becomes obvious that the filter medium, under the outer layer of fibres, is coated in

  1. Water use in the development and operation of geothermal power plants.

    Energy Technology Data Exchange (ETDEWEB)

    Clark, C. E.; Harto, C. B.; Sullivan, J. L.; Wang, M. Q. (Energy Systems); ( EVS)

    2010-09-17

    , reservoir characteristics, and local climate have various effects on elements such as drilling rate, the number of production wells, and production flow rates. Over the life cycle of a geothermal power plant, from construction through 30 years of operation, plant operations is where the vast majority of water consumption occurs. Water consumption refers to the water that is withdrawn from a resource such as a river, lake, or non-geothermal aquifer that is not returned to that resource. For the EGS scenarios, plant operations consume between 0.29 and 0.72 gal/kWh. The binary plant experiences similar operational consumption, at 0.27 gal/kWh. Far less water, just 0.01 gal/kWh, is consumed during operations of the flash plant because geofluid is used for cooling and is not replaced. While the makeup water requirements are far less for a hydrothermal flash plant, the long-term sustainability of the reservoir is less certain due to estimated evaporative losses of 14.5-33% of produced geofluid at operating flash plants. For the hydrothermal flash scenario, the average loss of geofluid due to evaporation, drift, and blowdown is 2.7 gal/kWh. The construction stage requires considerably less water: 0.001 gal/kWh for both the binary and flash plant scenarios and 0.01 gal/kWh for the EGS scenarios. The additional water requirements for the EGS scenarios are caused by a combination of factors, including lower flow rates per well, which increases the total number of wells needed per plant, the assumed well depths, and the hydraulic stimulation required to engineer the reservoir. Water quality results are presented in Chapter 5. The chemical composition of geofluid has important implications for plant operations and the potential environmental impacts of geothermal energy production. An extensive dataset containing more than 53,000 geothermal geochemical data points was compiled and analyzed for general trends and statistics for typical geofluids. Geofluid composition was found to vary

  2. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    1 - Description of program or function: THALES, which stands for 'Thermal Hydraulic Analysis of Loss-of-coolant, Emergency core cooling and Severe core damage', is a computer code system for analyzing progression of core melt accident of light water reactors. The code was developed for Level 2 PSA (probabilistic safety assessment) and applicable to a wide range of postulated accident scenarios. Its outcomes are thermal hydraulic conditions in the reactor coolant system and the containment which are necessary for analyzing fission product release and transport behavior during the accident. The code system consists of following three member codes: (1) THALES-PM for accident progression in the primary and the secondary system of PWRs, (2) THALES-BM for accident progression in the reactor coolant system of BWRs, and (3) THALES-CV for accident progression in the containment of PWRs and BWRs. The THALES-PM and the THALES-BM codes carry out two categories of analysis. The first one is overall thermal-hydraulic analysis in the reactor coolant system. The reactor coolant system is divided into multi-volumes and each volume is further separated into a liquid region and a gas region by a movable mixture level. System pressure, mixture level in each volume, coolant temperature in each region, flow rate between volumes, etc. are calculated. The other one is core heatup and meltdown analysis. The reactor core is radially and axially divided into many nodes. Fuel and cladding temperature, cladding oxidation rate, hydrogen generation rate, core melt fraction, etc. are calculated. The THALES-CV code is for containment response analysis. It divides the containment into multiple compartments, each of which is further separated into a liquid region and a gas region by a movable mixture level. Containment pressure, mixture level in each compartment, coolant temperature in each region, flow rate between compartments, etc. are calculated. The code can treat coolant blowdown from the

  3. Institutional impediments to using alternative water sources in thermoelectric power plants.

    Energy Technology Data Exchange (ETDEWEB)

    Elcock, D. (Environmental Science Division)

