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Sample records for blowdown

  1. Blowdown heat transfer experiment, (1)

    International Nuclear Information System (INIS)

    Blowdown heat transfer experiment has been carried out with a transparent test section to observe phenomena in coolant behavior during blowdown process. Experimental parameters are discharge position, initial system pressure, initial coolant temperature, power supply to heater rods and number of heater rods. At initial pressure 7-12 ata and initial power 6-50 kw per one heater rod, the flow condition in the test section is a major factor in determining time of DNB occurrence and physical process to DNB during blowdown. (auth.)

  2. Monodromy Substitutions and Rational Blowdowns

    CERN Document Server

    Endo, Hisaaki; van Horn-Morris, Jeremy

    2010-01-01

    We introduce several new families of relations in the mapping class groups of planar surfaces, each equating two products of right-handed Dehn twists. The interest of these relations lies in their geometric interpretation in terms of rational blowdowns of 4-manifolds, specifically via monodromy substitution in Lefschetz fibrations. The simplest example is the lantern relation, already shown by the first author and Gurtas to correspond to rational blowdown along a -4 sphere; here we give relations that extend that result to realize the "generalized" rational blowdowns of Fintushel-Stern and Park by monodromy subsitution, as well as several of the families of rational blowdowns discovered by Stipsicz-Szab\\'o-Wahl.

  3. BWR drywell behavior under steam blowdown

    International Nuclear Information System (INIS)

    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. Blowdown Simulation of CO2 Pipelines

    OpenAIRE

    Collard, A

    2015-01-01

    Pipelines are the most practical option for transporting large volumes of captured CO2 to appropriate storage sites as part of the Carbon Capture and Storage (CCS) process. Proper maintenance, including periodic blowdown of pipelines or pipeline sections, is necessary for their safe operation, a pre-requisite for the public acceptance of CCS. Given the relatively high Joule-Thomson coefficient of CO2, blowdown can present significant risks to pipeline infrastructure. Depressurisation will res...

  5. 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

  6. PPOOLEX experiments with two parallel blowdown pipes

    International Nuclear Information System (INIS)

    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 factors

  7. LOFT blowdown experiment safety analysis methodology

    International Nuclear Information System (INIS)

    An unprecedented blowdown experiment safety analysis (ESA) has been performed for the first two scheduled nuclear experiments in the Loss-of-Fluid Test (LOFT) facility. The ESA methodology is a unique approach needed to estimate conservatively the maximum consequences that will occur during an experiment. Through use of this information an acceptable risk in terms of adequate protection of the facility, personnel, and general public can be balanced with the requirements of the experiment program objectives. As an example, one of the LOFT program objectives is to evaluate the performance and effectiveness of emergency core cooling systems (ECCS) while relying on the same ECCSs (and backup ECCSs) to effectively perform as plant protection systems (PPS). The purpose of this paper is to present the LOFT blowdown ESA methodology

  8. Optimization of Boiler Blowdown and Blowdown Heat Recovery in Textile Sector

    Directory of Open Access Journals (Sweden)

    Sunudas T

    2013-09-01

    Full Text Available Boilers are widely used in most of the processing industries like textile, for the heating applications. Surat is the one of the largest textile processing area in India. In textile industries coal is mainly used for the steam generation. In a textile industry normally a 4% of heat energy is wasted through blowdown. In the study conducted in steam boilers in textile industries in surat location, 1.5% of coal of total coal consumption is wasted in an industry by improper blowdwon. This thesis work aims to prevent the wastage in the coal use by optimizing the blowdown in the boiler and maximizing the recovery of heat wasting through blowdown.

  9. 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.

  10. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  12. PPOOLEX experiments with a modified blowdown pipe outlet

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

    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

  13. PPOOLEX experiments with a modified blowdown pipe outlet

    International Nuclear Information System (INIS)

    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

  14. 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...

  15. Application of RELAP5 to a pipe blowdown experiment

    International Nuclear Information System (INIS)

    The application of the RELAP5 computer program to a pipe blowdown experiment is described in this paper. The basic hydrodynamic model, constitutive relations, and special process models included in RELAP5 are also briefly discussed. The results of this application confirm the effectiveness of using a choked flow model

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

    International Nuclear Information System (INIS)

    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

  17. Containment steam blowdown analysis: experimental and numerical comparisons

    International Nuclear Information System (INIS)

    This paper compares the numerical simulation with the experimental data of a steam blowdown event in a light water reactor containment building. A three step approach was used to analyze the steam jet behavior. First, the temperature and pressure data of a stem 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. Second, 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. Finally, 2-Dimensional and 3-Dimensional CFD models of the discharged steam jets were computed using PHOENICS, a commercially available CFD package. It was found that RELAP5 is reasonably capable in predicting the general temperature and pressure trends in the RPV. However, due to modeling compromises and the code's built-in capabilities, RELAP5 1-Dimensional predictions of containment temperature and pressure did not compare well with measured data. On the other hand, with minor modifications to the k-ε turbulence model, the 2-Dimensional and 3-Dimensional PHOENICS CFD solutions compared extremely well with the measured data. (author)

  18. Multiple blowdown pipe experiments with the PPOOLEX facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-15

    This report summarizes the results of the experiments with two steel 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 the blowdown pipes to the condensation pool. The main purpose of the experiment series was to study chugging phenomena (rapid condensation) while steam is discharged through two parallel blowdown pipes into the condensation pool filled with sub-cooled water. Particularly, the aim was to study if the pipe material (polycarbonate) used in the earlier experiment series with two blowdown pipes has had an effect on the general chugging behaviour and measured loads. In the experiments the initial temperature of the pool water was 20 deg. C. The steam flow rate ranged from 220 g/s to 2 350 g/s and the temperature of incoming steam from 148 deg. C to 207 deg. C. The formation and collapse of steam bubbles and the movement of the steam/water interface inside the pipes was non-synchronous. There could be even a 70 ms time difference between the occurrences of steam bubble collapses at the outlets of the two pipes. There was no clear pattern in which pipe the steam bubble first starts to collapse. Several successive bubbles could collapse first in either pipe but then the order changed for a single or several cycles. High pressure loads were measured inside the blowdown pipes due to rapid condensation of the steam volumes in the pipes and resulting water hammer effects. The loads seemed to be higher in pipe 1 than in pipe 2. An explanation for this could be a possible unequal distribution of steam flow between the two pipes. The pipe material has an effect on the condensation phenomena inside the blowdown pipes. A huge difference in the measured pressure curves inside the pipes could be observed compared to the experiments with the polycarbonate pipes. With the same test conditions the amplitude of the

  19. Multiple blowdown pipe experiments with the PPOOLEX facility

    International Nuclear Information System (INIS)

    This report summarizes the results of the experiments with two steel 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 the blowdown pipes to the condensation pool. The main purpose of the experiment series was to study chugging phenomena (rapid condensation) while steam is discharged through two parallel blowdown pipes into the condensation pool filled with sub-cooled water. Particularly, the aim was to study if the pipe material (polycarbonate) used in the earlier experiment series with two blowdown pipes has had an effect on the general chugging behaviour and measured loads. In the experiments the initial temperature of the pool water was 20 deg. C. The steam flow rate ranged from 220 g/s to 2 350 g/s and the temperature of incoming steam from 148 deg. C to 207 deg. C. The formation and collapse of steam bubbles and the movement of the steam/water interface inside the pipes was non-synchronous. There could be even a 70 ms time difference between the occurrences of steam bubble collapses at the outlets of the two pipes. There was no clear pattern in which pipe the steam bubble first starts to collapse. Several successive bubbles could collapse first in either pipe but then the order changed for a single or several cycles. High pressure loads were measured inside the blowdown pipes due to rapid condensation of the steam volumes in the pipes and resulting water hammer effects. The loads seemed to be higher in pipe 1 than in pipe 2. An explanation for this could be a possible unequal distribution of steam flow between the two pipes. The pipe material has an effect on the condensation phenomena inside the blowdown pipes. A huge difference in the measured pressure curves inside the pipes could be observed compared to the experiments with the polycarbonate pipes. With the same test conditions the amplitude of the

  20. PWR-blowdown heat transfer separate effects program

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Condensation pool experiments with steam using DN200 blowdown pipe

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-08-01

    This report summarizes the results of the condensation pool experiments with steam using a DN200 blowdown pipe. Altogether five experiment series, each consisting of several steam blows, were carried out in December 2004 with a scaled-down test facility designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to increase the understanding of different phenomena in the condensation pool during steam discharge. (au)

  2. Blowdown Wind Tunnels: Latest Citations from the Aerospace Database

    Science.gov (United States)

    1996-01-01

    The bibliography contains citations concerning the design, construction, operation, and performance of blowdown wind tunnels. The use of compressed gas, mechanical piston, or combustion exhaust to provide continuous or short-duration operation from transonic to hypersonic approach velocities is discussed. Also covered are invasive and non-invasive aerothermodynamic instrumentation, data acquisition and reduction techniques, and test reports on aerospace components. Comprehensive coverage of wind tunnel force balancing systems and supersonic wind tunnels are covered in separate bibliographies.

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

    International Nuclear Information System (INIS)

    A small-scale experiment using Freon-11 at 1300F (54.40C) and 65 psia (0.45 MPa) 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 61-cm 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 1300F (54.40C) and 60 psig (0.517 MPa) 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 conditions of thermodynamic nonequilibrium 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

  4. Increased Frequency of Large Blowdown Formation in Years With Hotter Dry Seasons in the Northwestern Amazon

    Science.gov (United States)

    Rifai, S. W.; Anderson, L. O.; Bohlman, S.

    2015-12-01

    Blowdowns, which are large tree mortality events caused by downbursts, create large pulses of carbon emissions in the short term and alter successional dynamics and species composition of forests, thus affecting long term biogeochemical cycling of tropical forests. Changing climate, especially increasing temperatures and frequency of extreme climate events, may cause changes in the frequency of blowdowns, but there has been little spatiotemporal analysis to associate the interannual variation in the frequency of blowdowns with annual climate parameters. We mapped blowdowns greater than 25 ha using a time series of Landsat images from 1984-2012 in the northwestern Amazon to estimate the annual size distribution of these blowdowns. The difference in forest area affected by blowdowns between the years with the highest and lowest blowdown activity were on the order of 10 - 30 times greater depending on location. Spatially, we found the probability of large blowdowns to be higher in regions with higher annual rainfall. Temporally, we found a positive correlation between the probability of large blowdown events and maximum dry season air temperature (R2 = 0.1-0.46). Mean and maximum blowdown size also increased with maximum dry season air temperature. The strength of these relationships varied between scene locations which may be related to cloud cover obscuring the land surface in the satellite images, or biophysical characteristics of the sites. Potentially, elevated dry season temperatures during the transition from the dry season to the wet season (October - December) may exacerbate atmospheric instabilities, which promote downburst occurrences. Most global circulation models predict dry season air temperatures to increase 2-5 ℃ in the northwestern Amazon by 2050. Should the blowdown disturbance regime continue increasing with elevated dry season temperatures, the northwestern Amazon is likely to experience more catastrophic tree mortality events which has direct

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

    International Nuclear Information System (INIS)

    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

  6. Reactivity transients during a blowdown in a MSIV closure ATWS

    International Nuclear Information System (INIS)

    Anticipated transients without scram (ATWS) events have received considerable attention in the past and are still a subject of great interest in severe-accident analysis. Of special interest is the effect of the low-pressure emergency core cooling system (ECCS) on the plant response following a blowdown by the automatic depressurization system (ADS). There is a potential for positive reactivity insertion due to the cold water injection of the low-pressure coolant injection (LPCI) system and the low-pressure core spray system in a boiling water reactor (BWR)/4. The main concern is whether a power excursion and pressure oscillation can occur in such an event. Furthermore, since thermal-hydraulic feedback plays an important role in these accidents, the uncertainty of the reactivity feedback coefficients used can impact the outcome of the analysis for such a power excursion. The objectives of the work reported in this paper are to study the consequences of the reactivity transients during a blowdown in an ATWS event with closure of the main steam isolation valves (MSIVs) and to evaluate the effect of the LPCI system and the sensitivity of plant response to the feedback coefficients. This work was performed with the Brookhaven National Laboratory plant analyzer

  7. 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

  8. A Hydraulic Blowdown Servo System For Launch Vehicle

    Science.gov (United States)

    Chen, Anping; Deng, Tao

    2016-07-01

    This paper introduced a hydraulic blowdown servo system developed for a solid launch vehicle of the family of Chinese Long March Vehicles. It's the thrust vector control (TVC) system for the first stage. This system is a cold gas blowdown hydraulic servo system and consist of gas vessel, hydraulic reservoir, servo actuator, digital control unit (DCU), electric explosion valve, and pressure regulator etc. A brief description of the main assemblies and characteristics follows. a) Gas vessel is a resin/carbon fiber composite over wrapped pressure vessel with a titanium liner, The volume of the vessel is about 30 liters. b) Hydraulic reservoir is a titanium alloy piston type reservoir with a magnetostrictive sensor as the fluid level indicator. The volume of the reservoir is about 30 liters. c) Servo actuator is a equal area linear piston actuator with a 2-stage low null leakage servo valve and a linear variable differential transducer (LVDT) feedback the piston position, Its stall force is about 120kN. d) Digital control unit (DCU) is a compact digital controller based on digital signal processor (DSP), and deployed dual redundant 1553B digital busses to communicate with the on board computer. e) Electric explosion valve is a normally closed valve to confine the high pressure helium gas. f) Pressure regulator is a spring-loaded poppet pressure valve, and regulates the gas pressure from about 60MPa to about 24MPa. g) The whole system is mounted in the aft skirt of the vehicle. h) This system delivers approximately 40kW hydraulic power, by contrast, the total mass is less than 190kg. the power mass ratio is about 0.21. Have finished the development and the system test. Bench and motor static firing tests verified that all of the performances have met the design requirements. This servo system is complaint to use of the solid launch vehicle.

  9. 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.

  10. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  11. 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)

  12. Interactive model of the steam generator blowdown system of the Kozloduy NPP units 5 and 6

    International Nuclear Information System (INIS)

    A program complex for full scale modelling of the WWER reactor dynamics 'RADUGA-EU' has been developed. The complex includes: program complex 'RADUGA - 7.3' - for non-stationary neutron-hydraulic calculations of reactors with 3-dimensional reactor core; package TPP (Thermal Power Plant) - for modeling of non-stationary and stationary processes in complicated thermal-hydraulic networks (including primary and secondary circuit of NPPs and TPPs); program 'GENERATOR' - for modeling of the electric generator; program 'MVTU (Modeling in technical devices)' - for modeling, analysis and optimisation of dynamical processes. The 'RADUGA-EhU' is used for calculations of thermal-hydraulic processes in the steam generator blowdown system of the Kozloduy NPP units 5 and 6. An interactive model of the blowdown system has been created. It is used not only for testing of technical solutions, but also as interactive environment for the development of the equipment operational regulations

  13. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  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. Investigation of Minimum Film boiling Phenomena on Fuel Rods Under Blowdown Cooling Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bajorek; Michael Gawron; Timothy Etzel; Lucas Peterson

    2003-06-30

    Blowdon cooling heat transfer is an important process that occurs early in a hypothetical large break loss-of-coolant accident (LOCA) in a pressurized water reactor. During blowdown, the flow through the hot assembly is a post-critical heat flux dispersed droplet flow. The heat transfer mechanisms that occur in blowdown cooling are complex and depend on droplet and heated surface interaction. In a safety analysis, it is of considerable importance to determine the thermal-hydraulic conditions leading to the minimum film boiling temperature, Tmin. A flow boiling rig for measurement of blowdown cooling heat transfer and quench phenomena on a nuclear fuel rod simulator was designed and constructed for operation at up to 12.4 MPa. The test section consisted of a concentric annulus, with a 9.5 mm OD nuclear fuel rod simulator at the center. The rod was contained within a 0.85 mm thick, 19 mm OD 316 stainless steel tube, forming the flow channel. Two types of rods were tested; one type was sheathed with Inconel 600 while the other was clad with Zircaloy-2. Water was injected into the test section at the top of the heated length through an injection header. This header was an annular sign that fit around the fuel rod simulator and within the stainless steel tube. Small spacers aligned the injection header and prevented contract with either the heater rod or the tube. A series of small diameter holes at the bottom of the header caused the formation of droplets that became entrained with the steam flow. The test section design was such that quench would take place on the rod, and not along the channel outer annulus.

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

    International Nuclear Information System (INIS)

    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

  17. A theoretical and numerical investigation of turbulent steam jets in BWR steam blowdown

    International Nuclear Information System (INIS)

    The preliminary results of PHOENICS and RELAP5 show that the current numerical models are adequate in predicting steam flow and stratification patterns in the upper Drywell of a BWR containment subsequent to a blow-down event. However, additional modeling is required in order to study detailed local phenomena such as condensation with non-condensables, natural convection, and stratification effects. Analytically, the intermittence modified similarity solutions show great promise. Once γ is accounted for, the jet's turbulent shear stress can be determined with excellent accuracy

  18. 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

  19. Assessment of SWBR safety-relief valve discharge line dynamic loads due to steam blowdown

    Energy Technology Data Exchange (ETDEWEB)

    Spoelstra, S.; Stoop, P.M. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Dijk, A.B. van [Stork Nucon BV (Netherlands)

    1997-06-01

    The Safety/Relief Valve Discharge Lines of the SBWR nuclear power plant are subject to dynamic loads due to steam blowdown after rapid opening of the Safety/Relief Valves. This paper describes the calculation of the thermal-hydraulic loads exerted on the piping system and the calculation of the resulting pipe stresses. These calculations have been performed using the CHARME and PS+CAEPIPE computer programs respectively. The calculated pipe stresses have been combined with the ones resulting from dead weight and thermal expansion and compared with ASME III criteria. (orig.).

  20. 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.

  1. Debris transport evaluation during LOCA blow-down using CFD methodology for OPR-1000 plant

    International Nuclear Information System (INIS)

    In response to GSI-191, 'Potential of PWR Sump Blockage Post-LOCA', NEI and the industry formed the PWR Sump Performance Task Force. The primary purpose of Task Force was to creation of a methodology document that could be used as guideline for PWR operators to address the issue. The NEI methodology document provides basic guidance on approach and various methods available. But some additional information be required in order to apply to specific plants, such as OPR-1000, and APR-1400 plant. According to the baseline evaluation of NEI 04-07, debris transport logic chart was composed of 4 transport phases. The present work aim to evaluate debris transport during LOCA blow-down, the first transport phase, based on CFD analysis. The target plant is Ulchin 3 and 4 which is OPR-1000 plant. Flow pattern strongly affects shape of containment, and disposition of components, such as steam generators, RCPs, and pipes, etc. The present work takes advantage of 3D CAD model so that real geometry of OPR-1000 plant is used. The analysis results give a clear figure about flow pattern in containment during LOCA blow-down, and fraction of debris transport to upper containment, which is one of major safety issues. (author)

  2. 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.

  3. PKL-tests, test series IIB (end of blowdown). Vol. 2

    International Nuclear Information System (INIS)

    As part of the federally subsidized research project 1500 287/A0, the system behavior of a 1300 MWe pressurized water reactor (PWR) was investigated during the depressurization phase (end-of-blowdown, EOB), as well as during the refill and reflood phases of a loss of coolant accident involving a large break in the reactor coolant loop. Appropriate modifications to the system and supplementary instrumentation have made it possible to simulate the EOB (as of 26 bar), the refill phase and reflood phase in sequence. This report includes a detailed description of the instrumentation and the data acquisition system used in Test Series PKL IIB. (orig.) With 6 refs., 2 tabs., 60 figs

  4. A Blowdown Cryogenic Cavitation Tunnel and CFD Treatment for Flow Visualization around a Foil

    Institute of Scientific and Technical Information of China (English)

    Yutaka ITO; Kazuya SAWASAKI; Naoki TANI; Takao NAGASAKI; Toshio NAGASHIMA

    2005-01-01

    Cavitation is one of the major problems in the development of rocket engines. There have been few experimental studies to visualize cryogenic foil cavitation. Therefore a new cryogenic cavitation tunnel of blowdown type was built. The foil shape is "plano-convex". This profile was chosen because of simplicity, but also of being similar to the one for a rocket inducer impeller. Working fluids were water at room temperature,hot water and liquid nitrogen. In case of Angle of Attack (AOA)=8°, periodical cavity departure was observed in the experiments of both water at 90℃ and nitrogen at -190℃ under the same velocity 10 m/sec and the same cavitation number 0.7. The frequencies were observed to be 110 and 90 Hz, respectively, and almost coincided with those of vortex shedding from the foil. Temperature depression due to the thermodynamic effect was confirmed in both experiment and simulation especially in the cryogenic cavitation.

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

    International Nuclear Information System (INIS)

    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)

  6. Investigations of the fluctuating pressure field in the suppression pool of the Marviken containment during blowdown

    International Nuclear Information System (INIS)

    From August 1972 until May 1973 blowdown tests were performed at the Marviken reactor plant. The tests were intended to provide information about the behaviour of a reactor safety containment with pressure suppression system in case of a loss-of-coolant accident resulting from a rupture in the primary circuit. Besides of experiments on the behaviour of the containment parallel experiments were conducted relative to the transport of iodine, the behaviour of components, and the tightness of the containment. Within this test program the Gesellschaft fuer Kernforschung measured the local pressure pulsation field in the water pool as well as the mass flows entering the pressure suppression system. The measurements were performed to provide first a general view of the vibration phenomena in the water pool to allow subsequent interpretation by means of physical models and processing by computation. (Auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Steam generator blowdown demineralisation plant (DARA). Influences on the separation of silica and corrosion products in Beznau nuclear power plant

    International Nuclear Information System (INIS)

    A constant reduction of the silica retention efficiency of the ion-exchanger has been observed in the steam-generator blow-down demineralisation plant (DARA), following resin-renewal. The reason for this phenomenon, as well as for the increased silica content of the processed water cannot be attributed to the ingress of foreign silica. Instead, the cause has been determined to be a blockage of the ion-exchanger by organic substances. The source of these substances has been pinpointed to activated hydrazine a component of which, hydrochinone, plays a major role. Tests have shown that the role of the blowdown as a means of reducing the deposition of corrosions products transported into the steam generator (SG) has hitherto been underestimated. (orig.)

  9. ALARM-P1: a computer program for pressurized water reactor blowdown analysis

    International Nuclear Information System (INIS)

    The computer program ALARM-P1 written in FORTRAN-IV for FACOM 230-75 is a part of the code series for evaluation of performance of the emergency core cooling system (ECCS) in pressurized water reactors according to the safety evaluation guidelines provided by the Atomic Energy Commission of Japan. ALARM-P1 is for analyzing the thermo-hydraulic phenomena during blowdown following a large break in the primary coolant system. ALARM-P1 models the PWR system fluid conditions including flow, pressure, mass inventory, fluid quality and heat transfer. It solves integral forms of fluid conservation and state equations for user-defined volumes treated as one-dimensional homogeneous, thermal-equilibrium elements with interconnecting flow paths and also finite difference forms of the one-dimensional heat conduction equations describing temperature profiles within solid material and the fluid-solid interface conditions. In addition, the ALARM-P1 provides the initial conditions for analysis of the last portion of the LOCA transient, a reflood phase, and the information for core heat-up analysis during the whole LOCA. This report describes the state-of-art methods and models of ALARM-P1 in June 1978 and gives information for users. (author)

  10. Scaling for hot leg break LBLOCA during post blowdown in the view of mass and energy release

    International Nuclear Information System (INIS)

    This study is on the discussion of scaling distortion using the SNUF for the hot leg break LBLOCA during post blowdown, based on the Ishii's three level scaling. The distortion effect was investigated through the experiment, and the mainly concerned dimensionless group is Froude number. Through the experiment, the violation of similarity in Froude number was revealed not to affect much on the overall system behavior. Thus, Ishii's three level scaling law was successfully applied and the experiment could be carried out without much loss of important information on system behaviors. However, for some parameters such as time scale, careful analysis is required. (author)

  11. A Main Steam Safety Valve (MSSV) With Fixed Blowdown According to ASME Section III,Part NC-7512

    International Nuclear Information System (INIS)

    In 1986, the NRC issued the Information Notice (IN) 86-05 'Main Steam Safety Valve test failures and ring setting adjustments'. Shortly after this IN was issued, the Code was revised to require that a full flow test has to be performed on each CL.2 MSSV by the manufacturer to verify that the valve was adjusted so that it would reach full lift and thus full relieving capacity and would re-close at a pressure as specified in the valve Design Specification. In response to the concern discussed in the IN, the Westinghouse Owners Group (WOG) performed extensive full flow testing on PWR MSSVs and found that each valve required a unique setting of a combination of two rings in order to achieve full lift at accumulation of 3% and re-closing at a blowdown of 5%. The Bopp and Reuther MSSV type SiZ 2507 has a 'fixed blowdown' i.e. without any adjusting rings to adjust the 'blowdown' so that the blowdown is 'fixed'. More than 1000 pieces of this type are successfully in nuclear power plants in operation. Many of them since about 25 years. Therefore it can be considered as a proven design. It is new that an optimization of this MSSV type SiZ 2507 fulfill the requirements of part NC-7512 of the ASME Section III although there are still no adjusting rings in the flow part. In 2000, for the Qinshan Candu unit 1 and 2 full flow tests were performed with 32 MSSV type SiZ 2507 size 8'' x 12'' at 51 bar saturated steam in only 6 days. In all tests the functional performance was very stable. It was demonstrated by recording the signals lift and system pressure that all valves had acceptable results to achieve full lift at accumulation of 3% and to re-close at blowdown of 5%. This is an advantage which gives a reduction in cost for flow tests and which gives more reliability after maintenance work during outage compared to the common MSSV design with an individual required setting of the combination of the two rings. The design of the type SiZ 2507 without any adjusting rings in the

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

    International Nuclear Information System (INIS)

    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

  13. Modification of blowdown heat transfer models for RELAP5-3D in accordance with appendix K of 10CFR50

    Energy Technology Data Exchange (ETDEWEB)

    Chin-Jang, Chang; Liang, T.K.S. [Nuclear Engineering Div. Institute of Nuclear Energy Research, Lung-Tan, Taiwan (China); Huan-Jen, Hung; Wang, L.C. [Power Research Institute, Taiwan Power Company (China)

    2001-07-01

    The objective of this paper is to implement the blowdown heat transfer models accepted by Appendix K of 10CFR50 into RELAP5-3D and to rename it as RELAP5-3D/K. Modifications of critical heat flux (CHF) model, post-CHF model, and the heat transfer logic for nucleate and transition boiling lockout are included. Also the assessments against separate-effect experiments were evaluated for RELAP 5-3D/K. From calculation results, the conservative predictions of surface peak temperatures using RELAP5-3D/K are obtained. It demonstrated that the blowdown heat transfer models were successfully modified and implemented into RELAP5-3D in accordance with Appendix K of 10CFR50. (authors)

  14. 蒸汽发生器排污管嘴压降研究及蒸汽发生器管板排污孔设置方式探讨%Research on AP1000 Steam Generator Blowdown Nozzle Pressure Drop and Blowdown Hole Configuration of Steam Generator Tube Sheet

    Institute of Scientific and Technical Information of China (English)

    王建平

    2015-01-01

    In order to improve the methodology of AP1000 Steam Generator blowdown nozzle pressure drop calculation, the paper provides the finite element model, simulates the blowdown flow path and the actual system conditions with STAR CCM+program. The paper provides the STAR CCM+ calculation result of blowdown nozzle drop (i.e. loss factor "K"). The paper also provides the simulation comparison result among three different configurations of blowdown path inside Steam Generator tube sheet.%本文通过对蒸汽发生器管板排污流道(包括排污管嘴)进行有限元建模分析,利用STAR CCM+程序模拟蒸汽发生器管板中实际流动情况,得到管嘴出口处压力值,为计算蒸汽发生器排污管嘴阻力系数K值提供有力支持。

  15. Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200 MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of

  16. Reactivity transients during a blowdown in a MSIV [Main Steam Isolation Valves] closure ATWS [Anticipated Transients Without Scram

    International Nuclear Information System (INIS)

    The objectives of this work are to study the consequences of the reactivity transients during a blowdown in an ATWS event with closure of the Main Steam Isolation Valves (MSIV), and to evaluate the effect of the LPCI (Low Pressure Coolant Injection) system and the sensitivity of plant response to the feedback coefficients. The present work was performed with the BNL Plant Analyzer (BPA). The BPA is a on-line, interactive BWR system code which models the non-homogeneous, non-equilibrium two-phase flow with a drift flux mixture model, the reactor kinetics with a point kinetic model, the thermal conduction with an integral method, and the control and plant protection systems with modern control theory. It also models the balance of plant (BOP) as well as the Mark I containment of a BWR/4. Thus, the BPA is a comprehensive engineering plant analyzer transients as well as accidents (e.g., ATWS and Small Break Loss of Coolant Accidents)

  17. 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)

    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

  18. Validation of a CATHENA fuel channel model for the post blowdown analysis of the high temperature thermal-chemical experiment CS28-1, I - Steady state

    International Nuclear Information System (INIS)

    To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal-chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H2 at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20-25 deg. C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using 'transparent' assumption for the CO2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA code

  19. LOFT LOCE transient thermal analysis for 6 in., 8 in., 10 in., and 12 in. primary coolant blowdown piping. Research, engineering, and construction report

    International Nuclear Information System (INIS)

    Several sections of the LOFT primary coolant blowdown piping were analyzed for temperature transients occurring during a Loss of Coolant Experiment (LOCE). The LOCE fluid conditions were chosen to conservatively represent the most severe operating conditions occurring in the piping. Temperature gradients will be used by the Applied Mechanics Branch to determine thermal stresses and the allowable thermal cycles for the piping. The only other significant thermal cycle (heat-up or cooldown) was not analyzed because the DTs for this cycle for the pipe sections analyzed will be small (less than 150F) and will have a very minor effect on the allowable number of thermal cycles. 8 inch-Sch 160, 10 inch-Sch 140, 12 inch-Sch 160, and a special 6 inch section of stainless steel piping were analyzed. The temperature gradients for each case were expressed in the DT form required for the ASME Section III pipe equations

  20. Ability of the TRAC-P1A computer program to predict blowdown, refill, and reflood phenomena during Semiscale Mod-1 experiments. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Demmie, P.N.

    1980-01-01

    A computer analysis of a Semiscale Mod-1 Loss-of-Coolant Experiment (LOCE) was performed using the TRAC-P1A computer program. The main purpose of this analysis was to contribute data for the assessment of the ability of TRAC-P1A to predict blowdown, refill, and reflood phenomena during a postulated Loss-of-Coolant Accident (LOCA). A TRAC-P1A Semiscale Mod-1 system model was created and TRAC-P1A was used to obtain initial conditions for Semiscale Mod-1 LOCE S-04-6. After this initialization, TRAC-P1A was used to simulate the first 60 seconds of this experiment. The results of this simulation are presented and discussed.

  1. System-level validation of CATHENA MOD-3.5D for early blowdown phase of large LOCA - RD-14M tests B0405-B0413

    Energy Technology Data Exchange (ETDEWEB)

    Gu, J.W.; McGee, G.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: guj@aecl.ca

    2008-07-01

    To investigate the integrated effect of multiple phenomena on CATHENA MOD-3.5d code uncertainty, for the early blowdown phase of large loss of coolant accident (LOCA), one RD-14M test series (B0405-B0413) is used to perform a system-level validation. The peak sheath temperature in the Fuel-Element-Simulator (FES) is selected as the key output parameter used to quantify the code bias and uncertainty in the validation. In the nine tests, the test conditions (break size, pump and power trip time, fluid sub-cooling and pressurizer isolation) are systematically varied and simulated, so that their effects on the magnitude and timing of the peak FES-sheath temperatures are demonstrated. The base test, B0405 is selected to perform sensitivity and uncertainty analyses. The sensitivity analyses show that the choice of film-boiling heat-transfer correlation has a significant effect on the prediction of the FES-sheath temperatures during the FES quenching period. Uncertainty analysis demonstrates a mean bias of about +20{sup o}C, with a range of about {+-}30{sup o}C to the upper and lower bounds. These results compare very well with the estimated code accuracy based on all nine tests of B0405-B0413. (author)

  2. Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments

    Energy Technology Data Exchange (ETDEWEB)

    Berta, V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-05-01

    Data from the Loss-of-Fluid Test (LOFT) Program help quantify the margin of safety inherent in pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). As early as 1979, questions arose concerning the accuracy of LOFT fuel rod cladding temperature data during several large-break LOCA experiments. This report analyzes how well externally-mounted fuel rod cladding thermocouples in LOFT accurately reflected actual cladding surface temperature during large-break LOCA experiments. In particular, the validity of the apparent core-wide fuel rod cladding quench exhibited during blowdown in LOFT Experiments L2-2 and L2-3 is studied. Also addressed is the question of whether the externally-mounted thermocouples might have influenced cladding temperature. The analysis makes use of data and information from several sources, including later, similar LOFT Experiments in which fuel centerline temperature measurements were made, experiments in other facilities, and results from a detailed FRAP-T6 model of the LOFT fuel rod. The analysis shows that there can be a significant difference (referred to as bias) between the surface-mounted thermocouple reading and the actual cladding temperature, and that the magnitude of this bias depends on the rate of heat transfer between the fuel rod cladding and coolant. The results of the analysis demonstrate clearly that a core-wide cladding quench did occur in Experiments L2-2 and L2-3. Further, it is shown that, in terms of peak cladding temperature recording during LOFT large-break LOCA experiments, the mean bias is 11.4 {plus_minus} 16.2K (20.5 {plus_minus} 29.2{degrees} F). The best-estimate value of peak cladding temperature for LOFT LP-02-6 is 1,104.8 K. The best-estimate peak cladding temperature for LOFT LP-LB-1 is 1284.0 K.

  3. 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

  4. Mathematical aspects of reactor blowdown

    International Nuclear Information System (INIS)

    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.)

  5. 基于MATLAB软件的蒸汽锅炉连续排污余热回收设计%Process Design of Waste Heat Recovery in Continuous Blowdown of Steam Boiler Based on MATLAB Software

    Institute of Scientific and Technical Information of China (English)

    刘舒佳; 关文吉; 刘伟; 冯圣红; 孙晓禹

    2016-01-01

    An energy-saving scheme is proposed to solve the waste heat and water resources in continuous blowdown of steam boiler. Matlab software is utilized to design calculation program for heat exchanger in heat recovery of steam boiler's continuous blowdown because of the complex problems in heat recovery design. Iterative methods are applied to obtain the heat gain of the deaerated water tank and softened water tank respectively. The heat exchange area of heat exchanger is calculated and investigated in different ways in the process of the waste heat recovery. The optimal plan for calculating heat exchange area is selected with comparison and analysis. The energy saving effect of the continuous waste heat recovery scheme for steam boiler is analyzed and calculated.%为解决蒸汽锅炉连续排污余热回收过程中废热及水资源浪费的问题提出一种节能方案.针对蒸汽锅炉连续排污余热回收过程中换热器设计计算较为复杂的问题,运用Matlab软件就蒸汽锅炉连续排污余热回收过程中换热器的设计计算编写程序,使用迭代的方式得出除氧水箱与软化水箱的分别得热量,探讨以不同的方式计算余热回收过程中换热器换热面积的问题;对比分析换热面积计算的2种方法,选取较优方案.分析计算提出的蒸汽锅炉连续排污余热回收方案的节能效果.

  6. Power Generation Systems Using Continuous Blowdown Waste Heat From Drum Boilers Driving an Organic Rankine Cycle%利用汽包锅炉连续排污余热的有机朗肯循环发电系统

    Institute of Scientific and Technical Information of China (English)

    刘强; 段远源; 万绪财

    2013-01-01

    A power generation system which used the continuous blowdown waste heat to drive an organic Rankine cycle (ORC) was developed to improve the energy efficiency. The blowdown waste heat was recovered by organic fluid, and then generates power by expansion through a turbine. The analysis model of thermal performance for the system was established. The performance of seven ORC working fluids including R227ea, RC318, R236ea, R245fa, R245ca, R123 and R113 were optimized using the GRG algorithm, and the maximum power output was obtained. The results show that the optimum turbine inlet temperature increases as the critical temperature of the working fluid decreases for the o2 cycle which has saturated vapor entering the turbine. However, the superheating in the o3 cycle reduces the waste heat utilization ratio. Supercritical ORC improves the match of temperature profiles between the heat source and the working fluid, which helps to increase the system power output. But the high operation pressure and heat transfer deterioration due to the large specific heat near the critical point must be considered in the system design. The thermal performance and the power output of R236ea are better than the six other fluids.%提出了一种利用汽包锅炉排污系统余热的有机朗肯循环发电系统,有机工质回收扩容器疏水的热量,并通过气轮机发电。建立了系统的热力性能分析模型,并对 R227ea、RC318、R236ea、R245fa、R245ca、R123和R113等7种工质的热力性能进行了优化。结果表明,临界温度高的工质,其 o2循环的最佳主气温度(蒸发温度)反而低;亚临界循环采用干流体时,过热不利于余热的利用;超临界循环可以改善热源与工质间的温度匹配,有利于增大系统输出功,但是其运行压力高、大比热区的传热恶化等问题是实际运行和设计需要考虑的因素;R236ea的热力性能优于其余6种工质。

  7. Dynamic loads by a blowdown incident DAISY-code of RPV-internals and primary loops of a PWR for the 0.1A-leck in the cold leg

    International Nuclear Information System (INIS)

    In this report the investigations of the dynamic loads for the RPV-internals and the primary loops of a pressurized water reactor with the fluid-structure code DAISY are documented for the case of a blowdown accident. For the first time the complete primary circuit of a PWR is modelled in a coupled code with regard to the fluid-structure interaction of the reactor internals with the surrounding water. In the performed calculations the consequences of a 0.1 A-leck, corresponding to the RSK-recommendations /RSK 81/, are investigated for the structures of the primary circuit and the RPV internals. After the exposition of the problems the scope of the work and the special problems are discussed. The applied computer technique for the fluid and structural part as well the coupling interface is explained. The model for the fluiddynamic part includes the four loops with the steamgenerators, the pumps and the pressure vessel with specific attention to a realistic modelling of the downcomer region. The structural model comprehends the RPV internals with particular emphasis on the core barrel. The required initial and boundary conditions and their realization is extensively discussed. The results of the different cases of initial and boundary conditions are presented and compared on diagrams. Finally the results are assessed and the influences of the simplifications and the assumptions are reviewed. The most important finding of this investigation is the fact that as consequence of a postulated 0.1 A-leak in the cold leg of a pressurized water reactor there is no risk for the structural integrity of all RPV internals and all components of the loops. (orig.)

  8. Pipe blowdown analysis using explicit numerical schemes

    International Nuclear Information System (INIS)

    Several explicit numerical schemes were investigated to solve the homogeneous equations of change for one-dimensional fluid flow and heat transfer. The most successful technique investigated is the alternating gradient method which is based on the two-step Lax-Wendroff procedure. Agreement with experimental results is very good. (U.S.)

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

    International Nuclear Information System (INIS)

    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 (CO2) 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), CO2 (model) and R134a (model) for depressurization in a pressure vessel were analyzed. (author)

  10. LOBI. Influence of PWR primary loops on blowdown. First results

    International Nuclear Information System (INIS)

    The LOBI test facility and the experimental programme are described with particular emphasis on the test facility design rationals and the programme objectives. The future small break test programme and the test facility modifications required for these tests are briefly referred to. The present status of the LOBI project is oultined by summarizing the four test series performed so far which were concerned with large break A1-tests, with small break scoping tests, and with large break A2- and B-tests in the framework of an interim test programme; all these tests were executed with the large downcomer. The fourth serie presently being performed is concerned with large break A1-tests with the small downcomer. Analysis results are presented of the comparison between prediction and experimental data of the very first LOBI test A1-04 which was used for a special, blind post-test prediction exercise with international participation (LOBI PREX)

  11. Blowdown transient for sodium-steam water SG for prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lele, H.G.; Srivastava, A.; Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K. [Reactor Safety Div., Bhabha Atomic Research Centre, Tromblay (India); Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Associate Director, Reactor Group, Chennai (India)

    2001-07-01

    Prototype Fast Breeder Test Reactor (PFBR) Steam Generator is once through steam generator in which water flows from bottom to top in 547 tubes, changing its state from highly subcooled to superheated state as it receives heat from sodium flowing from top to bottom in the shell side. Depressurization of steam generator from the dump valve provided at bottom is protective action. It prevents further possibility of water steam leak into sodium and subsequent sodium - water reaction. To perform depressurization transient analysis of PFBR appropriate thermal hydraulic modeling of SG is essential. Correct thermal hydraulic modelling needs simulation of sodium system, steam water system with different states from highly subcooled to superheated, coupling between sodium and steam-water system, SG tube and shell and different valve action. The computer code DPPFBR is developed with capability to simulate all these systems and phenomena encountered during transient. Different models of the code have been validated and code has been used for analysing depressurization transient. This paper describes various models used in the code and results of analysis for typical scenario. (author)

  12. A Blowdown Cryogenic Cavitation Tunnel and CFD Treatment for Flow Visualization around a Foil

    OpenAIRE

    伊藤,優; Ito, Yutaka; 沢崎, 和也; SAWASAKI, Kazuya; 谷, 直樹; Tani, Naoki; 長崎, 孝夫; NAGASAKI, TAKAO; 長島, 利夫; Nagashima, Toshio

    2005-01-01

    Cavitation is one of the major problems in the development of rocket engines. There have been fewexperimental studies to visualize cryogenic foil cavitation. Therefore a new cryogenic cavitation tunnel ofblowdown type was built. The foil shape is "piano-convex". This profile was chosen because of simplicity, butalso of being similar to the one for a rocket inducer impeller. Working fluids were water at room temperature,hot water and liquid nitrogen. In case of Angle of Attack (AOA)=8 °, perio...

  13. Simulation of the initial blowdown phase in a pressure suppression system (vent clearing and pool swell)

    International Nuclear Information System (INIS)

    The nonuniformity during vent clearing depends more on a local steam jet formation, the flow paths and the lengths of flow paths than on the flow resistance. For a more realistic simulation of power plants a stronger nodalisation with subdivision of single compartments into several nodes is necessary. In order to reduce the uncertainty in the prediction of pool swell, the model would require a calculation of pressure and velocity fields in the pool by an Euler or Lagrangemesh. This would require at least a 2-D model. A possible solution could be to determine the bubble expansion by a separate 2-D calculation and transmit it to the 1-D model. (orig./HP)

  14. Blowdown transient for sodium-steam water SG for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Test Reactor (PFBR) Steam Generator is once through steam generator in which water flows from bottom to top in 547 tubes, changing its state from highly subcooled to superheated state as it receives heat from sodium flowing from top to bottom in the shell side. Depressurization of steam generator from the dump valve provided at bottom is protective action. It prevents further possibility of water steam leak into sodium and subsequent sodium - water reaction. To perform depressurization transient analysis of PFBR appropriate thermal hydraulic modeling of SG is essential. Correct thermal hydraulic modelling needs simulation of sodium system, steam water system with different states from highly subcooled to superheated, coupling between sodium and steam-water system, SG tube and shell and different valve action. The computer code DPPFBR is developed with capability to simulate all these systems and phenomena encountered during transient. Different models of the code have been validated and code has been used for analysing depressurization transient. This paper describes various models used in the code and results of analysis for typical scenario. (author)

  15. Results of calculation of the dynamic behaviour of pressure suppression system during blowdown

    International Nuclear Information System (INIS)

    The computational model is based on several simplifications: The concrete parts of the containment are assumed to be rigid under the applied loadings, so that only the spherical shell with its annular condensation chamber will be investigated. As there is a plane of symmetry in the structure and in the loadings (and hence in the response as well) only half of the structure must be analyzed. A useful method to compute the behaviour of such a complex shell structure is the Finite Elements Method. Here the programme STRUDL-DYNAL was used, which has a linear, triangular shallow shell element with 5 degrees of freedom and with lumped inertia properties. In order to determine the necessary refinement of the discretization, the dynamic behaviour of the most important parts of the containment structure was analyzed individually. The computations showed that the lowest eigenfrequency of a simple shell may have a rather complex mode shape, e.g. a high circumferential order of cylindrical or conical shells and that higher frequencies may have simpler mode shapes. This behaviour requires a relative fine grid for discretization, as there must be sufficient degrees of freedom for the correct representation of the complex low modes. With respect to these effects, the structure was discretized by a spatial grid of 230 joints and 420 triangular finite elements. The resulting problem has about 1,200 degrees of freedom. The computation of the first 30 eigenfrequencies between 10 and 50 cps and of the corresponding mode shapes took about 75 min at 2,000 K memory size. There are some modes where the whole structure is vibrating; so at 10 cps the containment is vibrating like a vertically clamped beam; at 29 cps the structure goes up and down; at 32 cps horizontal cross sections are deformed elliptically. In addition there is a great number of modes with only parts of the structure vibrating at large amplitudes, especially the cylindrical and conical shell parts of the containment while the other parts remain quiet. The computed modes are a reliable basis to perform a transient analysis. The measured pressure distribution is used to derive a dynamic loading with interpolated time histories between the places where signals are measured. The computed response of the containment is compared with the corresponding data measured at Brunsbuettel. (orig./HP)

  16. Experimental and computed results for fluid-structure interactions with impacts in the HDR blowdown experiment

    International Nuclear Information System (INIS)

    This paper describes a new type of HDR experiment (V34) and compares the experimental results with the FLUX-code results. As novel feature, the core barrel is not rigidly clamped to the vessel as in earlier experiments but supported with gaps such that the core barrel can move freely upwards for about 2 mm and horizontally for 0.3 mm at the upper flange. At the lower core-barrel edge, snubbers restrict the horizontal motion to about + 1.4 mm and -2.8 mm. The experimental results show that the core barrel is deflected sidewards until it hits the snubber at the lower edge and then swings back to hit the opposite snubber. By this some kinetic energy is lost due to plastic snubber deformations. At the same time, the measurements show that the core barrel lifts rather uniformly from its support upwards until it hits the upper constraint. Several bounces up and down are observed until the core barrel becomes fixed probably due to friction from the side. This situation has been pre- and post-computed with the new FLUX-version which contains a very effective algorithm to treat supports with gaps and resultant impacts. For treatment of plastic supports, a simple model is added. Pre-computations were not meaningful because of large deviations in the pre-estimated initial gaps. However the computed pressure-field is not influenced very much by these parameters and predicted very well. This was favoured by the isothermal fluid initial conditions. Post-computations show sufficient agreement with respect to computed core barrel motion. The axial motion is described very well. Some problems remain which are due to the model for the upper flange support. Impacts do not result in greatly enlarged loadings, strains or accelerations for this situation. (orig./RW)

  17. Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

    International Nuclear Information System (INIS)

    Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

  18. Safety and Economy Analysis on Blowdown Method of Hydrogen Cooling System for Steam Turbo-Generator Sets%汽轮发电机氢气排污方法的安全性及经济性分析

    Institute of Scientific and Technical Information of China (English)

    祝艳平; 李西军; 吴红波; 杨红兵

    2012-01-01

    介绍了600MW汽轮发电机组氢气系统运行中经常发生的氢气纯度下降快、氢气持续微漏需要经常补氢或排污的问题,不当的氢气排污方法会造成氢气浪费、氢气纯度提升速度慢,还会给机组设备和人员的安全带来极大威胁.用计算、分析和试验验证的方法,研讨了补氢排污过程中氢气纯度和氢压的变化.分析计算和试验验证表明:在相同的氢气用量下连续式补氢排污的安全性和经济性要优于间断式补氢排污的方法.