    2011-08-03

    ), and with the local political organizations that can influence decisions regarding the use of the alternative source. Often a plan to use reclaimed water will work only if local politics and power plant goals converge. Even then, lengthy negotiations are often needed for the plans to come to fruition. (3) Regulatory requirements for planning and developing associated infrastructure such as pipelines, storage facilities, and back-up supplies that can require numerous approvals, permits, and public participation, all of which can create delays and increased costs. (4) Permitting requirements that may be difficult to meet, such as load-based discharge limits for wastewater or air emissions limitations for particulate matter (which will be in the mist of cooling towers that use reclaimed water high in dissolved solids). (5) Finding discharge options for cooling tower blowdown of reclaimed water that are acceptable to permitting authorities. Constituents in this wastewater can limit options for discharge. For example, discharge to rivers requires National Pollutant Discharge Elimination System (NPDES) permits whose limits may be difficult to meet, and underground injection can be limited because many potential injection sites have already been claimed for disposal of produced waters from oil and gas wells or waters associated with gas shale extraction. (6) Potential liabilities associated with using alternative sources. A power plant can be liable for damages associated with leaks from reclaimed water conveyance systems or storage areas, or with mine water that has been contaminated by unscrupulous drillers that is subsequently discharged by the power plant. (7) Community concerns that include, but are not limited to, increased saltwater drift on farmers fields; the possibility that the reclaimed water will contaminate local drinking water aquifers; determining the 'best' use of WWTP effluent; and potential health concerns associated with emissions from the

  4. An Innovative System for the Efficient and Effective Treatment of Non-Traditional Waters for Reuse in Thermoelectric Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    John Rodgers; James Castle

    2008-08-31

    This study assessed opportunities for improving water quality associated with coal-fired power generation including the use of non-traditional waters for cooling, innovative technology for recovering and reusing water within power plants, novel approaches for the removal of trace inorganic compounds from ash pond effluents, and novel approaches for removing biocides from cooling tower blowdown. This research evaluated specifically designed pilot-scale constructed wetland systems for treatment of targeted constituents in non-traditional waters for reuse in thermoelectric power generation and other purposes. The overall objective of this project was to decrease targeted constituents in non-traditional waters to achieve reuse criteria or discharge limitations established by the National Pollutant Discharge Elimination System (NPDES) and Clean Water Act (CWA). The six original project objectives were completed, and results are presented in this final technical report. These objectives included identification of targeted constituents for treatment in four non-traditional water sources, determination of reuse or discharge criteria for treatment, design of constructed wetland treatment systems for these non-traditional waters, and measurement of treatment of targeted constituents in non-traditional waters, as well as determination of the suitability of the treated non-traditional waters for reuse or discharge to receiving aquatic systems. The four non-traditional waters used to accomplish these objectives were ash basin water, cooling water, flue gas desulfurization (FGD) water, and produced water. The contaminants of concern identified in ash basin waters were arsenic, chromium, copper, mercury, selenium, and zinc. Contaminants of concern in cooling waters included free oxidants (chlorine, bromine, and peroxides), copper, lead, zinc, pH, and total dissolved solids. FGD waters contained contaminants of concern including arsenic, boron, chlorides, selenium, mercury

  5. Control of supersonic axisymmetric base flows using passive splitter plates and pulsed plasma actuators

    Science.gov (United States)

    Reedy, Todd Mitchell

    An experimental investigation evaluating the effects of flow control on the near-wake downstream of a blunt-based axisymmetric body in supersonic flow has been conducted. To better understand and control the physical phenomena that govern these massively separated high-speed flows, this research examined both passive and active flow-control methodologies designed to alter the stability characteristics and structure of the near-wake. The passive control investigation consisted of inserting splitter plates into the recirculation region. The active control technique utilized energy deposition from multiple electric-arc plasma discharges placed around the base. The flow-control authority of both methodologies was evaluated with experimental diagnostics including particle image velocimetry, schlieren photography, surface flow visualization, pressure-sensitive paint, and discrete surface pressure measurements. Using a blowdown-type wind tunnel reconstructed specifically for these studies, baseline axisymmetric experiments without control were conducted for a nominal approach Mach number of 2.5. In addition to traditional base pressure measurements, mean velocity and turbulence quantities were acquired using two-component, planar particle image velocimetry. As a result, substantial insight was gained regarding the time-averaged and instantaneous near-wake flow fields. This dataset will supplement the previous benchmark point-wise laser Doppler velocimetry data of Herrin and Dutton (1994) for comparison with new computational predictive techniques. Next, experiments were conducted to study the effects of passive triangular splitter plates placed in the recirculation region behind a blunt-based axisymmetric body. By dividing the near-wake into 1/2, 1/3, and 1/4 cylindrical regions, the time-averaged base pressure distribution, time-series pressure fluctuations, and presumably the stability characteristics were altered. While the spatial base pressure distribution was

  6. Development Of Nutrient And Water Recycling Capabilities In Algae Biofuels Production Systems. Final Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lundquist, Tryg [California Polytechnic State Univ. (CalPoly), San Luis Obispo, CA (United States). Civil and Environmental Engineering Dept.; Spierling, Ruth [California Polytechnic State Univ. (CalPoly), San Luis Obispo, CA (United States); Poole, Kyle [California Polytechnic State Univ. (CalPoly), San Luis Obispo, CA (United States); Blackwell, Shelley [California Polytechnic State Univ. (CalPoly), San Luis Obispo, CA (United States); Crowe, Braden [California Polytechnic State Univ. (CalPoly), San Luis Obispo, CA (United States); Hutton, Matt [California Polytechnic State Univ. (CalPoly), San Luis Obispo, CA (United States); Lehr, Corinne [California Polytechnic State Univ. (CalPoly), San Luis Obispo, CA (United States). Dept. of Chemistry and Biochemistry

    2018-01-25

    conventional heated mixed lab digester yielded 0.22 LCH4/g VS with 0.25 g VS/L-d and 30oC. The highest yield (0.30 LCH4/g VS) was achieved by the unmixed lab digesters operated at a constant 20oC. All digesters were operated with a 40-d hydraulic residence time. 6. In general, 50-75% of initial particulate N and P could be solubilized during anaerobic digestion and available for subsequent rounds of algae cultivation. 7. Bench-scale experiments showed the recovery from hydrothermal liquefaction (HTL) wastewater of carbon via anaerobic digestion and of nutrients to grow algae. To satisfy the nitrogen demand of algae cultivation, HTL wastewater would be diluted 400-fold, which was found to eliminate inhibition of algae growth by HTL wastewater. 8. Anaerobic digestion methane yield was lower for algal biomass containing coagulants such as would be used to aid harvesting or dewatering. Depending on doses, starch-based coagulant decreased yield by 10-14% and aluminum chlorohydrate decreased it by 14-26%. The lowest yield was 0.28 L CH4/g volatile solids introduced to the digesters. 9. Algae harvested from raceways operated on recycled water had methane yields 13% higher than algae from raceways operated on both recycled water and nutrients provided by algae digestate. The slightly lower yield was expected due to the presence of previously digested biomass from the digestate fertilizer. 10. Defined media was replenished with nutrients and recycled repeatedly in sequential batch growth of Chlorella sorokiniana (DOE 1412). This laboratory study tested for inhibition and accumulation of inhibiting compounds (allelopathic or auto-inhibitory substances), information that would help estimate the blowdown ratio needed for an integrated system. In laboratory experiments in which water was recycled a total of five times, each successive round of reuse resulted in an average 4±3% reduction in log-phase specific growth rates. However, linear-phase growth

  7. A Synergistic Combination of Advanced Separation and Chemical Scale Inhibitor Technologies for Efficient Use of Imparied Water As Cooling Water in Coal-based Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jasbir Gill