  19. Nuclear steam generator

    International Nuclear Information System (INIS)

    A nuclear steam generator has a blowdown pump arranged to pump water from the blowdown line through a filter for return to the steam generator. The piping is arranged so that the pump may operate to reverse the direction of pumping through the blowdown line whereby reverse circulation may be established during wet lay up of the steam generator. A blower is arranged to withdraw nitrogen from an upper elevation in the steam generator and inject the nitrogen into the blowdown line in combination with the pumped reverse circulation during wet lay up. (author)

  20. 76 FR 72311 - Hazardous Waste Management System; Identification and Listing of Hazardous Waste; Final Exclusion

    Science.gov (United States)

    2011-11-23

    .... D001 Ignitability. D002 Corrosivity. D003 Reactivity. D007 Chromium. D008 Lead. D018 Benzene. D022... metals for eight samples for the RKI fly ash and RKI bottom ash, and RKI scrubber water blowdown; 2..., herbicides, dioxins/ furans, PCBs and metals for eight samples for the RKI scrubber water blowdown;...

  1. Department of Energy's team's analyses of Soviet designed VVERs

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document provides Appendices A thru K of this report. The topics discussed respectively are: radiation induced embrittlement and annealing of reactor pressure vessel steels; loss of coolant accident blowdown analyses; LOCA blowdown response analyses; non-seismic structural response analyses; seismic analyses; S'' seal integrity; reactor transient analyses; fire protection; aircraft impacts; and boric acid induced corrosion. (FI).

  2. 10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

    Science.gov (United States)

    2010-01-01

    ... 10 CFR part 50 or part 52 of this chapter. The guides on limiting conditions for operation for light..., floor and sample station drains), steam generator blowdown streams, chemical waste streams, low purity... paragraph C.1. 3 Such in-plant control measures may include treatment of steam generator blowdown...

  3. 直喷式直流电弧等离子化学气相沉积法金刚石单晶外延层制备研究%Growth of Diamond Single Crystal Epitaxial Layer by DC Arcjet Plasma Chemical Vapor Deposition at Blow-down Mode

    Institute of Scientific and Technical Information of China (English)

    刘杰; 黑立富; 陈广超; 李成明; 宋建华; 唐伟忠; 吕反修

    2014-01-01

    采用非循环直流喷射(直喷式)直流电弧等离子化学气相沉积法,在Ar/H2/CH4气氛下,成功制备了金刚石单晶外延层.试验采用的是3 mm×3 mm×1.2 mm的高温高压Ib型金刚石单晶衬底.研究了不同衬底温度和甲烷浓度对金刚石单晶外延层的形貌,速率和晶体质量的影响.采用光学显微镜,激光共聚焦表征了样品的形貌,利用千分尺测量其生长速率,利用Raman表征其晶体质量,采用OES诊断Ar/H2/CH4等离子气氛下C2、CH与Hβ的相对浓度.研究表明,温度和甲烷浓度对单晶刚石形貌和质量产生了明显的影响.在衬底为温度980℃,甲烷浓度在1.5%的条件下,生长速率达到了36 μm/h,并且晶体质量较好(半高宽仅为1.88 cm-1).同时发现生长参数对金刚石单晶外延层的生长模式有着显著地影响.

  4. Effects of residual scale inhibitors on the performance of reverse osmosis membrane in blow-down water from circulating water system%循环水排污水中残余阻垢剂对反渗透膜性能的影响

    Institute of Scientific and Technical Information of China (English)

    杨伟; 刘芳; 高雅; 闫茜; 张利

    2015-01-01

    循环水排污水中残余的阻垢剂会导致其水质的变化,从而影响反渗透膜性能。本文以循环水中常用的阻垢剂聚天冬氨酸(PASP)、羟基亚乙基=膦酸(HEDP)和氨基三亚甲基膦酸(ATMP)为研究对象,首先考察了它们的阻垢性能,然后在此基础上,通过静态浸泡试验和动态试验考察了它们的存在对反渗透膜性能的影响。研究结果表明,PASP、HEDP和ATMP中,PASP的阻垢性能最优,阻垢率高达84.21%,三者均会对反渗透膜的表面结构、组成成分、膜通量以及脱盐率产生一定的影响。当PASP、HEDP和ATMP的浓度分别为50mg/L、10mg/L和30mg/L时,在反渗透系统连续运行10h后,膜通量分别下降5.53%、4.89%和9.09%,小于空白时的18.95%;此外,脱盐率有不同程度的提高。%Residual scale inhibitors can cause water quality changes in circulating cooling process,thus affect the performances of reverse osmosis membrane. This study investigated the scale inhibition performance of scale inhibitors,PASP,HEDP and ATMP. The effects on the performance of reverse osmosis membrane were investigated by static test and dynamic test. Results showed that among PASP,HEDP and ATMP,the scale inhibition performances of PASP were best and scale inhibition rate could be as high as 84.21%. All the three materials had certain effects on the surface of the reverse osmosis membrane structure,composition and membrane flux. When the concentration of PASP, HEDP and ATMP was 50mg/L,10mg/L and 30mg/L respectively,the membrane flux decreased by 5.53%,4.89%,9.09%,less than 18.95% of the blank solution. In addition,desalination rates increased.

  5. 核电厂VVP101BA排放扩容器底部排污管线出口水温超标问题的处理%Treatment of VVP101BA Nuclear Power Plant Emissions Expanding Vessel Bottom Blowdown Pipeline Outlet Water Temperature Exceed the Standard Problem

    Institute of Scientific and Technical Information of China (English)

    张宪

    2015-01-01

    By raising the hydrophobic pipeline shaped water sealing bend door height and improve the cooling water water level discharge capacity expansion device, can guarantee the VVP101BA emission expanding vessel accumulate enough water, thereby cooling the hydrophobic effect. At the same time, modify the magnetic level gauge interface position, so as to solve the emissions expanding vessel magnetic type level gauge without reading problems.%通过抬高疏水管线门形水封弯管的高度和提高排放扩容器的冷却水水位高度,可保证VVP101BA排放扩容器积蓄足够的水量,从而达到冷却疏水的效果。同时,修改了磁力式翻板液位计接口的位置,从而解决了排放扩容器磁力式翻板液位计无读数的问题。

  6. 循环水排污水中杀菌剂、缓蚀阻垢剂对混凝效果的影响%Effects of Bactericides,Corrosion and Scale Inhibitors on Coagulation in Blow-down Water From Circulating Water System

    Institute of Scientific and Technical Information of China (English)

    杨伟; 刘芳; 樊丰涛; 张利

    2015-01-01

    When the blow‐down water from circulating water system is treated by coagulant‐sedimentation method ,the water treatment agents in it would have a certain effect on coagulation effect .For seeking influence of the agents on the coagulation process and its mechanism , the dosages of coagulant ,coagulant aid and operating conditions were optimized firstly ,and then the effects of single and compound water treatment agents on coagulation were investigated . In addition ,the effects of agents on optimum dosing amount of coagulant were investigated and flocculation obtained under different conditions was analyzed by SEM .The results showed that the optimized dosages of PAC and PAM were 40 mg/L and 0.8 mg/L ,respectively ,and the optimal operating conditions were fast stirring for 3 min at 300 r/min and slow stirring for 10 min at 90 r/min ,under which the coagulation effect of blow‐down water was best and turbidity removal rate reached 93.49% .Polyaspartic acid (PASP) had a greatest influence on coagulation of blow‐down water and residual turbidity raged from 0.85 to 1.78 nephelometric turbidity unit (NTU ) , w hile dodecyl dimethyl benzyl ammonium chloride (1227) had a greatest influence on the optimum amount of PAC ,which ranged from 20 to 70 mg/L .Residual turbidity was greater than 0.98 NTU at the existence of compound agents . According to SEM photos , the difference in microstructures of flocculation from blow‐down water at different coagulation conditions led to different coagulation effect .%混凝沉淀法处理循环水排污水时,由于排污水中残余药剂的影响,混凝往往达不到理想效果。为了寻求药剂对混凝过程及机理的影响,首先对混凝剂(PAC)、助凝剂(PAM )的投加量以及混凝条件进行优化,然后在优化的混凝条件下,考察单体药剂以及复配药剂对排污水混凝效果的影响,还考察了药剂对 PAC最佳投加量的影响,并采用扫描电镜观测不同条件下絮凝体的微观结构。结果表明,PAC 和 PAM 的最佳投加量分别为40 mg/L、0.8 mg/L ,最佳混凝条件为快速搅拌时间3 min、慢速搅拌速率90 r/min、快速搅拌速率300 r/min、慢速搅拌时间10 min;在此条件下,排污水浊度去除率可达93.49%。PASP对排污水混凝效果影响最大,且剩余浊度波动性大,波动范围0.85~1.78 NTU (Nephelometric Turbidity Unit);1227对 PAC最佳投加量影响最大,波动范围20~70 mg/L。排污水存在复配药剂时,其混凝后得剩余浊度都大于无药剂时的0.98 N T U。不同条件下的排污水混凝所得絮凝体结构存在明显的差异,从而导致了混凝效果的不同。

  7. 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,...

  8. 40 CFR 98.250 - Definition of source category.

    Science.gov (United States)

    2010-07-01

    ..., naphtha, kerosene, distillate fuel oils, residual fuel oils, lubricants, or asphalt (bitumen) through...; asphalt blowing operations; blowdown systems; storage tanks; process equipment components (compressors..., tanker truck, and similar loading operations; flares; sulfur recovery plants; and non-merchant...

  9. APPLICATION OF MEMBRANE TECHNOLOGY TO POWER GENERATION WATERS

    Science.gov (United States)

    Three membrane technlogies (reverse osmosis, ultrafiltration, and electrodialysis) for wastewater treatment and reuse at electric generating power plants were examined. Recirculating condenser water, ash sluice water, coal pile drainage, boiler blowdown and makeup treatment waste...

  10. LOFT reactor vessel 290/sup 0/ downcomer stalk instrument penetration flange stress analysis

    Energy Technology Data Exchange (ETDEWEB)

    Finicle, D.P.

    1978-06-06

    The LOFT Reactor Vessel 290/sup 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.

  11. A demonstration experiment of steam-driven, high-pressure melt ejection

    International Nuclear Information System (INIS)

    A steam blowdown test was performed at the Surtsey Direct Heating Test Facility to test the steam supply system and burst diaphragm arrangement that will be used in subsequent Surtsey Direct Containment Heating (DCH) experiments. Following successful completion of the steam blowdown test, the HIPS-10S (High-Pressure Melt Streaming) experiment was conducted to demonstrate that the technology to perform steam-driven, high-pressure melt ejection (HPME) experiments has been successfully developed. In addition, the HIPS-10S experiment was used to assess techniques and instrumentation design to create the proper timing of events in HPME experiments. This document discusses the results of this test

  12. Secondary coolant purification system with demineralizer bypass

    International Nuclear Information System (INIS)

    Apparatus and method are provided for a nuclear stream supply system for adequately controlling the chemistry of the secondary coolant. The invention includes means for the addition of volatile chemicals, a full flow condensate demineralizer, continuous blowdown capability, radiation detection means, a condensate demineralizer bypass line, and an auxiliary demineralizer bypass line, and an auxiliary demineralizer sized to handle full blowdown flow. The auxiliary demineralizer is cut into the system and the steam generator feedwater flow is bypassed around the full flow condensate demineralizer whenever radioactivity is detected in the secondary coolant

  13. Code development and analysis program. RELAP4/MOD7 (Version 2): user's manual

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-08-01

    This manual describes RELAP4/MOD7 (Version 2), which is the latest version of the RELAP4 LPWR blowdown code. Version 2 is a precursor to the final version of RELAP4/MOD7, which will address LPWR LOCA analysis in integral fashion (i.e., blowdown, refill, and reflood in continuous fashion). This manual describes the new code models and provides application information required to utilize the code. It must be used in conjunction with the RELAP4/MOD5 User's Manual (ANCR-NUREG-1335, dated September 1976), and the RELAP4/MOD6 User's Manual (CDAP-TR-003, dated January 1978).

  14. 40 CFR 1700.5 - Discharges not requiring control.

    Science.gov (United States)

    2010-07-01

    ... Marine Pollution Control Device to mitigate adverse impacts on the marine environment: (a) Boiler Blowdown: the water and steam discharged when a steam boiler is blown down, or when a steam safety valve is... originates from port facilities. (k) Stern Tube Seals and Underwater Bearing Lubrication: the seawater...

  15. 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.

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

    International Nuclear Information System (INIS)

    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

  17. 40 CFR 471.33 - New source performance standards (NSPS).

    Science.gov (United States)

    2010-07-01

    ... operator demonstrates, on the basis of analytical methods set forth in or approved pursuant to 40 CFR part... control scrubber blowdown. Subpart C—NSPS Pollutant or pollutant property Maximum for any 1 day Maximum... times. (bb) Hydrostatic tube testing and ultrasonic testing wastewater—Subpart C—NSPS. There shall be...

  18. Dynamic behavior of pipe under loss of coolant accident

    International Nuclear Information System (INIS)

    This paper consists of six chapters. In the chapter one, a survey of past researches in the field of pipe whip and blowdown thrust force is described. The purpose of this paper is also included. In the chapter two, the aim and specifications of Pipe Rupture Test Facility installed at JAERI is described. In the chapter three, the results of blowdown thrust force experiments and the estimation of analytical method regarding blowdown thrust force using RELAP4 and its post processor BLOWDOWN code. In the chapter four, the results of pipe whip experiments using 4, 6, and 8 inch test pipes and pipe whip restraints under BWR LOCA conditions. The influence of clearance, overhang length and diameter of test pipes on the pipe whip behavior is described. In the chapter five, the results of pipe whip analyses using a general purpose finite element program ADINA are described. The maximum restraint force due to impact of a ruptured pipe and pipe whip restraints can be predicted by this method. In addition, it is concluded that a simplified analysis method based on an energy balance is useful in order to determine the limit of the overhang length. In the chapter six, is described an example of application of the simplified method to real BWR plant piping and the whole conclusions. (author)

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

    NARCIS (Netherlands)

    Louwerse, Marcus; Jansen, Henri; Elwenspoek, Miko

    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 use

  20. 77 FR 14010 - Millennium Pipeline Company, LLC; Notice of Availability of the Environmental Assessment for the...

    Science.gov (United States)

    2012-03-08

    ... coolers, unit blowdown silencers, a filter-separator with a liquids tank, and an emergency electrical power generator. Pipeline facilities required for the project include approximately 545 feet of new 36..., Washington, DC 20426. Any person seeking to become a party to the proceeding must file a motion to...

  1. 75 FR 58315 - Hazardous Waste Management System; Identification and Listing of Hazardous Waste; Direct Final...

    Science.gov (United States)

    2010-09-24

    ... hazardous wastes until excluded. See 66 FR 27266 (May 16, 2001). III. EPA's Evaluation of the Waste... Company at 62 FR 37694 (July 14, 1997) and 62 FR 63458 (December 1, 1997) where the delisted waste leached... disposal scenario for Eastman's RKI scrubber water blowdown. EPA applied the DRAS described in 65 FR...

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

    International Nuclear Information System (INIS)

    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

  3. Compounding disturbance interactions in a Southern Rocky Mountain subalpine forest

    Science.gov (United States)

    Caldwell, M. K.; Wessman, C. A.; Buma, B.; Poore, R.

    2014-12-01

    Landscape disturbances are important shaping agents to ecosystem processes, services and structure. When multiple disturbances occur, they create novel ecosystem trajectories. It is unknown what happens to ecosystem resiliency and services, such as carbon storage, when multiple disturbances occur in a short time period. Routt National Forest, Colorado is a subalpine forest which experienced multiple disturbances including a blowdown (1997), logging (1999-2001), fire (2002) and insects (spruce and pine beetle, multiple years). The objective of this study is to determine recovery patterns post- disturbance as they pertain to resilience and carbon storage. Recovery from a single landscape disturbance for individual species typically have a predictable response. In order to study recovery from multiple disturbances, we measured plots in 2010-2013 across the multiple disturbances. Further, we simulated plots to the year 2113 using the Forest Vegetation Simulator to quantify carbon storage. Our sampling design captured disturbance interactions, where we considered 1. fire, 2. blowdown + fire, with a gradient across blowdown severity, 3. blowdown + logging + fire, 4. beetle-kill, 5. logged + beetle kill, 6. blowdown + beetle kill, 7. logged + blowdown, and 8. control. We counted species, diameter and height of each tree within the 162 (15x15m) square plots. Results between fire and other disturbances varied by individual species. Lodgepole pine regeneration was strongly driven by other disturbances along a severity gradient. Logging prior to fire seems to create varying abiotic conditions, increasing lodgepole seedling density post-fire. Engelmann spruce regeneration was linked to the presence of aspen post-fire + other disturbances, a function of shade provided by aspen. In turn, soil moisture drives aspen regeneration. Incoming aspen seedlings aid carbon storage, recovering to pre-fire between 60-80 years, post disturbance. Upon preliminary analysis, those plots absent

  4. 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.

  5. Assessment of RELAP5/MOD3/K using semi scale test S-02-4

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sun Tack; Choi, Han Rim; Huh, Jae Yong; Lee, Nam Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report presents the results of the RELAP5/MOD3/K assessment utilizing a Semi scale large break loss-of-coolant experiment Test S-02-4. Blowdown heat transfer test S-02-4 is a 200 % double ended cold leg break experiment performed in Semi scale Mod-1 facility in 1975 for the purpose of investigating the thermal and hydraulic phenomena accompanying a hypothetical large break LOCA in a pressurized water reactor system. Through comparisons between data and best-estimate RELAP5 calculation, the capabilities of RELAP5 to calculate the large break loss-of-coolant accident (LOCA) were assessed. Emphasis was placed on the capability of the code to calculate break flow rates during system blowdown phase and the peak cladding temperature (PCT) behavior. 37 figs., 2 tabs., 7 refs. (Author) .new.

  6. Comparison of an integral response scaling method with Ishii's scaling method and its validation using RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    An integral response scaling method for a reduced-height test facility is suggested and the scaling laws derived from it are compared with Ishii's scaling. In the present scaling method it turns out that flow velocities in the vertical channel and through the break area or injection area should be preserved. RELAP5/MOD3.2 code calculations of pot-boiling, blowdown, heat transfer in Steam Generator(SG) and off-take are conducted for the validation of the present scaling method. Four scaled-down models are designed based on the present method and Ishii's scaling method given length and area scales of 1/5 and 1/100, respectively. RELAP5/MOD3.2 calculations show that the scaled-down model based on the present scaling method well maintains the similarity of the nondimensional mixture level in pot-boiling, the nondimensional pressure in blowdown and the heat transfer coefficient in SG

  7. Experimental investigation on unsteady pressure fluctuation of rotor tip region in high pressure stage of a vaneless counter-rotating turbine

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    An experimental investigation has been performed to study the unsteady pressure fluctuation of rotor tip region in high pressure stage of a vaneless counter-rotating turbine.The experiment is carried out on a blow-down short duration turbine facility.The investigation indicates that the blow-down short duration turbine facility is capable of substituting continuous turbine facilities in most turbine testing.Through this experimental investigation,a distinct blade-to-blade variation is observed.The results indicate that the combined effects of vane wake,tip leakage flow,complicated wave systems and rotor wake induce the remarkable blade-to-blade variations.The results also show that the unsteady effect is intensified along the flow direction.

  8. Chemical control in steam systems by using a stabilized inorganic product with gain of energy and speed in detecting contaminations; Controle quimico em geradores de vapor, pelo uso de agente inorganico estabilizado, com ganhos de energia e celeridade na deteccao de contaminacoes

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Barny de; Pereira, Renato Andre Nunes [Kurita do Brasil, Rio de Janeiro, RJ (Brazil)

    2010-07-01

    This paper shows the basic conditions to control the relation between phosphate and sodium in high pressure boilers by applying a stabilized chemical product ensuring operation with low variability and energy gain by the eliminating of corrective blowdown. It presents the routine and the relevant benefits provided by a strong monitoring program of phosphate application in high pressure boilers as an important tool do detect deviations and to get better control of silica solubilization in this pressure level. (author)

  9. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    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

  10. Depressurization of Vertical Pipe with Temperature Gradient Modeled with WAHA Code

    OpenAIRE

    Oriol Costa; Iztok Tiselj; Leon Cizelj

    2012-01-01

    The subcooled decompression under temperature gradient experiment performed by Takeda and Toda in 1979 has been reproduced using the in-house code WAHA version 3. The sudden blowdown of a pressurized water pipe under temperature gradient generates a travelling pressure wave that changes from decompression to compression, and vice versa, every time it reaches the two-phase region near the orifice break. The pressure wave amplitude and frequency are obtained at different locations of the pipe's...

  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. Loss of coolant accident analysis and evolution of emergency core cooling system for an inpile irradiation facility

    International Nuclear Information System (INIS)

    This paper deals with the Loss of Coolant Accident (LOCA) analysis of an inpile facility using RELAP4/MOD6 computer code. The present study is the culmination of a three part LOCA analysis done earlier by the authors. Blowdown analysis had been extended to include reflood part of the transient. Based on the analysis an Emergency Core Cooling System (ECCS) has been evolved. (author). 5 figs., 2 tabs

  13. Details on spot remote sensing satellite propulsion unit

    Science.gov (United States)

    Corai, J. C.

    1984-03-01

    The SPOT propulsion system is described. The SPOT platform includes a propulsion module equipped with a hydrazine reaction control system (RCD). This RCS belongs to the attitude and orbit control system. It comprises essentially two branches of catalytic thrusters fed by surface tension tanks through a circuit providing their interconnection by latching insulation valves. It operates in blow-down mode. Each equipment is qualified individually.

  14. Analysis of pressure safety valves for fire protection on offshore oil and gas installations

    DEFF Research Database (Denmark)

    Bjerre, Michael Skov; Eriksen, Jacob; Andreasen, Anders;

    2016-01-01

    The effectiveness of fire Pressure Safety Valves (PSV) has been investigated when offshore process equipment is exposed to a fire. Simulations of several typical offshore pressure vessels have been performed using the commercial software VessFire. The pressure vessels are exposed to a small jet f...... protection for typical offshore fire scenarios and that blowdown valves and passive fire protection should be considered as alternatives....

  15. Review on Recent Advances in Pulse Detonation Engines

    OpenAIRE

    Pandey, K. M.; Pinku Debnath

    2016-01-01

    Pulse detonation engines (PDEs) are new exciting propulsion technologies for future propulsion applications. The operating cycles of PDE consist of fuel-air mixture, combustion, blowdown, and purging. The combustion process in pulse detonation engine is the most important phenomenon as it produces reliable and repeatable detonation waves. The detonation wave initiation in detonation tube in practical system is a combination of multistage combustion phenomena. Detonation combustion causes rapi...

  16. Pulse Detonation Rocket Engine Research at NASA Marshall

    Science.gov (United States)

    Morris, Christopher I.

    2003-01-01

    Pulse detonation rocket engines (PDREs) offer potential performance improvements over conventional designs, but represent a challenging modeling task. A quasi 1-D, finite-rate chemistry CFD model for a PDRE is described and implemented. A parametric study of the effect of blowdown pressure ratio on the performance of an optimized, fixed PDRE nozzle configuration is reported. The results are compared to a steady-state rocket system using similar modeling assumptions.

  17. Preliminary experiment and analysis of supersonic inlet buzz

    OpenAIRE

    Hongprapas, Sorarat

    1996-01-01

    The inlet buzz phenomenon was investigated experimentally and analytically. An external-compression axisymmetric inlet model, having 74 mm in cowl lip diameter, was tested in the 229x229 sq mm blowdown wind tunnel at Mach 2.4. The test facility has shown potential for the supersonic inlet research. The occurrence of inlet buzz was indicated by the continuous shock oscillation and the static pressure fluctuation. The hypothesis based on internal pressure measurement and shadowgr...

  18. Multiple disturbance interactions and drought influence fire severity in Rocky Mountain subalpine forests

    OpenAIRE

    C. Bigler; D. Kulakowski; T. T. Veblen

    2005-01-01

    Disturbances such as fire, insect outbreaks, and blowdown are important in shaping subalpine forests in the Rocky Mountains, but quantitative studies of their interactions are rare. We investigated the combined effects of past disturbances, current vegetation, and topography on spatial variability of the severity of a fire that burned approximately 4500 ha of subalpine forest during the extreme drought of 2002 in northwestern Colorado. Ordinal logistic regression was used to spatially model f...

  19. Combined effects experiments with the condensation pool test facility

    Energy Technology Data Exchange (ETDEWEB)

    Puustinen, M. [Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland)

    2007-01-15

    This report summarizes the results of the condensation pool experiments in spring 2006, where steam and steam/air mixture was blown into the pool through a DN200 blowdown pipe. Altogether three experiments, each consisting of several blows, were carried out with a scaled down test facility designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to study the effects of non-condensable gas present in the discharge flow. Particularly pressure pulses inside the blowdown pipe and at the pool bottom caused by chugging were of interest. The test pool was an open stainless steel tank with a wall thickness of 4 mm and a bottom thickness of 5 mm containing 15 m3 of water. The nearby PACTEL test facility was used as a steam source. During the experiments the initial pressure of the steam source was 0.5 MPa and the pool water bulk temperature ranged from 40 C to 70 C. The test facility was equipped with high frequency instrumentation for capturing different aspects of the investigated phenomena. The data acquisition program recorded data with the frequency of 10 kHz. A digital high-speed video camera was used for visual observation of the pool interior. Air, in quantities even less than 1 %, reduced the condensation rate considerably. The high pressure pulses registered inside the blowdown pipe due to water hammer propagation during chugging almost disappeared when the combined discharge period of steam and air started. With noncondensable gas fractions above 3 % the damping of pressure oscillations inside the blowdown pipe was practically complete. Air quantities in the vicinity of 2 % started to have an effect also on the oscillations measured by the pressure sensor at the pool bottom. Both the amplitude and frequency of the pressure pulses decreased considerably. The experiments demonstrated that even small quantities of noncondensable gas can have a strong diminishing effect on pressure oscillations and structural loads

  20. The development and application of overheating failure model of FBR steam generator tubes

    International Nuclear Information System (INIS)

    The following items have been studied to evaluate overheating failure of FBR steam generator heat transfer tubes: 1) To establish a structural integrity analysis method, 2) To improve and validate blow down analytical method, 3) To quantitatively validate the entire overheating analysis model by sodium water reaction data. Based on the above studies, the analytical method was applied to PFR superheater leak event and the Monju steam generator accidental analysis. The followings were quantitatively shown through the analysis: 1. The most important cause that multi-tube failure occurred in the 1987 PFR superheater-2 leak is that the superheater did not equip a fast steam dump system at the time of the leak event. 2. Overheating failure will not occur under any operational conditions of Monju in both steady state and transient phases such as water/steam blow-down. 3. Although safety margin becomes small when the water/steam flow rate becomes small during the blow-down, the modification of the plant such as hastening blow-down by equipping more relief valves will drastically improve the safety margin. (J.P.N.)

  1. Determination of the thermal transport delay characteristics of a heater-rod-thermocouple system used for measuring the time-to-critical heat flux

    International Nuclear Information System (INIS)

    Fast response thermocouple installations were used to measure time-to-CHF during the rod bundle test phase of the C-E EPRI Blowdown Heat Transfer Program. The CHF measured by these thermocouple installations occurs following the simulation of a complete rupture and offset of the inlet piping of a Pressurized Water Reactor during a Loss of Coolant Accident. The thermocouples were installed in ceramic cylinders within fuel rod simulators which were heated by passing direct current through thier walls. Such blowdown tests subjected the thermocouples to rapid heatup rates of from 300 to 5000F/sec. starting within approximately one second from the time of simulated rupture. The primary elements contributing to the heat transport delay for the system composed of the heater rod and the thermocouple are the clearance between the rod wall and the ceramic and the thermocouple time constant.An analytical model was developed in conjuction with an iterative non-linear least squares fitting technique which allowed the determination of these two transport delays terms from calibration tests data. A heater rod was subjected to a current pulses while hanging in air. Transient temperature profiles during these pulse tests were fit to the closed form analytical equation by varying the gap size and the thermocouple time constant. The transport time lag terms derived from fitting the calibration test data for a known power pulse could be applied to determine the actual time-to-CHF and post-CHF heatup rate from measured blowdown test data

  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. Condensation pool experiments with non-condensable gas and Fluent 5 simulations

    International Nuclear Information System (INIS)

    The formation, size and distribution of non-condensable gas bubbles in the condensation pool of the Olkiluoto nuclear power plant (NPP) in a conceivable loss-of-coolant accident (LOCA) was studied experimentally with a scaled down condensation pool test rig. Particularly, it was important to find out if any air bubbles flowed inside the emergency core cooling system (ECCS) strainer close to the pool wall and bottom. The effect of non-condensable gas on the performance of an ECCS pump was also examined. Computational fluid dynamics (CFD) calculations with the Fluent 5 code were made to support the design of the test rig and the planning of the experiments. Compressed air was blown to the test pool through blowdown pipes or, alternatively, air was injected directly into the intake pipe of the ECCS pump. The first large air bubbles forming at the blowdown pipe outlet touched the ECCS strainer. When two blowdown pipes were used simultaneously, a lot of air bubbles were detected inside the strainer during the first 30 seconds. A 3-7 % volume fraction of air injected directly into the pump intake pipe was enough to make the pump head and flow collapse. (orig.)

  4. Three-dimensional thermal-hydraulic response in LBLOCA based on MARS-KS calculation

    International Nuclear Information System (INIS)

    Three-dimensional (3D) thermal-hydraulic analysis of an accident in Nuclear Power Plant (NPP) has been extended to use since Best-Estimate (BE) calculation was allowed for safety analysis. The present study is to discuss why and how big differences can be obtained from the 1D and 3D thermal-hydraulic calculations for large break Loss-of-Coolant Accident (LBLOCA). Calculations are performed using MARS-KS code with one-dimensional (1D) modeling and with 3D modeling for reactor vessel of Advanced Power Reactor (APR1400). For the 3D modeling, the MULTI-D component of the MARS-KS code is applied. Especially, a hot channel having a size of one fuel assembly is also simulated. From the comparison of the calculation results, four differences are found: lower blowdown Peak Cladding Temperature (PCT) in 3D calculation, instantaneous stop of cladding heat-up, extent of blowdown quenching, and milder and longer reflood process in 3D calculation. The flow distribution in the core in 3D calculation could be one of the reasons for those differences. From the sensitivity study, the initial temperature at the reactor vessel upper head is found to have strong effect on the blowdown quenching, thus the reflood PCT and needs a careful consideration. (author)

  5. Pressure suppression pool mixing in passive advanced BWR plants

    International Nuclear Information System (INIS)

    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

  6. The Improvement of Plant Efficiency by Testing and Revising of the Reactor Thermal Power Calculation Program

    International Nuclear Information System (INIS)

    Since the uncertainty of flow measurement mostly affects the result of reactor thermal power calculation, reactor power in most of Nuclear Power Plants(NPPs) is controlled by excore Nuclear Instrumentation System(NIS) based on SPPC which has less uncertainty of flow measurement by using venture-meter. Real time monitoring system for reactor thermal power of Kori unit 3 and 4 has been established since 1992, and plant efficiency was improved by detecting errors and revising the program in 2012 following the engineering judgement that reactor thermal power varies according to steam generator blowdown flow change, unit conversion constant, and thermal expansion coefficient, etc. The reactor thermal power calculation program for Kori unit 3 and 4 was developed in 1992 and operated for 20 years without any correction or revision. Based on the engineering judgement that reactor thermal power varies according to change of steam generator blowdown flow, we conducted a research and found a couple of errors in steam generator blowdown specific volume, unit conversion constants for differential pressure of main feed water inlet flow, and thermal expansion coefficient of venture-meter which measures main feed water flow for steam generator. By correcting the errors in reactor thermal power program, generator power increased by 3.2 MWe for two units, Kori 3 and 4. Considering recent capacity factor of the plant, additional net electricity of 26,434 MWh was produced annually

  7. Mark III confirmatory test program: one third scale, three vent air tests

    International Nuclear Information System (INIS)

    A series of air blowdown tests was run to evaluate pool swell phenomena for the Mark III pressure suppression containment concept. The tests were performed at the Pressure Suppression Test Facility which consists of an integrated system of drywell, vent system, and suppression pool. The volumetric scale factor used for facility design was nominally 1:130, based on the BWR 6/251 series Mark III containment design. The pool and vent system both represented one-third scale mockups of an 8-degree sector of the Mark III containment, including a vertical row of three 157/8 in. (403 mm)-diameter horizontal vents. Test parameters changed were blowdown flow restrictor size and top vent centerline submergence. The transient responses of the pressurizer, drywell, vent system, suppression pool, and wetwell air space were measured and analyzed for use in formulating and/or further confirming the analytical models used for predicting loss-of-coolant accident transients. Results supported previously reported conclusions. Air blowdown tests with comparable drywell pressure transients were shown to have somewhat higher pool swell velocities than previously reported steam tests. The air tests provided additional evidence that bubble breakthrough elevation is not dependent upon charging rate but is determined almost exclusively by initial vent submergence. Total impulse values on the pool ceiling for the air tests were found to be lower than comparable steam tests

  8. Vibration phenomena in large scale pressure suppression tests

    International Nuclear Information System (INIS)

    Structure and fluid vibration phenomena were observed during blow-down experiments simulating a LOCA in the GKSS full scale multivent pressure suppression test facility. The source related excitations during the two regimes of condensation oscillation and of chugging are described together 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 blow-down 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 modes of two of the wetwell's walls when coupled by the wetwell pool. Further peaks found during blow-down in the spectra of signals at higher frequencies correspond to global vibration modes of the wetwell. (author)

  9. PPOOLEX experiments on dynamic loading with pressure feedback

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-15

    This report summarizes the results of the dynamic loading experiments (DYN series) carried out with the scaled down, two compartment PPOOLEX test facility designed and constructed at LUT. Steam was blown into the 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 experiments was to study dynamic loads caused by different condensation modes. Particularly, the effect of counterpressure on loads due to pressure oscillations induced by chugging was of interest. Before the experiments the condensation pool was filled with isothermal water so that the blowdown pipe outlet was submerged by 1.03-1.11 m. The initial temperature of the pool water varied from 11 deg. C to 63 deg. C, the steam flow rate from 290 g/s to 1220 g/s and the temperature of incoming steam from 132 deg. C to 182 deg. C. Non-condensables were pushed from the dry well into the gas space of the wet well with a short discharge of steam before the recorded period of the experiments. As a result of this procedure, the system pressure was at an elevated level in the beginning of the actual experiments. An increased counterpressure was used in the last experiment of the series. The diminishing effect of increased system pressure on chugging intensity and on measured loads is evident from the results of the last experiment. The highest pressure pulses both inside the blowdown pipe and in the condensation pool were about half of those measured with a lower system pressure but otherwise with similar test parameters. The experiments on dynamic loading gave expected results. The loads experienced by pool structures depended strongly on the steam mass flow rate, pool water temperature and system pressure. The DYN experiments indicated that chugging and condensation within the blowdown pipe cause significant dynamic loads in case of strongly sub-cooled pool water. The level of pool water temperature is decisive

  10. PPOOLEX experiments on dynamic loading with pressure feedback

    International Nuclear Information System (INIS)

    This report summarizes the results of the dynamic loading experiments (DYN series) carried out with the scaled down, two compartment PPOOLEX test facility designed and constructed at LUT. Steam was blown into the 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 experiments was to study dynamic loads caused by different condensation modes. Particularly, the effect of counterpressure on loads due to pressure oscillations induced by chugging was of interest. Before the experiments the condensation pool was filled with isothermal water so that the blowdown pipe outlet was submerged by 1.03-1.11 m. The initial temperature of the pool water varied from 11 deg. C to 63 deg. C, the steam flow rate from 290 g/s to 1220 g/s and the temperature of incoming steam from 132 deg. C to 182 deg. C. Non-condensables were pushed from the dry well into the gas space of the wet well with a short discharge of steam before the recorded period of the experiments. As a result of this procedure, the system pressure was at an elevated level in the beginning of the actual experiments. An increased counterpressure was used in the last experiment of the series. The diminishing effect of increased system pressure on chugging intensity and on measured loads is evident from the results of the last experiment. The highest pressure pulses both inside the blowdown pipe and in the condensation pool were about half of those measured with a lower system pressure but otherwise with similar test parameters. The experiments on dynamic loading gave expected results. The loads experienced by pool structures depended strongly on the steam mass flow rate, pool water temperature and system pressure. The DYN experiments indicated that chugging and condensation within the blowdown pipe cause significant dynamic loads in case of strongly sub-cooled pool water. The level of pool water temperature is decisive

  11. Experimental investigations of BWR pressure suppression pool behavior under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    The experiments discussed in this paper look into different processes which may occur during a loss-of-coolant accident in the pressure suppression pool of a Boiling Water Reactor (BWR). These processes include: a) development of a thermal stratification, b) bubble dynamics and related water flow during continuous release of air and c) air blowdown and associated water slug phenomenon in the water pool. The experiments have been performed in the THAI test facility, which is a cylindrical vessel of 9.2 m height, 3.2 m diameter and with a gas volume of 60 m3. The variation in the investigated test parameters included, steam and air mass flux, initial water pool temperature, blowdown pressures, downcomer submergence, etc. A systematic variation of the test parameters allowed better understanding of the phenomena. Experiments discussed in this paper were performed with a vertical downcomer of 0.1 m diameter and 2 m submergence depth in the water pool. For the blowdown experiments, a separate interconnecting vessel of 1 m3 volume was used to inject air at pressures between 3 bar and 10 bar. A high speed camera (1000 fps) was installed to visualize the formation and propagation of air bubbles in the suppression pool and the resulting pool swelling phenomena. Customized instrumentation applied during the tests included grids of densely spaced thermocouples and of pressure transducers at various locations in order to capture the temperature distribution in the pool and the water slug induced pressure loadings, respectively. The present paper discusses the main outcome of the selected experiments. On the whole the experimental data may be very useful for code validation. (authors)

  12. Effect of pipeline rupture transient release modelling on predicted consequences

    Energy Technology Data Exchange (ETDEWEB)

    Johnston, C.R.; Springer, W.A.J.; Rowe, R.D. [Calgary Univ., Dept. of Mechanical Engineering, Calgary, AB (Canada)

    1998-09-01

    A mathematical model was developed to predict the consequences of a rupture in a natural gas pipeline. The model was a real-fluid, non-isentropic blowdown (RFB) model. A comparison of this model and the widely accepted double exponential model presented some interesting similarities and differences. The mass flow rates predicted by the two models were in close agreement, but the double exponential model was not able to predict the release of fluid as liquid. The RFB model predicted that 25 per cent of the mass released would be liquid.

  13. Analysis of transient coolant void formation during a guillotine-type HX tube rupture event in the Star-LM system employing a supercritical CO2 Brayton cycle

    International Nuclear Information System (INIS)

    One proposed concept for the STAR-LM Lead Fast Reactor (LFR) incorporates a supercritical CO2 gas turbine Brayton cycle to achieve high cycle efficiency and reduced plant footprint. In this design, 100+% of core full power is transferred by natural circulation from the core, located at the bottom of the reactor vessel, to in-vessel heat exchangers (HXs) located at the top of the vessel in the annulus between the core shroud and vessel inner wall. Although this approach extremely simplifies the plant design, the presence of the HXs within the vessel raises concerns regarding the potential rupture of a HX tube that would initiate a high-pressure blowdown of CO2 into the lead coolant. The principal issue is to what extent, if any, is void entrained downwards with the coolant and then upwards through the core where adverse reactivity effects or degraded heat removal could result. To address this question, a scoping analysis of transient void formation during a guillotine-type HX tube rupture event in the STAR-LM employing a supercritical CO2 Brayton cycle has been performed. The void formation process is evaluated by solving a coupled set of ordinary differential equations describing: i) the supercritical CO2 blowdown, ii) bubble center-of-mass trajectory, iii) bubble growth rate, iv) bubble gas internal energy, and v) discrete bubble formation rate due to Taylor instability at the bubble/coolant interface. The results indicate that for thermal hydraulic conditions consistent with the current STAR-LM design, the peak blowdown rate from a single tube rupture is ∼ 2.5 kg/sec. The void formation process is dominated by large coherent gas bubbles that penetrate minimally downwards into the coolant due to the large coolant density. Rather, the gas pockets are predicted to periodically rise due to buoyancy and vent to the core cover gas region, as opposed to being swept downwards with the coolant. Moreover, the total CO2 fraction that is rendered in the form of discrete

  14. Two-phase flow dynamics in ECC

    International Nuclear Information System (INIS)

    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.)

  15. Structural modelling and testing of failed high energy pipe runs: 2D and 3D pipe whip

    OpenAIRE

    Reid, SR; Wang, B.; Aleyaasin, M

    2011-01-01

    Copyright @ 2011 Elsevier The sudden rupture of a high energy piping system is a safety-related issue and has been the subject of extensive study and discussed in several industrial reports (e.g. [2], [3] and [4]). The dynamic plastic response of the deforming pipe segment under the blow-down force of the escaping liquid is termed pipe whip. Because of the potential damage that such an event could cause, various geometric and kinematic features of this phenomenon have been modelled from th...

  16. The RELAP-UK MK 4 transient thermal-hydraulic code summary and input data description

    International Nuclear Information System (INIS)

    RELAP-UK MK IV is the latest UK code in the RELAP series of transient thermal-hydraulic codes for water reactor safety and fault analysis. It is the first version with full capability for PWR blowdown. The major improvements over earlier versions are a drift flux model, the Bryce flow-dependent slip correlation, a revised bubble rise model and a generalised 'heat slab' option. Other developments include a simple rewetting model, Fanning friction factor and change of area pressure drop models, time-dependent boundary nodes and an option to input pump speed history. RELAP-UK MK IV is available on the Harwell IBM 370-168 computer. (author)

  17. Steam line rupture experiments with the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

    The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

  18. Periodical shedding of cloud cavitation from a single hydrofoil in high-speed cryogenic channel flow

    OpenAIRE

    伊藤,優; Ito, Yutaka; 長崎, 孝夫; NAGASAKI, TAKAO

    2009-01-01

    In order to explain criteria for periodical shedding of the cloud cavitation, flow patterns of cavitation around a plano-convex hydrofoil were observed using a cryogenic cavitation tunnel of a blowdown type. Two hydrofoils of similarity of 20 and 60 mm in chord length with two test sections of 20 and 60 mm in width were prepared. Working fluids were water at ambient temperature, hot water and liquid nitrogen. The parameter range was varied between 0.3 and 1.4 for cavitation number, 9 and 17 m...

  19. Analysis of heat transfer from fuel rods with externally attached thermocouples

    International Nuclear Information System (INIS)

    This paper describes the development of 2 and 3 dimensional finite element heat conduction models to simulate the behaviour of the external thermocouples attached to the LOFT fuel rods during the blowdown phase of a large break loss-of-coolant accident. To establish the model and determine the thermal coupling between the thermocouple and the fuel rod extensive use was made of two series of experiments performed at INEL in the LOFT Test Support Facility (LTSF). These experiments were high pressure reflood experiments with fluid conditions 'typical' of those seen during the bottom-up flow period of the LOFT experiments. (author)

  20. Hypersonic Wind Tunnels: Latest Citations from the Aerospace Database

    Science.gov (United States)

    1996-01-01

    The bibliography contains citations concerning the design, construction, operation, performance, and use of hypersonic wind tunnels. References cover the design of flow nozzles, diffusers, test sections, and ejectors for tunnels driven by compressed air, high-pressure gases, or cryogenic liquids. Methods for flow calibration, boundary layer control, local and freestream turbulence reduction, and force measurement are discussed. Intrusive and non-intrusive instrumentation, sources of measurement error, and measurement corrections are also covered. The citations also include the testing of inlets, nozzles, airfoils, and other components of hypersonic aerospace vehicles. Comprehensive coverage of supersonic and blowdown wind tunnels, and force balance systems for wind tunnels are covered in separate bibliographies.