    2010-08-30

    commercial product commonly used for silica/silicate control. Additional pilot cooling tower testing confirmed the bench study. We also developed a molecule to inhibit calcium carbonate precipitation and calcium sulfate precipitation at high supersaturations. During Phase 3, a long-term test of the EDI system and scale inhibitors was done at Nalco's cooling tower water testing facility, producing 850 gallons of high purity water (90+% salt removal) at a rate of 220 L/day. The EDI system's performance was stable when the salt concentration in the concentrate compartment (i.e. the EDI waste stream) was controlled and a CIP was done after every 48 hours of operation time. A combination of EDI and scale inhibitors completely eliminated blowdown discharge from the Pilot cooling Tower. The only water-consumption came from evaporation, CIP and EDI concentrate. Silica Inhibitor was evaluated in the field at a western coal fired power plant.

  8. THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2001-01-01

    . - The most important feature of THYDE-W is the conservation of mass, momentum and energy. 2 - Method of solution: In THYDE-P, a PWR plant is regarded as a net- work of various coolant components which may be classified into nodes and junctions. The one-dimensional mass, momentum and energy equations are suitably integrated in each node and junction. In integrating the resulting equations with respect to time, a non- linear implicit method is used on the basis of the Newton method. The Jacobian matrix of the basic equations can be reduced to a simple form by the network theory, which is one of the characteristics of THYDE-P. To solve the basic equations by the non-linear implicit method, various smoothing functions with respect to time are introduced for mode changes such as phase change and flow re- versal. New models for a steam generator and a pressurizer are implemented. A THYDE-P calculation is started by a steady-state adjustment, where the basic equations are exactly solved without time derivatives. THYDE-P is able to calculate through both blowdown and re- fill-reflood phases without any change of models and physical conditions of the coolant. A model which takes non-equilibrium effects into account is newly implemented. 3 - Restrictions on the complexity of the problem: It is required that the network has at least one mixing junction except for the core heatup calculation mode and that a normal node without heat source (or sink) must be placed at both the top and bottom ends of the core. After so reticulating the plant, we have a number of nodes and junctions separately, strictly in numeric order in accordance with the following rules: (a) Normal nodes (except linkage nodes) should be numbered in numerical order chain-wise from one mixing junction to another according to the direction of the steady state chain flow. (b) Then linkage nodes should be numbered in numeric order chain- wise from the corresponding mixing junction. (c) Special nodes should be numbered

  9. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    ranging from 4 to 21 bar with three different static inventories of non-condensable gas. Condensation and heat transfer rates were evaluated employing several methods, notably from measured temperature gradients in the HTP as well as measured condensate formation rates. A detailed mass and energy accounting was used to assess the various measurement methods and to support simplifying assumptions required for the analysis. Condensation heat fluxes and heat transfer coefficients are calculated and presented as a function of pressure to satisfy the objectives of this investigation. The major conclusions for those tests are summarized below: (1) In the steam blow-down tests, the initial condensation heat transfer process involves the heating-up of the containment heat transfer plate. An inverse heat conduction model was developed to capture the rapid transient transfer characteristics, and the analysis method is applicable to SMR safety analysis. (2) The average condensation heat transfer coefficients for different pressure conditions and non-condensable gas mass fractions were obtained from the integral test facility, through the measurements of the heat conduction rate across the containment heat transfer plate, and from the water condensation rates measurement based on the total energy balance equation. 15 (3) The test results using the measured HTP wall temperatures are considerably lower than popular condensation models would predict mainly due to the side wall conduction effects in the existing MASLWR integral test facility. The data revealed the detailed heat transfer characteristics of the model containment, important to the SMR safety analysis and the validation of associated evaluation model. However this approach, unlike separate effect tests, cannot isolate the condensation heat transfer coefficient over the containment wall, and therefore is not suitable for the assessment of the condensation heat transfer coefficient against system pressure and noncondensable