  1. Loading operations for spacecraft propulsion subsystems

    Science.gov (United States)

    Purohit, G. P.; Nordeng, H. O.; Ellison, J. R.

    1992-07-01

    This paper provides a broad overview of loading operations for pressurized blowdown monopropellant and pressure regulated integral bipropellant propulsion subsystems used in geosynchronous communication satellites. Propellant chemical composition, cleanliness, processing, and handling requirements are addressed. Ground servicing equipment (GSE) and propellant transfer procedures for the various loading configurations are discussed. Effects of helium solubility and helium saturation levels in both GSE carts and propellant tanks are examined. Predicted equilibrium pressures for actual postload tank pressures are compared against extensive loading data on Hughes bipropellant spacecraft. Helium tank pressurization and manifold pressurization practices are described. Propellant loading facility requirements and safety requirements are discussed.

  2. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point

  3. RELAP5-3D code for supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point. (author)

  4. LOCA verification and data bank

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. Design of particle bed reactors for the space nuclear thermal propulsion program

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H.; Powell, J.R.; Todosow, M.; Maise, G.; Barletta, R.; Schweitzer, D.G. [Brookhaven National Lab., Upton, NY (United States)

    1996-02-01

    This paper describes the design for the Particle Bed Reactor (PBR) that was considered for the Space Nuclear Thermal Propulsion (SNTP) Program. The methods of analysis and their validation are outlined first. Monte Carlo methods were used for the physics analysis, several new algorithms were developed for the fluid dynamics, heat transfer and transient analysis; and commercial codes were used for the stress analysis. We carried out a critical experiment, prototypic of the PBR to validate the reactor physics; blowdown experiments with beds of prototypic dimensions were undertaken to validate the power-extraction capabilities from particle beds. In addition, materials and mechanical design concepts for the fuel elements were experimentally validated. (author).

  7. Effects of non-uniform core flow on peak cladding temperature: MOXY/SCORE sensitivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S.C.

    1979-08-15

    The MOXY/SCORE computer program is used to evaluate the potential effect on peak cladding temperature of selective cooling that may result from a nonuniform mass flux at the core boundaries during the blowdown phase of the LOFT L2-4 test. The results of this study indicate that the effect of the flow nonuniformity at the core boundaries will be neutralized by a strong radial flow redistribution in the neighborhood of core boundaries. The implication is that the flow nonuniformity at the core boundaries has no significant effect on the thermal-hydraulic behavior and cladding temperature at the hot plane.

  8. Liquid neon heat transfer as applied to a 30 tesla cryomagnet

    Science.gov (United States)

    Papell, S. S.; Hendricks, R. C.

    1975-01-01

    A 30-tesla magnet design is studied which calls for forced convection liquid neon heat transfer in small coolant channels. The design also requires suppressing boiling by subjecting the fluid to high pressures through use of magnet coils enclosed in a pressure vessel which is maintained at the critical pressure of liquid neon. This high pressure reduces the possibility of the system flow instabilities which may occur at low pressures. The forced convection heat transfer data presented were obtained by using a blowdown technique to force the fluid to flow vertically through a resistance heated, instrumented tube.

  9. Function analysis of steam isolation valves

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  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)

    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. 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.

  13. Verification study of LOCA analysis code THYDE-P

    International Nuclear Information System (INIS)

    THYDE-P is a code to analyze loss-of-coolant accidents (LOCA) of the pressurized water reactor (PWR). In this report, the blowdown portion of THYDE-P sample calculation Run 10 is presented along with THYDE-P inputs requirements. Run 10 forms a portion of a series of THYDE-P sample calculations to be performed by the evaluation model option on a specified plant design and is characterized by a simple nodalization such as a single active core node and discharge coefficient 0.6. (author)

  14. Assessment of TRACE Code for GE Level Swell Test to Review Industrial Code

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chanyi; Cheng, Ae Ju; Bang, Young Seok; Hwang, Taesuk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    Korea Institute of Nuclear Safety(KINS) has reviewed the industrial code for safety analysis of nuclear power plant, in which TRACE and MARS-KS codes are being used to support the understanding of specific phenomena and code prediction. For this aspect, the TRACE code was assessed for the GE Level Swell Experiment. General Electric (GE) performed a series of experiments to investigate thermal-hydraulic phenomena such as critical flow, void distribution, and liquid-vapor mixture swell during blowdown conditions. These GE Level swell experiments are frequently simulated to verify safety analysis codes as a separate effect test. TRACE code calculations with version 5.0 patch 4 for GE Level Swell experiment 1004-3 have been performed to assess the applicability of the TRACE code for verification of industrial code. An Assessment analysis of the TRACE version 5.0 patch 4 code was carried out for GE Level Swell experiments 1004-3 by comparison purpose with SPACE. Overall, TRACE predicted the pressure and axial void fractions at different times reasonably well for 1004-3 blowdown test, while SPACE tends to underestimate the pressure. It was also found that results of void fraction distribution should be compared at different time to discuss the accuracy of the SPACE code against this test.

  15. Preliminary Sensitivity Study of Upper Head Nodalization for LBLOCA in APR-1400

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu; Yoo, Seung Hun; Cho, Dae-Hyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the key-way bypass was determined to be - 0.3 %. The steady state condition which is the initial condition for LBLOCA was obtained by MARS-KS calculation. Up to now, it was assumed that the temperature of the upper dome in APR-1400 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. In this study, the preliminary sensitivity study of original and modified nodalization for LBLOCA was performed, and the effect of upper head nodalization and temperature was evaluated qualitatively. In this study, the preliminary sensitivity study of original and modified nodalization for upper head in APR-1400 was performed, and the effect of upper head nodalization and temperature on LBLOCA PCT was evaluated qualitatively. Through the transient calculation, it was confirmed that the upper head temperature affects the water inventory in the upper head at the early stage of LBLOCA so it does the blowdown quenching and following reflood PCT significantly. The results in this study were caused by very conservative upper head temperature determination.

  16. 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.

  17. The Effect of Temperature of Opening and Closed Cooling Water on Selecting Plate Heat Exchangers%汽轮机低压缸排污系统设计

    Institute of Scientific and Technical Information of China (English)

    郭锋; 刘杨

    2012-01-01

    汽轮机在低压缸靠近轴承两端设置有排污口,此接口主要用于排放低压缸凝结水,由于靠近汽轮机轴承,因此这部分凝结水含有油,但由于排放量小,绝大部分工程将此路排污水直接排至凝汽器,含油废水排放至凝汽器会污染凝结水,结合低压缸排污参数特点设计一种水封排污罐,这种设计简单,便于布置,以供同行参考借鉴.%There are blowdown interfaces at bearing side of LP. They are using for condensate of the LP. Because of they are near the bearing,so oil maybe contain in the condensate. Because the quantity of condensate is little,so it is discharged to the condenser nearly all projects. It will affect condensate quality. In this paper,there is a new kind of water-sealing blowdown tank designed,it is easily designed and fixed.

  18. Dispersant trial at ANO-2: Results from a short-term trial prior to SG replacement

    Energy Technology Data Exchange (ETDEWEB)

    Fruzzetti, K.; Frattini, P. [Electric Power Research Inst., Palo Alto, CA (United States); Robbins, P. [Entergy Operations, Arkansas Nuclear One, Russellville, AR (United States); Miller, A. [Pedro Point Technology, Inc., Pacifica, CA (United States); Varrin, R.; Kreider, M. [Dominion Engineering Inc., McLean, VA (United States)

    2002-07-01

    Corrosion products in the secondary side of pressurized water reactor (PWR) steam generators (SGs) primarily deposit on the SG tubes. These deposits can inhibit heat transfer, lead to thermal-hydraulic instabilities through blockage of tube supports, and create occluded regions where corrosive species can concentrate along tubes and in tube-to-tube support plate crevices. The performance of the SGs is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. A potential strategy for minimizing deposition of corrosion products on SG internal surfaces is to use an online dispersant to help prevent the corrosion products from adhering to the steam generator surfaces. By inhibiting the deposition of the corrosion products, the dispersant can facilitate more effective removal from the SGs via blowdown. This type of strategy has been employed at fossil boilers for many decades. However, due to the use of inorganic (sulfur and other impurities) polymerization initiators, polymeric dispersants had not been utilized in the nuclear industry. Only recently has a poly-acrylic acid dispersant, developed by BetzDearborn (PAA), been available that meets the criteria for nuclear application. This paper summarizes the results of the short-term PAA dispersant trial in Winter/Spring 2000, lasting approximately 3 months, performed at Arkansas nuclear one unit 2 (ANO-2)-including the chronology of the trial, the increase in blowdown iron removal efficiency with use of the dispersant, and observed effects on SG performance. (authors)

  19. An experimental investigation of critical flow rates of subcooled water through a thick orifice with a small diameter

    International Nuclear Information System (INIS)

    To study critical flow phenomenon in a thick orifice, and to generate technical data to evaluate the performance of break simulator design for small break accidents in a nuclear power plant, critical flow tests have been performed at the Blowdown and Condensation Loop. Steady state and blowdown critical flow tests have been performed using eight different-shaped thick orifices. The steady state flow data show that the critical mass flux can be expressed as a function of the discharge coefficient and initial conditions. Based upon the test results, a semi-empirical model has been developed. Comparison between the model prediction and test data from various sources showed that the critical mass flux through a thick orifice (0.4 ≤ L/d ≤ 2.0) with a small diameter (2.0 ≤ d ≤ 12.7 mm) can be accurately predicted. The characteristics of two typical break simulators, which simulate break geometries during a small break loss-of-coolant-accident, were analyzed and provisions for the design of a break simulator for a small scale test facility have been suggested. (author)

  20. Reactive Additive Stabilization Process (RASP) for hazardous and mixed waste vitrification

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

    Solidification of hazardous/mixed wastes into glass is being examined at the Savannah River Site (SRS) for (1) nickel plating line (F006) sludges and (2) incinerator wastes. Vitrification of these wastes using high surface area additives, the Reactive Additive Stabilization Process (RASP), has been determined to greatly enhance the dissolution and retention of hazardous, mixed, and heavy metal species in glass. RASP lowers melt temperatures (typically 1050-- 1150{degrees}C), thereby minimizing volatility concerns during vitrification. RASP maximizes waste loading (typically 50--75 wt% on a dry oxide basis) by taking advantage of the glass forming potential of the waste. RASP vitrification thereby minimizes waste disposal volume (typically 86--97 vol. %), and maximizes cost savings. Solidification of the F006 plating line sludges containing depleted uranium has been achieved in both soda-lime-silica (SLS) and borosilicate glasses at 1150{degrees}C up to waste loadings of 75 wt%. Solidification of incinerator blowdown and mixtures of incinerator blowdown and bottom kiln ash have been achieved in SLS glass at 1150{degrees}C up to waste loadings of 50% using RASP. These waste loadings correspond to volume reductions of 86 and 94 volume %, respectively, with large associated savings in storage costs.

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

    International Nuclear Information System (INIS)

    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)

  2. Reactive Additive Stabilization Process (RASP) for hazardous and mixed waste vitrification

    International Nuclear Information System (INIS)

    Solidification of hazardous/mixed wastes into glass is being examined at the Savannah River Site (SRS) for (1) nickel plating line (F006) sludges and (2) incinerator wastes. Vitrification of these wastes using high surface area additives, the Reactive Additive Stabilization Process (RASP), has been determined to greatly enhance the dissolution and retention of hazardous, mixed, and heavy metal species in glass. RASP lowers melt temperatures (typically 1050-- 1150 degrees C), thereby minimizing volatility concerns during vitrification. RASP maximizes waste loading (typically 50--75 wt% on a dry oxide basis) by taking advantage of the glass forming potential of the waste. RASP vitrification thereby minimizes waste disposal volume (typically 86--97 vol. %), and maximizes cost savings. Solidification of the F006 plating line sludges containing depleted uranium has been achieved in both soda-lime-silica (SLS) and borosilicate glasses at 1150 degrees C up to waste loadings of 75 wt%. Solidification of incinerator blowdown and mixtures of incinerator blowdown and bottom kiln ash have been achieved in SLS glass at 1150 degrees C up to waste loadings of 50% using RASP. These waste loadings correspond to volume reductions of 86 and 94 volume %, respectively, with large associated savings in storage costs

  3. Morpholine decomposition products in the secondary cycle of CANDU-PHWR plants

    International Nuclear Information System (INIS)

    Trace amounts of organic compounds resulting from the decomposition of morpholine additive used for erosin-corrosion control were determined in CANDU-PHWR steam-condensate cycles. Most of the morpholine breakdown products (2-(2-aminoethoxy) ethanol, ethanolamine, ammonia, methylamine, ethylamine, ethylene glycol, glycolic and acetic acids) identified during thermal-decomposition tests in the laboratory were detected in the steam-condensate cycles investigated, thus confirming the proposed morpholine reaction scheme. Their relative concentration in cycle components is affected by the use of condensate polishing, the presence of contaminants in the feeding morpholine solutions, the presence of non-ionic or weakly-ionized organic matter in the makeup water, and the organic contaminants introduced into the cycle by condenser leaks. Comparison of the analytical results before and after feeding the morpholine into the cycle of one of the plants investigated confirms that the thermal decomposition of this additive contributes significantly to the formation of glycolic and acetic acids, reported to be responsible for a cation conductivity increase of about 0.009 and 0.0675 mS/m in steam-generator blowdowns and moisture separator/reheater drains, respectively. Finally, an important fraction of these breakdown products is removed by the blowdown of the steam generator, the deaerator and the condensate polisher. (orig.)

  4. Experiment data report for semiscale Mod-1 test S-04-1 (baseline ECC test)

    International Nuclear Information System (INIS)

    Recorded test data are presented for Test S-04-1 of the Semiscale Mod-1 Baseline ECC Test Series. This test is 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 system. Test S-04-1 was conducted from an initial cold leg fluid temperature of 5420F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using system volume scaled coolant injection parameters. System flow was set to achieve a core fluid temperature differential of 660F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. Optimization of the operation of liquid radioactive waste treatment plants

    International Nuclear Information System (INIS)

    An analysis was made of the possibilities of optimizing the operation of liquid radioactive waste treatment plants of the V-1 nuclear power plant, this with the aim of reducing the amount or influencing the composition of these wastes. Two treatment plants were in the centre of attention, contributing most to the production of radioactive concentrate. The first is designed for unorganized releases from the primary circuit, for water from decontamination, special laundries, etc., the second for surface blowdowns from the steam generators. The best operating mode of treatment plants for minimizing the amount of liquid radioactive wastes will be achieved by selecting the most favourable operating temperature and flow rate of the treated medium. The first mentioned treatment plant treats waste waters by evaporization and by subsequent processing of the condensate on ion exchange filters; here substantial improvement was achieved mainly by incorporating forced circulation of the liquid phase between the evaporators. In optimizing the operating regime of the treatment plant for surface blowdowns, attention was mainly devoted to loosening, washing, regeneration and flushing and to the possibility of separately processing used solutions. The studies and experiments yielded draft operating regulations for treatment plants. (Z.M.)

  9. PBF-LOCA test series test LOC-11 test result report

    International Nuclear Information System (INIS)

    This report presents the results of Loss-of-Coolant (LOC) Test LOC-11, the first test of the Loss-of-Coolant Accident (LOCA) Test Series conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The primary objective of the test was to evaluate the behavior of pressurized water reactor (PWR) fuel under LOCA conditions similar to those postulated during a simulated double-ended cold leg break in a PWR. Test LOC-11 consisted of four, separately shrouded, fresh fuel rods of PWR design, with initial plenum pressure as a variable. Maximum cladding temperatures of up to 10700K (corresponding to high ductility α-phase Zircaloy) were sought during Test LOC-11. The fuel rods were exposed to a series of three blowdowns from different power and coolant conditions. The final blowdown resulted in the maximum measured cladding temperature of 10340K. Upon disassembly of the test train the rods were found to be uniformly covered with a dark grey oxide. Posttest results indicated slight cladding circumferential swelling of the pressurized rods and slight collapse of the relatively unpressurized rods. The results are compared with the posttest analyses to aid in understanding the coolant thermal-hydraulic behavior and fuel rod behavior

  10. Calculation of Semiscale test S-06-3 using RELAP4/MOD6 and RELAP-UK Mk IV

    International Nuclear Information System (INIS)

    Calculations have been carried out using the thermal hydraulics codes RELAP4/MOD6 and RELAP-UK Mk IV to simulate a test carried out in the Semiscale facility at INEL. This experiment, simulates a large cold leg break LOCA in a PWR and consists of a blowdown, a refill and a reflood phase. Predictions of the fluid behaviour by MOD6 and RELAP-UK during blowdown were reasonably good although because of the sensitivity of clad temperatures to the core flow in a 'stagnation' situation, the resulting temperatures exhibited significant differences. MOD6 produced a good prediction of the quench behaviour of the average fuel pins when the experimental clad temperatures at the start of reflood were used to initialise the calculation. The quench behaviour was found to be quite insensitive to fluid conditions in the intact loop. These were not well predicted by the code but the discrepancy may be due in part to the neglect of metalwork heat in the calculation. (author)

  11. Waste water treatment options for SAGD oil production facilities

    Energy Technology Data Exchange (ETDEWEB)

    Portelance, S.N. [WorleyParsons MEG Ltd., Calgary, AB (Canada)

    2008-07-01

    Steam assisted gravity drainage (SAGD) water treatment facilities produce concentrated waste streams that contain high concentrations of total dissolved solids. The waste streams are typically partially recycled to upstream processes or injected into wells. However, these methods can result in the precipitation of silicate compounds and chemical imbalances in upstream water treatment processes. This study simulated 2 SAGD processes and MVC and once-through steam generator (OTSG) waste water treatment options. MVC waste water treatments were simulated with sulfuric acid only; with sulfuric acid and magnesium oxide; and low TH-high silica OTSG blowdown. Results of the simulations showed that the waste water generated was adequately treated with a combination of acid and magox. Further reductions in pH reduced silica contents and alkalinity. Costs for the treatment were estimated at $6.17 per metre{sup 3} for MVC waste water and $1.77 m{sup 3} for blowdown waste water. The addition of magox lowered the cost for silica removal to $4.60 per m{sup 3}. It was concluded that waste water treatment is needed to make produced water treatment options viable with the oil sands industry. 2 refs., 3 tabs., 10 figs.

  12. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, January--March 1976

    Energy Technology Data Exchange (ETDEWEB)

    Zane, J. O.; Farman, R. F.; Hanson, D. J.; Peterson, A. C.; Ybarrondo, L. J.; Berta, V. T.; Naff, S. A.; Crocker, J. G.; Martinson, Z. R.; Smolik, G. R.; Cawood, G. W.; Quapp, W. J.; Ramsthaler, J. H.; Ransom, V. H.; Scofield, M. P.; Dearien, J. A.; Bohn, M. P.; Burnham, B. W.; James, S. W.; Lee, W. H.; Lime, J. F.; Nalezny, C. L.; MacDonald, P. E.; Thompson, L. B.; Domenico, W. F.; Rice, R. E.; Hendrix, C. E.; Davis, C. B.

    1976-06-01

    Light water reactor sfaety research performed January through March 1976 is summarized. Results of the Semiscale Mod-1 blowdown heat transfer test series relating to those phenomena that influence core fluid and heat transfer effects are analyzed, and preliminary analyses of the recently completed reflood heat transfer test series are summarized for the forced and gravity feed reflood tests. The first nonnuclear LOCE in the LOFT program was successfully completed and preliminary results are presented. Preliminary results are given for the PCM 8-1 RF Test, the PCM-2A Test, and the Irradiation Effects Scoping Test 2 in the Thermal Fuel Behavior Program. Model development and verification efforts reported in the Reactor Behavior Program include checkout of RELAP4/MOD5 Update 1, development of a new hydrodynamic model for two-phase separated flows, development of the RACHET code to assess the assumptions in current fuel behavior codes of uniform stress and strain in the cladding, modifications of the containment code BEACON, analysis of results from the Halden Assembly IFA-429 helium sorption experiment, development of correlations for the thermal conductivity of UO/sub 2/ and (U,Pu)O/sub 2/, and evaluation of RALAP4 through comparison of calculated results with data from the GE Blowdown Heat Transfer and Semiscale experiments.

  13. 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

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

    International Nuclear Information System (INIS)

    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

  15. 某电厂膜法用于循环水系统的方案优化%Optimization of Membrane Method Used for Circulating Water System in Thermal Power Plant

    Institute of Scientific and Technical Information of China (English)

    姜琪; 李瑞瑞; 雷方俣; 苏艳

    2014-01-01

    以某水源为中水的废水“零排放”型火电厂,设置了循环水排污水回用系统,设计处理工艺为混凝澄清预处理-超滤-反渗透,反渗透淡水作为循环水及锅炉补给水系统水源。由于采用膜技术回收利用循环水排污水的工程实例中均存在膜系统运行不稳定,甚至无法正常运行的情况,并影响锅炉供水安全。对循环水处理工艺进行了试验研究。结果表明,超滤-反渗透装置水源采用石灰处理后的中水,循环水排污水采用碱性软化旁流处理工艺,可以减缓膜污堵,确保循环水系统稳定运行及锅炉给水安全,并满足全厂废水“零排放”的要求。%Reclaimed water is used as water source in a wastewater zero discharge thermal power plant .The original de-signed treatment process is coagulation and clarification pretreatment- ultrafiltration - reverse osmosis for blowdown water and the reverse osmosis freshwater is used as make-up water source of circulating water and the boiler .In existed projects of blowdown water recycling system using membrane technology ,membrane systems run not stably or even can’t run normal-ly and affect the security of boiler water supply .The circulating water treatment process is experimentally studied in this arti-cle .The results indicate that the reclaimed water treated by lime cad be used as make-up water source of ultrafiltration and reverse system and blowdown water can be treated with alkali bypass flow treatment ,which can slow membrane fouling ,en-sure stable operation of the circulating water system and security of boiler feed water and meet the demand of wastewater zero discharge .

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

    International Nuclear Information System (INIS)

    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

  17. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T.

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

    International Nuclear Information System (INIS)

    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

  19. Experimental investigation of air bubble flows in a water pool

    International Nuclear Information System (INIS)

    This paper presents experimental results on rising bubbles in the wetwell of a boiling water reactor (BWR) in a loss-of-coolant accident in the pressure suppression pool (PSP). This accident scenario includes three processes: blowdown and associated water slug phenomena, bubble dynamics and related water flow during continuous release of gases and development of a thermal stratification. The paper covers the middle phase where air is fed through a downcomer. The developments of bubble formation and bubble flow are investigated by means of high speed videos. Diameter, velocity, formation frequency and breakup distance of bubbles are evaluated using automated image evaluation procedures. The experiments have been performed in the cylindrical vessel of the THAI test facility with a height of 9.2 m and a diameter of 3.2 m. (author)

  20. Ceramic and coating applications in the hostile environment of a high temperature hypersonic wind tunnel. [Langley 8-foot high temperature structures tunnel

    Science.gov (United States)

    Puster, R. L.; Karns, J. R.; Vasquez, P.; Kelliher, W. C.

    1981-01-01

    A Mach 7, blowdown wind tunnel was used to investigate aerothermal structural phenomena on large to full scale high speed vehicle components. The high energy test medium, which provided a true temperature simulation of hypersonic flow at 24 to 40 km altitude, was generated by the combustion of methane with air at high pressures. Since the wind tunnel, as well as the models, must be protected from thermally induced damage, ceramics and coatings were used extensively. Coatings were used both to protect various wind tunnel components and to improve the quality of the test stream. Planned modifications for the wind tunnel included more extensive use of ceramics in order to minimize the number of active cooling systems and thus minimize the inherent operational unreliability and cost that accompanies such systems. Use of nonintrusive data acquisition techniques, such as infrared radiometry, allowed more widespread use of ceramics for models to be tested in high energy wind tunnels.

  1. 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.

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

    International Nuclear Information System (INIS)

    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 used for constellation maintenance of Cubesats, which measure 10 × 10 × 10 cm and have a mass less than 1 kg. The feed and thruster system contains an active valve, control electronics, a particle filter and an axisymmetric converging–diverging nozzle, all fabricated as separate modules. A novel method is used to integrate these modules by placing them on or in a glass tube package. The assembly method is shown to be a valid method but the valve module needs to be improved considerably

  3. RELAP4/MOD6/U4/J3: a JAERI improved version of RELAP4/MOD6 for transient thermal-hydraulic analysis of LWR including effects of BWR core spray

    International Nuclear Information System (INIS)

    The RELAP4/MOD6/U4/J3 code is the latest version of RELAP4/MOD6/Update4 improved in JAERI. The major improvements and modifications included in this version have been carried out aiming at small break LOCA analysis and BWR-LOCA analysis after core spray initiation. For example, a CCFL calculation model and a spray heat transfer model have been added for BWR-LOCA analysis. Using these models, through calculation from the beginning of blowdown to the end of reflood in BWR-LOCA was made practicable. Furthermore, the analyses of operational transients of LWR were facilitated greatly by an addition of a trip reset function. In this report, the description of the improvements and modifications included in this version, the input data description, and the results of two sample problems are contained. (author)

  4. Simulation of pressure waves in the coolant loop of PWR type reactors with a network of one-dimensional flow channels, taking the structural flexibility into account

    International Nuclear Information System (INIS)

    The DAPSY code is explained to be a universal tool for simulating and describing dynamic load effects on pipings, internals and components, and valves in the coolant loop. Excitation of pressure waves primarily is due to pipe rupture which leads to rapid pressure reduction. This is why the code very carefully calculates critical blowdown rates also for the case of only partial rupture with reduced outflow, as thus the course of disturbance is described that affects the system. A network method is presented for calculation of multidimensional geometries. As the pressure wave phenomena are observed in a low-compressibility fluid and in a system with sometimes very flexible structural components, the fluid-structure interactions are taken into account. The model presented allows to consider either quasi-static structural behaviour, or dynamic interaction of fluid and structure, depending on the configuration characteristics. (orig./HP)

  5. ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3.07.9 - steady-state film boiling in upflow

    International Nuclear Information System (INIS)

    Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

  6. Plant application uncertainty evaluation of LBLOCA analysis using RELAP5/MOD3/KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Chung, Bub Dong; Hwang, Tae Suk; Lee, Guy Hyung; Chang, Byung Hoon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the revised ECCS rule has been developed and this report describes the application of new REM to large break LOCA. A computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/ MOD3.1 was used as the best estimated code for the analysis and Kori unit 3 and 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by using Monte-Carlo sampler for each response surface. This study shows that plant application uncertainty can be quantified and demonstrates the applicability of the new realistic evaluation methodology. (Author) 29 refs., 40 figs., 8 tabs.

  7. Intact loop pump performance during the Semiscale Mod-1 isothermal test series

    International Nuclear Information System (INIS)

    An analysis was performed on the Semiscale Mod-1 intact loop pump data taken during the Semiscale Mod-1 isothermal test series. The pump was shown to directly affect intact loop and vessel flow rates during the early portion of the simulated loss-of-coolant accidents (LOCAs). Comparison of pump performance data taken during the Semiscale Mod-1 isothermal tests with data obtained during previous steady state and transient tests indicated that the pump head degraded more rapidly during the Semiscale Mod-1 tests. Calculations using the pump model contained in the RELAP4 computer program are compared with these pump performance data. Areas of operation for the Semiscale Mod-1 pump are defined for the transient two-phase steam-water flows that occurred during several isothermal blowdown tests, and suggested refinement in the two-phase characteristics of the Semiscale Mod-1 pump model is offered. 13 references

  8. 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.

  9. Comparison of BEACON and COMPARE reactor cavity subcompartment analyses

    International Nuclear Information System (INIS)

    In this study, a more advanced best-estimate containment code, BEACON-MOD3A, was ued to calculate force and moment loads resulting from a high-energy blowdown for two reactor cavity geometries previously analyzed with the licensing computer code COMPARE-MOD1A. The BEACON force and moment loads were compared with the COMPARE results to determine the safety margins provided by the COMPARE code. The forces and moments calculated by the codes were found to be different, although not in any consistent manner, for the two reactor cavity geometries studied. Therefore, generic summary statements regarding margins cannot be made because of the effects of the detailed physical configuration. However, differences in the BEACON and COMPARE calculated forces and moments can be attributed to differences in the modeling assumptions used in the codes and the analyses

  10. Depressurisation studies. Phase 2: results of Tests 127 and 128

    International Nuclear Information System (INIS)

    A basic experimental programme involving the sudden depressurisation of a simple pipe system containing water at 3.45 to 17.24MPa pressure and temperature in the range of 200 to 2500C has been concluded. Measurements were made of the transient density, pressure, and temperature variations in a two phase fluid in the system during discharge. Phase 1 tests investigated blowdown from straight pipes 4m long with constant internal diameters of 73 and 32 mm. Phase 2 tests incorporated a reservoir added to the 32mm pipe. In this, the second of three reports on Phase 2 tests, the test assembly, instrumentation and experimental procedure are briefly described. The conditions and results are reported for two of the tests in which the liquid in the long discharge pipe was initially subcooled by 100C and 150C while the reservoir was at saturation conditions with a steam dome present. (UK)

  11. System design of the Pioneer Venus spacecraft. Volume 10: Propulsion/orbit insertion subsystem studies

    Science.gov (United States)

    Rosenstein, B. J.

    1973-01-01

    The Pioneer Venus orbiter and multiprobe missions require spacecraft maneuvers for successful accomplishment. This report presents the results of studies performed to define the propulsion subsystems required to perform those maneuvers. Primary goals were to define low mass subsystems capable of performing the required missions with a high degree of reliability for low cost. A review was performed of all applicable propellants and thruster types, as well as propellant management techniques. Based on this review, a liquid monopropellant hydrazine propulsion subsystem was selected for all multiprobe mission maneuvers, and for all orbiter mission maneuvers except orbit insertion. A pressure blowdown operating mode was selected using helium as the pressurizing gas. The forces associated with spacecraft rotations were used to control the liquid-gas interface and resulting propellant orientation within the tank.

  12. Thermal-hydraulic evaluation study of the effectiveness of emergency core cooling system for light water reactors

    International Nuclear Information System (INIS)

    In order to evaluate the core cooling capability of the emergeny core cooling system, which is a safety guard system of light water reactors for a loss-of-coolant accident, a variety of large scale test were performed. Through the results, many phenomena were investigated and the predictabity of analytical codes were examined. The tests conducted were a single-vessel blowdown test, emergency core cooling test in a PWR simulation facility, spray cooling test for a BWR, large scale reflood test and a separate effect test on countercurrent flow. These test results were examined to clarify thermal-hydraulic phenomena and the effect of various test parameters and were utilized to improve predictability of the analytical codes. Some models for flow behavior in the upper core were also developed. By evaluating the effectiveness of various emergency core cooling system configurations, more effective cooling system than the current one was proposed and demonstrated. (author)

  13. Experiment prediction for Loft Nonnuclear Experiment L1-4

    International Nuclear Information System (INIS)

    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 282.20C, 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

  14. Multispecies absorption spectroscopy of detonation events at 100  kHz using a fiber-coupled, time-division-multiplexed quantum-cascade-laser system.

    Science.gov (United States)

    Rein, Keith D; Roy, Sukesh; Sanders, Scott T; Caswell, Andrew W; Schauer, Frederick R; Gord, James R

    2016-08-10

    A mid-infrared fiber-coupled laser system constructed around three time-division-multiplexed quantum-cascade lasers capable of measuring the absorption spectra of CO, CO2, and N2O at 100 kHz over a wide range of operating pressures and temperatures is demonstrated. This system is first demonstrated in a laboratory burner and then used to measure temperature, pressure, and concentrations of CO, CO2, and N2O as a function of time in a detonated mixture of N2O and C3H8. Both fuel-rich and fuel-lean detonation cases are outlined. High-temperature fluctuations during the blowdown are observed. Concentrations of CO are shown to decrease with time for fuel-lean conditions and increase for fuel-rich conditions. PMID:27534467

  15. A Simplified Model for Detonation Based Pressure-Gain Combustors

    Science.gov (United States)

    Paxson, Daniel E.

    2010-01-01

    A time-dependent model is presented which simulates the essential physics of a detonative or otherwise constant volume, pressure-gain combustor for gas turbine applications. The model utilizes simple, global thermodynamic relations to determine an assumed instantaneous and uniform post-combustion state in one of many envisioned tubes comprising the device. A simple, second order, non-upwinding computational fluid dynamic algorithm is then used to compute the (continuous) flowfield properties during the blowdown and refill stages of the periodic cycle which each tube undergoes. The exhausted flow is averaged to provide mixed total pressure and enthalpy which may be used as a cycle performance metric for benefits analysis. The simplicity of the model allows for nearly instantaneous results when implemented on a personal computer. The results compare favorably with higher resolution numerical codes which are more difficult to configure, and more time consuming to operate.

  16. 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.

  17. Estimation of the hydrodynamic effects of a LOCA in A 4-loop PWR

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Potapov, S. [Electricite de France (EDF-SEP / AMV), 92 - Clamart (France); Tephany, F. [Electricite de France (EDF SEPTEN), 69 - Villeurbanne (France)

    2001-07-01

    The PWR safety studies involve an analysis of the consequences of a hypothetical rupture of a primary pipe. From the opening tune, the blowdown at the break causes the propagation of an acoustic wave through the whole primary circuit, as well as pipe whipping. The local pressure gaps due to the depressurization wave propagation may induce component recoils and internal structure movements. In parallel with the acoustic wave propagation, the circuit empties progressively first with a monophasic regime and later with a diphasic one. This paper presents a hydrodynamic simulation of the flows in the primary circuit of 4-loop PWR during a LOCA. The results concern the propagation of the depressurization acoustic wave along the circuit, coupled with the transient fluid flows. (authors)

  18. Liquid neon heat transfer as applied to a 30 tesla cryomagnet

    Science.gov (United States)

    Papell, S. S.; Hendricks, R. C.

    1975-01-01

    Since superconducting magnets cooled by liquid helium are limited to magnetic fields of about 18 teslas, the design of a 30 tesla cryomagnet necessitates forced convection liquid neon heat transfer in small coolant channels. As these channels are too small to handle the vapor flow if the coolant were to boil, the design philosophy calls for suppressing boiling by subjecting the fluid to high pressures. Forced convection heat transfer data are obtained by using a blowdown technique to force the fluid vertically through a resistance-heated instrumented tube. The data are obtained at inlet temperatures between 28 and 34 K and system pressures between 28 to 29 bars. Data correlation is limited to a very narrow range of test conditions, since the tests were designed to simulate the heat transfer characteristics in the coolant channels of the 30 tesla cryomagnet concerned. The results can therefore be applied directly to the design of the magnet system.-

  19. Effect of spacing between two adjoining circular cylinders on flow around two-dimensional circular cylinder rows. 2. Tow and three rows of transverse arrangement

    International Nuclear Information System (INIS)

    This paper describes an effect of spacing between two adjoining circular cylinders on flow around two-dimensional circular cylinder bundles. The experiment was carried out in an N.P.L blow-down type wind-tunnel with a working section of 500 mm x 500 mm x 2000 mm, and under the Reynolds number 1.3 x 104. The surface-pressure distributions on the circular cylinder were measured and the drag coefficient was determined from these measurements. The flow-pattern around circular cylinders was observed. The power spectrum in the turbulent wake behind circular cylinders was also measured. It was found that the pressure on the rear surface of circular cylinders becomes lower and the drag coefficient increases as the spacing ratio decreases, while the step-change in the drag coefficient occurs at the spacing ratio where the flow pattern around the downstream circular cylinder changes. (author)

  20. Water world

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Lynda

    2011-03-15

    Water is an important element in oilsands operations: three barrels of water are required to produce one barrel of oil; therefore, companies have to implement technologies to reduce their use of water and increase reuse. To do so, several technologies are available: the use of both heavy to light (HTL) and steam assisted gravity drainage (SAGD) eliminates the need for sour water disposal facilities; chemistry optimization enhances the phase separation of water and oil; and steam blowdown reduces the need for makeup water. Then to dispose of this water, companies can use disposal wells, landfills or salt caverns. Oilsand operators are implementing different processes in order to reduce their water use and their footprint at the same time.

  1. 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.

  2. Analytical modelling of hydrogen transport in reactor containments

    International Nuclear Information System (INIS)

    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

  3. Assessment of RELAP5/Mod3/KAERI using semiscale test S-06-3

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Choi, Han Rim; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report presents the results of the RELAP5/MOD3/KAERI assessment utilizing a semi scale large break loss-of-coolant experiment Test S-06-3. Test S-06-3 is a 200% double ended cold break experiment performed in semi scale Mod-1 facility in 1987 for the purpose of investigating the thermal and hydraulic phenomena accompanying a hypothetical large break LOCA in a pressurized were reactor system. Through comparisons between data and best-estimate RELAP5 calculation, the capabilities of RELAP5 to calculate the large break LOCA accident were assessed. Emphasis was placed on the capability of the code to calculate break flow rates during system blowdown phase, emergency core cooling system injection bypass during refill phase, quenching during reflood phase, and the peak cladding temperature behavior throughout the whole experiment. (Author) 12 refs., 38 figs., 2 tabs.

  4. Application of ADINA fluid element for transient response analysis of fluid-structure system

    International Nuclear Information System (INIS)

    Pressure propagation and Fluid-Structure Interaction (FSI) in 3D space were simulated by general purpose finite element program ADINA using the displacement-based fluid element which presumes inviscid and compressible fluid with no net flow. Numerical transient solution was compared with the measured data of an FSI experiment and was found to fairly agree with the measured. In the next step, post analysis was conducted for a blowdown experiment performed with a 1/7 scaled reactor pressure vessel and a flexible core barrel and the code performance was found to be satisfactory. It is concluded that the transient response of the core internal structure of a PWR during the initial stage of LOCA can be analyzed by the displacement-based finite fluid element and the structural element. (orig.)

  5. CFD analysis on the critical flow for nozzle design

    International Nuclear Information System (INIS)

    We discussed on the implementation of an upwind method for a new hyperbolic two-dimensional two-fluid model including the interfacial pressure jump term in the momentum equations. This model consists of a complete set of 8 equations including 2-mass, 4-momentum, and 2-internal energy conservation equations having all real eigenvalues. Usually the two-fluid models have been solved by donor-cell differencing method, which has been much employed in the commercial system codes. This method, however, introduced a large amount of numerical diffusion. In order to remove the excessive numerical diffusion which, the upwind scheme has been employed for this model. Based on this model system with HLL scheme among the approximate Riemann solvers, we first make a pilot 2-D code and then solve some benchmark problems: two-phase shock wave problem, water faucet problem, and phase separation problem. Lastly, we calculate the Edwards pipe blowdown problem among typical critical flow problems

  6. 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.

  7. Adsorption of sulfate in PWR steam generators: Laboratory tests

    International Nuclear Information System (INIS)

    Following observation of an apparent difference in the hideout mechanism for sulfate compared to that of other highly soluble species during chemical injection tests at several PWRS, a laboratory test program, discussed in this report was implemented to quantify sulfate adsorption on metal surfaces. Approximately 350 ug/m2 of sulfate could be adsorbed on Alloy 600 from neutral solutions at 300 degree C. Less adsorption was observed at lower temperature as well as at increased pH. The adsorbed sulfate could be desorbed into pure water over a period of several days subsequent to termination of sulfate ingress. Thus, a prompt shutdown to hot standby with maximization of blowdown should minimize the long term impact of sulfate steam generator corrosion subsequent to a period of significant sulfate or cation resin ingress. The only other species which exhibited significant adsorption was phosphate which also has a tetrahedral ionic structure in solution

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

    International Nuclear Information System (INIS)

    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

  9. Cooling tower water conditioning study. [using ozone

    Science.gov (United States)

    Humphrey, M. F.; French, K. R.

    1979-01-01

    Successful elimination of cooling tower treatment chemicals was demonstrated. Three towers functioned for long periods of time with ozone as the only treatment for the water. The water in the systems was reused as much as 30 times (cycles of concentration) without deleterious effects to the heat exchangers. Actual system blow-down was eliminated and the only makeup water added was that required to replace the evaporation and mist entrainment losses. Minimum water savings alone are approximately 75.1 1/kg/year. Cost estimates indicate that a savings of 55 percent was obtained on the systems using ozone. A major problem experienced in the use of ozone for cooling tower applications was the difficulty of accurate concentration measurements. The ability to control the operational characteristics relies on easily and accurately determined concentration levels. Present methods of detection are subject to inaccuracies because of interfering materials and the rapid destruction of the ozone.

  10. CHEMCON User's Manual, Version 3.1

    International Nuclear Information System (INIS)

    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

  11. The 7.4 per cent cold leg break without accumulator operation

    International Nuclear Information System (INIS)

    A simulation technique for the loss-of-coolant failure analysis of light-water-cooled nuclear reactor is described. It has been used to analyze transient processes during a hypothetical accident and to estimate the effectiveness of built-in safety systems. The model PMK-NHV was established for these types of simulation in the Paks Nuclear Power Plant, Hungary. The first test on this simulation facility is described: a 7.4 per cent cold leg break from full power covering the blowdown phase of the accident. The pre-test analysis using the RELAP4/mod6 computer code, the evaluation of the measured data, the interpretation of the test results and the post-test calculations are presented. The work was performed within the IAEA Standard Problem Exersice (SPE). (R.P.)

  12. The effect of encroachments on structure impact loads during a pool swell transient based on small-scale testing

    International Nuclear Information System (INIS)

    Experiments were conducted to investigate suppression pool dynamics in boiling water reactor (BWR) containments which have large overhanging structures attached to the drywell wall. Several 1/10 linear scale air blowdown tests utilizing Froude scaling (balance of gravity and inertia forces) were performed in this tests series. The drywall pressure was measured and high speed movies were made of the pool response. The resultant pool response was a function of encroachment size. Small encroachments did not significantly alter the response obtained for he unobstructed pool. For the large radial and circumferential encroachment, however, the increased inertia of the extra water lifted by the rising bubble delayed the transient, resulting in much lower pool swell velocities. This led to a stable liquid surface at higher elevations, but the surface curvature coupled with the relatively low pool surface velocities significantly mitigates structure impact loadings

  13. Comparison of code calculations with experiments on containment response during LOCA conditions

    International Nuclear Information System (INIS)

    A series of experiments were performed on a one-tenth scale model of PHWR containment, incorporating pressure suppression system. The pressure-temperature transients in the model containment observed during simulated LOCA (Loss of Coolant) blowdown conditions were compared against calculated results form computer code PACSR, for purposes of verification of the code. Comparison of results indicated that calculated values of peak pressure in various compartment were significantly higher than observed ones. This disagreement was attributed mainly to modelling for energy absorption from containment atmosphere to structural surfaces, this effect being particularly important in a scaled down model. Good agreement between calculation and experiment was obtained after heat transfer correlation for energy absorption on surfaces were modified in the code. The study demonstrates the conservatism of the results from the code. (author). 6 refs., 1 tab., 9 figs

  14. Development of General-Purpose Software to Analyze the Static Thermal Characteristic of Nuclear Power Plant

    Science.gov (United States)

    Nakao, Yoshinobu; Koda, Eiichi; Takahashi, Toru

    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.

  15. Permeability of Consolidated Incinerator Facility Wastes Stabilized with Portland Cement

    Energy Technology Data Exchange (ETDEWEB)

    Walker, B.W.

    1999-08-23

    The Consolidated Incinerator Facility (CIF) at the Savannah River Site (SRS) burns low-level radioactive wastes and mixed wastes as method of treatment and volume reduction. The CIF generates secondary waste, which consists of ash and off-gas scrubber solution. Currently the ash is stabilized/solidified in the Ashcrete process. The scrubber solution (blowdown) is sent to the SRS Effluent Treatment Facility (ETF) for treatment as waste water. In the past, the scrubber solution was also stabilized/solidified in the Ashcrete process as blowcrete and will continue to be treated this way for listed waste burns and scrubber solution that do not meet the Effluent Treatment Facility (ETF) Waste Acceptance Criteria (WAC).

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

    International Nuclear Information System (INIS)

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

  17. Electrochemical ion exchanger in the water circuit to measure cation conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Bengtsson, Bernt; Ingemarsson, Rolf; Settervik, Gustav [Ringhals AB, Vaeroebacka (Sweden); Velin, Anna [Vattenfall Research and Development AB, Stockholm (Sweden)

    2011-03-15

    At Ringhals Nuclear Power Plant (NPP), more than four years of successful operation with a full-scale electrode ionization (EDI) unit for the recycling of steam generator blowdown gave the inspiration to modify and scale down this EDI process. As part of this project, the possibility of replacing the cation exchanger columns used for cation conductivity analysis with some small and integrated electrochemical ion exchange cells was explored. Monitoring the cation conductivity requires the use of a small cation resin column upstream of the conductivity probe and is one of the most important analyses at power plants. However, when operating with high alkaline treatment in the steam circuit, there is the disadvantage of rapid exhaustion of the resins, necessitating frequent replacement or regeneration. This causes interruptions in the monitoring and gives rise to a high workload for the maintenance staff. This paper reports on the optimization and testing of two different two-compartment electrochemical cells for possible replacement of the cation resin columns for analyzing cation conductivity in the secondary steam circuit at Ringhals NPP. Field tests during start-up conditions and more than four months of steady operation together with real and simulated tests for impurity influences indicate that an electrical ion exchange (ELIX) process could be successfully used to replace the resin columns in Ringhals while operating with high-pH all-volatile treatment (AVT) using hydrazine and ammonia. Installation of an ELIX system downstream of a particle filter and upstream of a small cation resin column will introduce additional safety and further reduce the maintenance and possible interruptions. Performance of the ELIX process together with other chemical additives (morpholine, ethanolamine, 3-methoxypropylamine, dimethylamine) and dispersants may be further evaluated to qualify the ELIX process as well as steam generator blowdown electrodeionization for wider use in

  18. Transient simulation of feedwater vaporization during a DBA LOP/LOCA using RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Harrell, J.R. [ENERCON Services, Inc., Atlanta, GA (United States); Fuller, R.W. [Entergy Operations, Inc., Port Gibson, MS (United States)

    1996-07-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station (GGNS) 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. The original design and testing 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. Given this condition, the appropriate testing criteria would be based on air with a relatively tight allowable limit. 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 leakage flow exists from the reactor vessel to the condenser through the feedwater piping during the reactor vessel blowdown phase. 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.

  19. Experimental investigation of critical flow of supercritical carbon dioxide

    Science.gov (United States)

    Mignot, Guillaume Paul H.

    A blowdown facility (0.125 m3) has been built to perform measurements of the critical flow rate of carbon dioxide over a wide range of conditions up to a supercritical pressure of 240 bars and up to a supercritical temperature of 260°C, i.e. three times the critical pressure and two times the critical temperature. The influence of the rupture geometry was investigated using a set of exit pipes with varying entrance shape, roughness and length to diameter ratio ranging from 3.7 to 168. The study showed that a rough sharp edge entrance tube had a lower critical mass flow rate compared to a smooth round entrance tube. For length to diameter ratios larger than 14.7, although two-phase effects were observed, the fluid behavior could be accurately modeled using a homogeneous equilibrium model with friction. For length to diameter ratio smaller than 14.7, the critical mass flux results exhibited a plateau, indicating that the critical mass flow rate was governed by the vena contracta. Stagnation pressure, stagnation temperature and mass time traces were scaled successfully using the initial mass and the initial mass flow rate. An exception was observed for the high density low temperature case due to non equilibrium effects occurring within the vessel. The compressibility of the flow in association with the contraction induced multidimensional and repetitive shock structures within the tube. These have been predicted with computational fluid dynamics modeling for perfect gas conditions. To measure experimentally the fluid state within the tube, an optical absorption technique has been developed, calibrated and tested in two geometries and during an integral blowdown test. Results showed that this new technique lead to the correct qualitative trends in the pressure measurements but that it needed to be calibrated against a more accurate high pressure database obtained for carbon dioxide.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

    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 epsilon-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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. 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)

    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

  3. 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.

  4. Numerical modelling of low-Reynolds number direct contact condensation in a suppression pool test facility

    International Nuclear Information System (INIS)

    Highlights: • A low-Reynolds number direct contact condensation mode was simulated. • Eulerian two-fluid approach was used without interfacial tracking. • The numerical results were validated with the steam blowdown test. • The surface divergence model predicted the condensation phenomena satisfactory. - Abstract: In the safety pressure suppression pool systems of Boiling Water Reactors (BWRs), the condensation rate has to be maintained high enough in order to fulfill their safety function. A major part of this condensation occurs as direct contact condensation (DCC), which governs different modes varying from vigorous chugging of collapsing bubbles to mild condensation on almost flat steam–water interface. This paper discusses the Computational Fluid Dynamics (CFD) simulations of the latter, low-Reynolds number weak condensation regime. The numerical simulations were performed with two CFD codes, NEPTUNECFD and OpenFOAM, in which the DCC phenomenon was modelled by using the Eulerian two-fluid approach of interpenetrating continua without interfacial tracking. The interfacial heat transfer between steam and water was modelled by using the DCC models based on the surface renewal and the surface divergence theories. Flow turbulence was solved by employing the standard k–∊ turbulence model. The CFD results of this study were validated against the test results of the POOLEX facility of Lappeenranta University of Technology. In the reference test STB-31, the condensation phenomena were limited to only occur on a stable steam–water interface by very low steam mass flux applied and thermal insulation of the blowdown pipe. The simulation results demonstrated that the surface divergence model predicted the condensation phenomena quite accurately both qualitatively and quantitatively while the surface renewal model overestimated it strongly

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

    International Nuclear Information System (INIS)

    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

  6. Basin-Wide Amazon Forest Tree Mortality From a Large 2005 Storm

    Science.gov (United States)

    Negron Juarez, R. I.; Chambers, J. Q.; Guimaraes, G.; Zeng, H.; Raupp, C.; Marra, D. M.; Ribeiro, G.; Saatchi, S. S.; Higuchi, N.

    2010-12-01

    Blowdowns are a recurrent characteristic of Amazon forests and are produced, among others, by squall lines. Squall lines are aligned clusters (typical length of 1000 km, width of 200 km) of deep convective cells that produce heavy rainfall during the dry season and significant rainfall during the wet season. These squall lines (accompanied by intense downbursts from convective cells) have been associated with large blowdowns characterized by uprooted, snapped trees, and trees being dragged down by other falling trees. Most squall lines in Amazonia form along the northeastern coast of South America as sea breeze-induced instability lines and propagate inside the continent. They occur frequently (~4 times per month), and can reach the central and even extreme western parts of Amazonia. Squall lines can also be generated inside the Amazon and propagate toward the equator. In January 2005 a squall line propagated from south to north across the entire Amazon basin producing widespread forest tree mortality and contributed to the elevated mortality observed that year. Over the Manaus region (3.4 x104 km2), disturbed forest patches generated by the squall produced a mortality of 0.3-0.5 million trees, equivalent to 30% of the observed annual deforestation reported in 2005 over the same area. The elevated mortality observed in the Central Amazon in 2005 is unlikely to be related to the 2005 Amazon drought since drought did not affect Central or Eastern Amazonia. Assuming a similar rate of forest mortality across the basin, the squall line could have potentially produced tree mortality estimated at 542 ± 121 million trees, equivalent to 23% of the mean annual biomass accumulation estimated for these forests. Our results highlight the vulnerability of Amazon trees to wind-driven mortality associated with convective storms. This vulnerability is likely to increase in a warming climate with models projecting an increase in storm intensity.

  7. Sensitivity of the Amazon rainforest to convective storms

    Science.gov (United States)

    Negron Juarez, R. I.; Chambers, J. Q.; Rifai, S. W.; Urquiza Munoz, J. D.; Tello, R.; Alegria Munoz, W.; Marra, D.; Ribeiro, G.; Higuchi, N.

    2012-12-01

    The Amazon rainforest is the largest contiguous continental tropical forest in the world and is a world center of carbon storage, biodiversity, biogeochemical cycles and biogeophysical processes that affect the Earth climate system. Yet anthropogenic activities have produced changes in the forest-climate system. Consequently, an increase in rainfall in both the Western and Central Amazon and a decrease in the Eastern Amazon are expected due to these anthropogenic activities. While the projected decrease in rainfall has been discussed under the context of drought, deforestation, and fires, the effect of an increase in rainfall, and associated convective processes, on forest ecosystems has been overlooked. Across the Amazon rainforest, Western Amazonia has the highest precipitation rates, wood productivity, soil fertility, recruitment and mortality rates. Yet our field-measured tree mortality data from blowdowns that occurred in Western and Central Amazonia do not show a statistical difference in tree mortality between these regions. However, downburst velocities associated with these disturbances were calculated to be lower in Western Amazonia than in the Central Amazon. This suggests the Western Amazon is more highly sensitive to intense convective systems. This result is particularly relevant given the expected increase in rainfall in the Western and Central Amazon. The increase in rainfall is associated with more intense convective systems that in turn imply an increase in low level jet stream (LLJ) intensity east of the Andes. The presence of the LLJ is the main cause of squall lines and an increase in LLJ intensity will therefore cause increased propagation of squall lines into the Amazon basin. More frequent and active squall lines have the potential to increase the intensity and frequency of downbursts responsible for large forest blowdowns that will affect the biogeophysical feedbacks on the forest ecosystem and carbon budget.

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

    International Nuclear Information System (INIS)

    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)

  9. Development of an on-line process for steam generator chemical cleaning

    Energy Technology Data Exchange (ETDEWEB)

    Semmler, J.; Guzonas, D.A.; Rousseau, S.C.; Snaglewski, A.P.; Chenier, M.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    An on-line, preventative chemical cleaning process for the removal of secondary side oxides from steam generators is being developed. An on-line chemical cleaning process uses a low concentration of a chelant ({approx}1-10 mg L{sup -1}) to partially dissolve and dislodge the secondary side oxides while the steam generator is in operation. The dissolved and dislodged oxides can then be removed by blowdown. Feasibility tests were carried out in which the operating conditions of a CANDU steam generator were simulated in an autoclave containing either loose powdered magnetite or sintered magnetite on Alloy 800 (I-800) steam generator tube surfaces. The extent of magnetite dissolution in on-line solvent formulations containing either ethylenediaminetetraacetic acid (EDTA) or N-(2-hydroxyethyl)ethylenedinitrilo-N,N',N'-triacetic acid (HEDTA) at temperatures of 256 and 263 degrees C were measured. Powdered magnetite dissolved faster than sintered magnetite using both types of chelant. Dissolution continued as fresh chelant was added. The half-life (t{sup 1/2}) of Fe-EDTA complexes at 256 degrees C was approximately 3 h, sufficient to allow removal by blowdown. Hydrazine and morpholine were equally effective as oxygen scavengers. Increased dissolved oxygen concentration was found to result in chelant decomposition, reduced solvent capacity and increased carbon steel corrosion. Total corrosion of several materials relevant to CANDU stations were measured in 96-h tests. To minimize corrosion, low concentration of chelant and a high concentration of an oxygen scavenger should be used. The results from these feasibility tests are currently being used to define the application conditions for large-scale tests of on-line chemical cleaning in a model steam generator. (author)

  10. 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.

  11. 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.

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

    International Nuclear Information System (INIS)

    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

  13. 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)

  14. Development of an on-line process for steam generator chemical cleaning

    International Nuclear Information System (INIS)

    An on-line, preventative chemical cleaning process for the removal of secondary side oxides from steam generators is being developed. An on-line chemical cleaning process uses a low concentration of a chelant (∼1-10 mg L-1) to partially dissolve and dislodge the secondary side oxides while the steam generator is in operation. The dissolved and dislodged oxides can then be removed by blowdown. Feasibility tests were carried out in which the operating conditions of a CANDU steam generator were simulated in an autoclave containing either loose powdered magnetite or sintered magnetite on Alloy 800 (I-800) steam generator tube surfaces. The extent of magnetite dissolution in on-line solvent formulations containing either ethylenediaminetetraacetic acid (EDTA) or N-(2-hydroxyethyl)ethylenedinitrilo-N,N',N'-triacetic acid (HEDTA) at temperatures of 256 and 263 degrees C were measured. Powdered magnetite dissolved faster than sintered magnetite using both types of chelant. Dissolution continued as fresh chelant was added. The half-life (t1/2) of Fe-EDTA complexes at 256 degrees C was approximately 3 h, sufficient to allow removal by blowdown. Hydrazine and morpholine were equally effective as oxygen scavengers. Increased dissolved oxygen concentration was found to result in chelant decomposition, reduced solvent capacity and increased carbon steel corrosion. Total corrosion of several materials relevant to CANDU stations were measured in 96-h tests. To minimize corrosion, low concentration of chelant and a high concentration of an oxygen scavenger should be used. The results from these feasibility tests are currently being used to define the application conditions for large-scale tests of on-line chemical cleaning in a model steam generator. (author)

  15. The derivation of two-fluid, three-field governing equations in porous media using time-volume averaging formulation and its application to develop a safety analysis code

    International Nuclear Information System (INIS)

    sharp liquid-gas interface. Vessel blowdown problem is set up to examine the blowdown capability of the code. Vessel injection problem is to see the two-phase mixing. And some other tests are performed to clarify the applicability of the collocated unstructured semi-implicit scheme to two-fluid three-field flow problem

  16. Calibration of a four-hole pyramid probe and area traverse measurements in a short-duration transonic turbine cascade tunnel

    Science.gov (United States)

    Main, A. J.; Day, C. R. B.; Lock, G. D.; Oldfield, M. L. G.

    1996-08-01

    A four-hole pyramid probe has been calibrated for use in a short-duration transonic turbine cascade tunnel. The probe is used to create area traverse maps of total and static pressure, and pitch and yaw angles of the flow downstream of a transonic annular cascade. This data is unusual in that it was acquired in a short-duration (5 s of run time) annular cascade blowdown tunnel. A four-hole pyramid probe was used which has a 2.5 mm section head, and has the side faces inclined at 60° to the flow to improve transonic performance. The probe was calibrated in an ejector driven, perforated wall transonic tunnel over the Mach number range 0.5 1.2, with pitch angles from -20° to + 20° and yaw angles from-23° to +23°. A computer driven automatic traversing mechanism and data collection system was used to acquire a large probe calibration matrix (˜ 10,000 readings) of non dimensional pitch, yaw, Mach number, and total pressure calibration coefficients. A novel method was used to transform the probe calibration matrix of the raw coefficients into a probe application matrix of the physical flow variables (pitch, yaw, Mach number etc.). The probe application matrix is then used as a fast look-up table to process probe results. With negligible loss of accuracy, this method is faster by two orders of magnitude than the alternative of global interpolation on the raw probe calibration matrix. The blowdown tunnel (mean nozzle guide vane blade ring diameter 1.1 m) creates engine representative Reynolds numbers, transonic Mach numbers and high levels (≈ 13%) of inlet turbulence intensity. Contours of experimental measurements at three different engine relevant conditions and two axial positions have been obtained. An analysis of the data is presented which includes a necessary correction for the finite velocity of the probe. Such a correction is non trivial for the case of fast moving probes in compressible flow.

  17. 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

  18. Characterizing experiments of the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

    This report summarizes the results of the characterizing test series in 2007 with the scaled down PPOOLEX facility designed and constructed at Lappeenranta University of Technology. Air and steam/air mixture was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool (wet well). Altogether eight air and four steam/air mixture experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the general behavior of the facility and the performance of basic instrumentation. Proper operation of automation, control and safety systems was also tested. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. The facility is equipped with high frequency measurements for capturing different aspects of the investigated phenomena. The general behavior of the PPOOLEX facility differs significantly from that of the previous POOLEX facility because of the closed two-compartment structure of the test vessel. Heat-up by several tens of degrees due to compression in both compartments was the most obvious evidence of this. Temperatures also stratified. Condensation oscillations and chugging phenomenon were encountered in those tests where the fraction of non-condensables had time to decrease significantly. A radical change from smooth condensation behavior to oscillating one occurred quite abruptly when the air fraction of the blowdown pipe flow dropped close to zero. The experiments again demonstrated the strong diminishing effect that noncondensable gases have on dynamic unsteady loadings experienced by submerged pool structures. BWR containment like behavior related to the beginning of a postulated steam line break accident was observed in the PPOOLEX test facility during the steam/air mixture experiments. The most important task of the research project, to produce experimental data for code simulation purposes, can be

  19. 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

  20. 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

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

    International Nuclear Information System (INIS)

    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

  2. A comprehensive, mechanistic heat transfer modeling package for dispersed flow film boiling – Part 1 – Development

    Energy Technology Data Exchange (ETDEWEB)

    Meholic, Michael J., E-mail: michael.meholic@unnpp.gov [Bettis Atomic Power Laboratory, West Mifflin, PA (United States); Aumiller, David L. [Bettis Atomic Power Laboratory, West Mifflin, PA (United States); Cheung, Fan-Bill [The Pennsylvania State University, Department of Mechanical and Nuclear Engineering, University Park, PA (United States)

    2015-09-15

    Highlights: • A comprehensive, mechanistic heat transfer modeling package has been developed. • Accounts for six interrelated heat transfer paths in Dispersed Flow Film Boiling. • Lagrangian subscale trajectory based dry contact heat transfer model. • Novel methodology to account for droplet convective enhancement. - Abstract: Accurate predictions of Dispersed Flow Film Boiling (DFFB) heat transfer are necessary during both the blowdown and reflood portions of a Loss-of-Coolant-Accident to ensure the correct initial fuel rod temperature distribution for the beginning of the reflood phase and ultimately, determining the peak cladding temperature. Numerous correlative, phenomenological, and mechanistic DFFB heat transfer models have been published; however, most of these models make simplifying assumptions that adversely impact their accuracy or are too computationally intensive to implement into current reactor safety codes. A comprehensive, mechanistic heat transfer modeling package has been developed to account for the six interrelated heat transfer paths in DFFB. Highlights of the model include a Lagrangian subscale trajectory based dry contact heat transfer model and a novel method of determining the two-phase convective heat transfer enhancement due to dispersed droplets intermittently altering the local vapor temperature distribution.

  3. Short term hydrogen generation following LOCA and loss of ECCS

    International Nuclear Information System (INIS)

    The purpose of the present study is to estimate the amount of hydrogen that can be generated due to metal water reaction following LOCA and loss of ECCS in a 500 MWe PHWR. A computer code HYGEN (Hydrogen Generation) written in FORTRAN calculates time-dependent fuel temperature during the post blowdown period and the amount of hydrogen generated as a result of metal water reaction. It is seen from the analyses that metal water reaction depends on fuel bundle power, its initial temperature and steam flow conditions. At present, four groups of channels have been analysed for different steam flow conditions, and it is found that, for an about 5 gm/sec steam flow condition, the maximum of amount of hydrogen is generated (5.76 x 104 gm-mole) due to the zircaloy - steam reaction. This amount of hydrogen, when considered mixed in volume V1 (drywell) of the reactor building, means that the global concentration reaches about 2.76% by volume. So, it is seen that in the short term, the global hydrogen concentration in the reactor building is well below the flammability limit of 4% by volume. (author) 4 refs., 1 tab., 10 figs

  4. Post test calculations of some FR2-In-pile-tests with the program system SSYST-2

    International Nuclear Information System (INIS)

    The computations presented were made as part of the R+D task 'In-pile tests on LWR fuel rod behavior' within the framework of the Nuclear Safety Project (PNS) of the Kernforschungszentrum Karlsruhe (KfK). It was the purpose of this task to investigate the influence exerted by the nuclear parameters on the fuel rod behavior in a loss-of-coolant accident (LOCA). Shortlength pressurized-water reactor (PWR) fuel rods were used as specimens instead of the electrical simulators normally used in out-of-pile experiments of this type. The investigations had been restricted to the second heat-up phase of a loss-of-coolant accident which means that they did not include blow-down. It is known today that the rods possibly suffer serious damage - ballooning and bursting - in this second phase of the accident preferably. The tests were performed in the DK-loop of the FR2 reactor at KfK. Some of the tests were posttest calculated using the SSYST-2 fuel rod behavior code. (orig./RW)

  5. Review on Recent Advances in Pulse Detonation Engines

    Directory of Open Access Journals (Sweden)

    K. M. Pandey

    2016-01-01

    Full Text Available Pulse detonation engines (PDEs are new exciting propulsion technologies for future propulsion applications. The operating cycles of PDE consist of fuel-air mixture, combustion, blowdown, and purging. The combustion process in pulse detonation engine is the most important phenomenon as it produces reliable and repeatable detonation waves. The detonation wave initiation in detonation tube in practical system is a combination of multistage combustion phenomena. Detonation combustion causes rapid burning of fuel-air mixture, which is a thousand times faster than deflagration mode of combustion process. PDE utilizes repetitive detonation wave to produce propulsion thrust. In the present paper, detailed review of various experimental studies and computational analysis addressing the detonation mode of combustion in pulse detonation engines are discussed. The effect of different parameters on the improvement of propulsion performance of pulse detonation engine has been presented in detail in this research paper. It is observed that the design of detonation wave flow path in detonation tube, ejectors at exit section of detonation tube, and operating parameters such as Mach numbers are mainly responsible for improving the propulsion performance of PDE. In the present review work, further scope of research in this area has also been suggested.

  6. Detonation duct gas generator demonstration program

    Science.gov (United States)

    Wortman, Andrew; Brinlee, Gayl A.; Othmer, Peter; Whelan, Michael A.

    1991-01-01

    The feasibility of the generation of detonation waves moving periodically across high speed channel flow is experimentally demonstrated. Such waves are essential to the concept of compressing requirements and increasing the engine pressure compressor with the objective of reducing conventional compressor requirements and increasing the engine thermodynamic efficiency through isochoric energy addition. By generating transient transverse waves, rather than standing waves, shock wave losses are reduced by an order of magnitude. The ultimate objective is to use such detonation ducts downstream of a low pressure gas turbine compressor to produce a high overall pressure ratio thermodynamic cycle. A 4 foot long, 1 inch x 12 inch cross-section, detonation duct was operated in a blow-down mode using compressed air reservoirs. Liquid or vapor propane was injected through injectors or solenoid valves located in the plenum or the duct itself. Detonation waves were generated when the mixture was ignited by a row of spark plugs in the duct wall. Problems with fuel injection and mixing limited the air speeds to about Mach 0.5, frequencies to below 10 Hz, and measured pressure ratios of about 5 to 6. The feasibility of the gas dynamic compression was demonstrated and the critical problem areas were identified.

  7. Best effort analysis of critical large loss-of-coolant accident in Darlington NGS

    International Nuclear Information System (INIS)

    A best-effort analysis of Emergency Coolant Injection System (ECIS) effectiveness has been performed for a critical large break loss of coolant accident (LOCA) in Darlington NGS. This analysis, and various sensitivity analyses were performed using the best-effort version of the TUF two-fluid thermal-hydraulics code. The objective of this project is to develop analytical tools and analysis methodology to quantify, within reasonable bounds of certainty, the effectiveness of the ECIS in Ontario Hydro nuclear generating stations to limit activity releases from fuel in the event of a large break LOCA. As part of Best Effort ECIS effectiveness methodology, and the pilot application of this methodology to the analysis of Large LOCA for Darlington NGS, the TUF code has been developed to: quantify the degree of blowdown cooling in a multiple parallel channel reactor core; establish the minimum moderator subcooling required to ensure that fuel channel integrity is maintained, and determine the maximum time that the moderator is required to act as a heat sink; quantify the effectiveness of the ECIS to limit the extent of fuel and fuel channel heatup. The methodology described in this paper, together with enhancements to account for the effects of fuel string relocation, higher void reactivity uncertainty allowance and flux tilt on the initial overpower transient, has been implemented in the Generic Safety Report analysis to update the Large LOCA Safety Report sections for the Bruce and Pickering NGS. (author). 9 refs., 12 figs

  8. Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant

    International Nuclear Information System (INIS)

    The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories are used to perform scaled experiments that simulate High Pressure Melt Ejection accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt (thermite) is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic air/steam/hydrogen atmospheres, and hydrogen generation and combustion, can be studied. Four Integral Effects Tests (IETs) have been performed with scale models of the Surry NPP to investigate DCH phenomena. The 1/61th scale Integral Effects Tests (IET-9, IET-10, and IET-11) were conducted in CTRF, which is a 1/6th scale model of the Surry reactor containment building (RCB). The 1/10th scale IET test (IET-12) was performed in the Surtsey vessel, which had been configured as a 1/10th scale Surry RCB. Scale models were constructed in each of the facilities of the Surry structures, including the reactor pressure vessel, reactor support skirt, control rod drive missile shield, biological shield wall, cavity, instrument tunnel, residual heat removal platform and heat exchangers, seal table room and seal table, operating deck, and crane wall. This report describes these experiments and gives the results

  9. Shutdown Decay Heat Removal analysis of a Westinghouse 3-loop pressurized water reactor: Case study

    International Nuclear Information System (INIS)

    This is one of six case studies for USI A-45 Decay Heat Removal (DHR) Requirements. The purpose of this study is to identify any potential vulnerabilities in the DHR systems of a typical Westinghouse 3-loop PWR, to suggest possible modifications to improve the DHR capability, and to assess the value and impact of the most promising alternatives to the existing DHR systems. The systems analysis considered small LOCAs and transient internal initiating events, and seismic, fire, extreme wind, internal and external flood, and lightning external events. A full-scale systems analysis was performed with detailed fault trees and event trees including support system dependencies. The system analysis results were extrapolated into release categories using applicable past PRA phenomenological results and improved containment failure mode probabilities. Public consequences were estimated using site specific CRAC2 calculations. The Value-Impact (VI) analysis of possible alternatives considered both onsite and offsite impacts arriving at several risk measures such as averted population dose out to a 50-mile radius and dollars per person rem averted. Uncertainties in the VI analysis are discussed and the issues of feed and bleed and secondary blowdown are analyzed

  10. 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.

  11. Distribution of amines and organic acids in the secondary side of Embalse Nuclear Power Station

    International Nuclear Information System (INIS)

    In this work we summarized the distribution of amines and organic acids generated by the thermal decomposition of morpholine in the secondary side of Embalse NPP. Sampling and analytical procedures to determine the concentration of formic, acetic and glycolic acids, morpholine, ammonia, methylamine, ethanolamine and 2(2-aminoethoxy)ethanol are described. Two sets of samples were collected in March 1995 and October 1996 in the following points: main steam line, composite steam generator blowdown, moisture separator, condensate extraction pump discharge and outlet feed pump. The general trend of the product distribution along the secondary side is similar to that reported for other CANDU NPP. In CNE methylamine and ethanolamine are more abundant than 2(2-aminoethoxy)ethanol due to faster decomposition of morpholine and less oxidizing conditions. Ammonia, and methylamine concentrate in the steam because of the lack of a de-aerator. The volatility of ethanolamine is low and its concentration in the steam generator is high. It could help to neutralize acid conditions in crevices and sludges. The concentration of organic acids in CNE is low as compared with other CANDU NPP, with formic acid being the predominant species. Differences in the relative concentrations of morpholine degradation products as compared to other CANDU NPP are discussed. (author)

  12. Final environmental statement related to the operation of Callaway Plant, Unit No. 1 (Docket No. 50-483)

    International Nuclear Information System (INIS)

    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

  13. 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)

    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

  14. SOLOX coke-oven gas desulfurization ppm levels -- No toxic waste

    Energy Technology Data Exchange (ETDEWEB)

    Platts, M. (Thyssen Still Otto Technical Services, Pittsburgh, PA (United States)); Tippmer, K. (Thyssen Still Otto Anlagentechnik GmbH, Bochum (Germany))

    1994-09-01

    For sulfur removal from coke-oven gas, the reduction/oxidation processes such as Stretford are the most effective, capable of removing the H[sub 2]S down to ppm levels. However, these processes have, in the past, suffered from ecological problems with secondary pollutant formation resulting from side reactions with HCN and O[sub 2]. The SOLOX gas desulfurization system is a development of the Stretford process in which the toxic effluent problems are eliminated by installing a salt decomposition process operating according to the liquid-phase hydrolysis principle. In this process, the gaseous hydrolysis products H[sub 2]S, NH[sub 3] and CO[sub 2] are returned to the untreated gas, and the regenerated solution is recycled to the absorption process. The blowdown from the absorption circuit is fed into a tube reactor where the hydrolysis process takes place. The toxic salts react with water, producing as reaction products the gases H[sub 2]S, NH[sub 3] and CO[sub 2], and the nontoxic salt Na[sub 2]SO[sub 4]. From the hydrolysis reactor the liquid stream flows into a fractionating crystallization plant. This plant produces a recycle stream of regenerated absorption solution and a second stream containing most of the Na[sub 2]SO[sub 4]. This second stream comprises the net plant waste and can be disposed of with the excess ammonia liquor or sprayed onto the coal.

  15. Effectiveness of area and dedicated water deluge in protecting objects impacted by crude oil/gas jet fires on offshore installations

    Energy Technology Data Exchange (ETDEWEB)

    Hankinson, G. [Loughborough Univ., Dept. of Chemical Engineering, Loughborough (United Kingdom); Lowesmith, B.J. [Advantica Technologies Ltd., Loughborough (United Kingdom)

    2004-03-01

    A joint industry project (JIP) was undertaken to study the use of water deluge to reduce the hazards of fires on offshore installations. The project involved an extensive programme of large-scale experiments studying the effectiveness of area and dedicated deluge in mitigating jet and pool fires, and was sponsored by 11 oil and gas companies and the UK Health and Safety Executive. The work was conducted at the Advantica (formerly British Gas Research and Technology) Spadeadam Test Site, Cumbria, UK. This paper concentrates on a small part of the work performed during the second phase of the project that involved evaluating the effectiveness of area water deluge and dedicated (object specific), water deluge in reducing the heat loading to an object impacted by a crude oil/gas ('live' crude) jet fire. The results demonstrate that a combination of area and dedicated deluge can significantly reduce the heat loading on a critical item of plant such that its temperature is maintained below that at which catastrophic failure might occur, or such that the rate of temperature rise is reduced to a level that provides time for emergency shut down and blow-down to take place. In both cases, escalation is inhibited. (Author)

  16. Investigation of a two-phase nozzle flow and validation of several computer codes by the experimental data

    International Nuclear Information System (INIS)

    Stationary experiments with a convergent nozzle are performed in order to validate advanced two-phase computer codes, which find application in the blowdown-phase of a loss-of-coolant accident (LOCA). The steam/water flow presents a broad variety of initial conditions: The pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of subcritical as well as critical experiments with different flow pattern is investigated. Additional air/water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiments. The layout of the nozzle and the applied measurement technique allow for a separate testing of physical models and the determination of empirical model parameters, respectively: In the four codes DUESE, DRIX-20, RELAP4/MOD6 and STRUYA the models - if they exist - for slip between the phases, thermodynamic non-equilibrium, pipe friction and critical mass flow rate are validated and criticised in comparison with the experimental data, and the corresponding model parameters are determined. The parameters essentially are a function of the void fraction. (orig.)

  17. Causes of SiO2 Content Exceeding Standard in the Boiler Water after the Typhoon%台风过后锅炉炉水二氧化硅超标原因分析

    Institute of Scientific and Technical Information of China (English)

    杨晨; 梁昌英

    2014-01-01

    Analyze causes of the SiO2 content exceeding standard in the water of high pressure steam boiler . Take following measures:increasing analysis of steam drum water and the steam , increasing continous blowdown of steam drum water ,improving the water desalination preparation process ,and increasing analysis of total silicon content ,active silicon content and colloid silicon content in the water which entering into the water desalination device .Then the water quality returns to normal .%介绍台风过后,高压蒸汽锅炉炉水二氧化硅超标的原因。采取增加对汽包炉水和蒸汽的分析次数、增大汽包炉水的连续排污,改进对脱盐水制备过程,增加对进脱盐水装置水的全硅含量、活性硅含量和胶体硅含量的分析等措施,使锅炉水质恢复正常。

  18. Testing of Flexible Ballutes in Hypersonic Wind Tunnels for Planetary Aerocapture

    Science.gov (United States)

    Buck, Gregory M.

    2007-01-01

    Studies were conducted for the In-Space Propulsion (ISP) Ultralightweight Ballute Technology Development Program to increase the technical readiness level of inflatable decelerator systems for planetary aerocapture. The present experimental study was conducted to develop the capability for testing lightweight, flexible materials in hypersonic facilities. The primary objectives were to evaluate advanced polymer film materials in a high-temperature, high-speed flow environment and provide experimental data for comparisons with fluid-structure interaction modeling tools. Experimental testing was conducted in the Langley Aerothermodynamics Laboratory 20-Inch Hypersonic CF4 and 31-Inch Mach 10 Air blowdown wind tunnels. Quantitative flexure measurements were made for 60 degree half angle afterbody-attached ballutes, in which polyimide films (1-mil and 3- mil thick) were clamped between a 1/2-inch diameter disk and a base ring (4-inch and 6-inch diameters). Deflection measurements were made using a parallel light silhouette of the film surface through an existing schlieren optical system. The purpose of this paper is to discuss these results as well as free-flying testing techniques being developed for both an afterbody-attached and trailing toroidal ballute configuration to determine dynamic fluid-structural stability. Methods for measuring polymer film temperature were also explored using both temperature sensitive paints (for up to 370 C) and laser-etched thin-film gages.

  19. Experiments to investigate the effect of flight path on direct containment heating (DCH) in the Surtsey test facility

    International Nuclear Information System (INIS)

    The goal of the Limited Flight Path (LFP) test series was to investigate the effect of reactor subcompartment flight path length on direct containment heating (DCH). The test series consisted of eight experiments with nominal flight paths of 1, 2, or 8 m. A thermitically generated mixture of iron, chromium, and alumina simulated the corium melt of a severe reactor accident. After thermite ignition, superheated steam forcibly ejected the molten debris into a 1:10 linear scale the model of a dry reactor cavity. The blowdown steam entrained the molten debris and dispersed it into the Surtsey vessel. The vessel pressure, gas temperature, debris temperature, hydrogen produced by steam/metal reactions, debris velocity, mass dispersed into the Surtsey vessel, and debris particle size were measured for each experiment. The measured peak pressure for each experiment was normalized by the total amount of energy introduced into the Surtsey vessel; the normalized pressures increased with lengthened flight path. The debris temperature at the cavity exit was about 2320 K. Gas grab samples indicated that steam in the cavity reacted rapidly to form hydrogen, so the driving gas was a mixture of steam and hydrogen. These experiments indicate that debris may be trapped in reactor subcompartments and thus will not efficiently transfer heat to gas in the upper dome of a containment building. The effect of deentrainment by reactor subcompartments may significantly reduce the peak containment load in a severe reactor accident. 8 refs., 49 figs., 6 tabs

  20. Results of direct containment heating integral experiments at 1/40th scale at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Binder, J.L.; McUmber, L.M.; Spencer, B.W.

    1993-09-01

    A series of integral tests have been completed that investigate the effect of scale and containment atmosphere initial composition on Direct Containment Heating (DCH) phenomena at 1/40 linear scale. A portion of these experiments were performed as counterparts to integral experiments conducted at 1/10th linear scale at Sandia National Laboratories. The tests investigated DCH phenomena in a 1/40th scale mockup of Zion Nuclear Power Plant geometry. The test apparatus was a scaled down version of the SNL apparatus and included models of the reactor vessel lower head, containment cavity, instrument tunnel, lower subcompartment structures and the upper dome. A High Pressure Melt Ejection (HPME) was produced using steam as a blowdown gas and iron-alumina thermite with chromium as a core melt simulant. The results of the counterpart experiments indicated no effect of scale on debris/gas heat transfer and debris metal oxidation with steam. However, the tests indicated a slight effect of scale on hydrogen combustion, the results indicating slightly more efficient combustion with increasing scale. The experiments demonstrated the effectiveness of the subcompartment structures in trapping debris exiting the cavity and preventing it from reaching the upper dome. The test results also indicated that a 50% air -- 50% steam atmosphere prevented hydrogen combustion. However, a 50% air - 50% nitrogen did not prevent hydrogen combustion in a HPME with all other conditions being nominally the same.

  1. Reactor Safety Research Programs Quarterly Report January - March 1980

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty 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 steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  2. An experimental parametric study of the high pressure melt ejection from two different scale reactor cavity models

    International Nuclear Information System (INIS)

    A parametric study of the high pressure melt ejection(HPME) from two small-scale(1/25th and 1/41st) transparent reactor cavity models of the YoungGwang(unit 1 and 2) has been performed. Wood's metal and water have been used as melt simulants while high pressure nitrogen and carbon dioxide are used as driver gases to simulate the blowdown steam and gas from the breach of the reactor pressure vessel. Experimental data for the fraction of melt simulant retained in the cavity model(Yf) during a postulated scenario of the HPME from PWR pressure vessel have been obtained as a function of various test parameters. These data have been used to develop a correlation for Yf that fits all the data(a total of 313 data points) within the standard deviation of 0.054 by means of dimensional analysis and nonlinear least-squares optimization technique. The basic effects of important parameters used to describe the HPME accident sequence on the Yf are also determined based on the correlation obtained here and experimental evidence. (Author)

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

    International Nuclear Information System (INIS)

    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)

  4. Suppression Pool Mixing and Condensation Tests in PUMA Facility

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  6. Fluid-structure interaction in BWR suppression pool systems. Final report. [PELE-IC code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. 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 [KOPEC, Yongin (Korea, Republic of); Kim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [KEPCO, Seoul (Korea, Republic of)

    1998-05-01

    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 Coodown (NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown pheonomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifice 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 test, 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.

  8. TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

    International Nuclear Information System (INIS)

    An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

  9. Preliminary regulatory audit calculation for Shinkori Units 3 and 4 LBLOCA

    International Nuclear Information System (INIS)

    The objective of this study is to perform a preliminary evaluation for Shinkori Units 3 and 4 LBLOCA by applying KINS Realistic Evaluation Methodology (REM). The following results were obtained: (1) From the evaluation for Shinkori Units 3 and 4 LBLOCA, the peak cladding temperature was evaluated to meet the regulatory requirement and the feasibility of the KINS-REM was identified. (2) The input decks that were developed in the previous studies, were reviewed and the evaluation model of the fluidic device was developed and applied for the audit calculation. (3) The treating method for the uncertainty of the gap conductance was developed and applied for the audit calculation. (4) The pre- and post-processing programs were developed for this study. (5) For the more detailed assessments, the information for the gap conductance, etc. should be improved and the effects of coolant bypass during blowdown, steam binding and so on were not sufficiently evaluated. KINS-REM should be advanced to evaluate these effects properly. The KINS methodology that was used in this study, can be further applied for independent regulatory audit calculations related to the licensing application on LOCA best estimate calculation

  10. 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)

  11. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  12. 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)

  13. Depressurization of Vertical Pipe with Temperature Gradient Modeled with WAHA Code

    Directory of Open Access Journals (Sweden)

    Oriol Costa

    2012-01-01

    Full Text Available The subcooled decompression under temperature gradient experiment performed by Takeda and Toda in 1979 has been reproduced using the in-house code WAHA version 3. The sudden blowdown of a pressurized water pipe under temperature gradient generates a travelling pressure wave that changes from decompression to compression, and vice versa, every time it reaches the two-phase region near the orifice break. The pressure wave amplitude and frequency are obtained at different locations of the pipe's length. The value of the wave period during the first 20 ms of the experiment seems to be correct but the pressure amplitude is overpredicted. The main three parameters that contribute to the pressure wave behavior are: the break orifice (critical flow model, the ambient pressure at the outlet, and the number of volumes used for the calculation. Recent studies using RELAP5 code have reproduced the early pressure wave (transient of the same experiment reducing the discharge coefficient and the bubble diameter. In the present paper, the long-term pipe pressure, that is, 2 seconds after rupture, is used to estimate the break orifice that originates the pressure wave. The numerical stability of the WAHA code is clearly proven with the results using different Courant numbers.

  14. Process Control for Simultaneous Vitrification of Two Mixed Waste Streams in the Transportable Vitrification System

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A.D. [Westinghouse Savannah River Company, AIKEN, SC (United States); Jantzen, C.M.; Brown, K.G.; Cicero-Herman, C.

    1998-05-01

    Two highly variable mixed (radioactive and hazardous) waste sludges were simultaneously vitrified in an EnVitCo Transportable Vitrification System (TVS) deployed at the Oak Ridge Reservation. The TVS was the result of a cooperative effort between the Westinghouse Savannah River Company and EnVitCo to design and build a transportable melter capable of vitrifying a variety of mixed low level wastes.The two waste streams for the demonstration were the dried B and C Pond sludges at the K-25 site and waste water sludge produced in the Central Neutralization Facility from treatment of incinerator blowdown. Large variations occurred in the sodium, calcium, silicon, phosphorus, fluorine and iron content of the co- blended waste sludges: these elements have a significant effect on the process ability and performance of the final glass product. The waste sludges were highly reduced due to organics added during processing, coal-pile runoff (coal and sulfides), and other organics, including wood chips. A batch-by-batch process control model was developed to control glass viscosity, liquidus, and reduction/oxidation, assuming that the melter behaved as a Continuously Stirred Tank Reactor.

  15. Structural modelling and testing of failed high energy pipe runs: 2D and 3D pipe whip

    Energy Technology Data Exchange (ETDEWEB)

    Reid, S.R., E-mail: steve.reid@abdn.ac.uk [School of Engineering, University of Aberdeen, Aberdeen AB24 3UE (United Kingdom); Wang, B.; Aleyaasin, M. [School of Engineering, University of Aberdeen, Aberdeen AB24 3UE (United Kingdom)

    2011-05-15

    The sudden rupture of a high energy piping system is a safety-related issue and has been the subject of extensive study and discussed in several industrial reports (e.g. ). The dynamic plastic response of the deforming pipe segment under the blow-down force of the escaping liquid is termed pipe whip. Because of the potential damage that such an event could cause, various geometric and kinematic features of this phenomenon have been modelled from the point of view of dynamic structural plasticity. After a comprehensive summary of the behaviour of in-plane deformation of pipe runs that deform in 2D in a plane, the more complicated case of 3D out-of-plane deformation is discussed. Both experimental studies and modelling using analytical and FE methods have been carried out and they show that, for a good estimate of the 'hazard zone' when unconstrained pipe whip motion could occur, a large displacement analysis is essential. The classical, rigid plastic, small deflection analysis (e.g. see ), is valid for estimating the initial failure mechanisms, however it is insufficient for describing the details and consequences of large deflection behaviour. - Highlights: > Dynamic plastic response of piping system under extreme loading (fluid escape). > Two and three dimensional analysis of the pipe whipping phenomena. > Comparison between theory and the experiments. > Determination of the Hazard Zone (HZ) and safety-related issues.

  16. Radioactive Waste Facilities at the Australian Atomic Energy Commission Research Establishment

    International Nuclear Information System (INIS)

    This paper describes the facilities,which are being provided for the collection, treatment and disposal of radioactive wastes at Lucas Heights in relation to the estimated arisings. Low-activity effluent is divided into three types: (a) Sewage; (b) Trades waste, arising from reactor cooling tower blow-down and engineering workshops and other inactive areas; and (c) Effluent arising from laboratories and other active areas. The effluent treatment plant for the latter type of effluent consists essentially of mixing and alkali dosing tanks, a sludge-blanket clarifier (using a calcium- iron-phosphate process) and holding tanks. Methods of concentrating the sludge and of secondary treatment are at present being investigated and are discussed. The discharge formula and the expected dilution obtained in the Woronora river are discussed, together with a dilution experiment carried out in the tidal waters. It is proposed to bury all low-activity solid waste after baling where appropriate and the choice and location of the disposal area is discussed. A facility for the storage and disposal of highly active solid waste is discussed. It is proposed to evaporate and store the medium- and high-activity liquid waste. Details are given of the capital and operating costs of the Effluent Treatment Plant and other waste handling facilities. (author)

  17. A modelling study of the multiphase leakage flow from pressurised CO2 pipeline.

    Science.gov (United States)

    Zhou, Xuejin; Li, Kang; Tu, Ran; Yi, Jianxin; Xie, Qiyuan; Jiang, Xi

    2016-04-01

    The accidental leakage is one of the main risks during the pipeline transportation of high pressure CO2. The decompression process of high pressure CO2 involves complex phase transition and large variations of the pressure and temperature fields. A mathematical method based on the homogeneous equilibrium mixture assumption is presented for simulating the leakage flow through a nozzle in a pressurised CO2 pipeline. The decompression process is represented by two sub-models: the flow in the pipe is represented by the blowdown model, while the leakage flow through the nozzle is calculated with the capillary tube assumption. In the simulation, two kinds of real gas equations of state were employed in this model instead of the ideal gas equation of state. Moreover, results of the flow through the nozzle and measurement data obtained from laboratory experiments of pressurised CO2 pipeline leakage were compared for the purpose of validation. The thermodynamic processes of the fluid both in the pipeline and the nozzle were described and analysed. PMID:26774983

  18. 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).

  19. 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.

  20. Handbook for estimating methane emissions from Canadian natural gas systems

    International Nuclear Information System (INIS)

    An overall framework by which companies in the natural gas industry can assess their methane emissions was described. A set of terms, source categories and nomenclature was established to allow for comparisons and easy aggregation of methane-emissions data for future industry reporting initiatives. This handbook deals only with methane emissions from non-combustion sources such as process venting and fugitive equipment leaks at gas transmission, storage and distribution facilities. The common assessment methods to estimate methane emissions for each applicable source type are summarized. Target source types include: (1) fugitive pipeline and equipment leaks, (2) emissions from third-party damage, (3) blowdown events, (4) purging activities, (5) still-column vents on glycol dehydrators, (6) compressor starts, (7) use of natural gas as the supply medium for gas-operated devices, and (8) venting by gas samples, analyzers, relief valves and regulators. The need for good quality assurance control measures to ensure reliable results is emphasized. 36 refs., 19 tabs., 7 figs., 1 appendix

  1. Pilot-scale cooling tower to evaluate corrosion, scaling, and biofouling control strategies for cooling system makeup water.

    Science.gov (United States)

    Chien, S H; Hsieh, M K; Li, H; Monnell, J; Dzombak, D; Vidic, R

    2012-02-01

    Pilot-scale cooling towers can be used to evaluate corrosion, scaling, and biofouling control strategies when using particular cooling system makeup water and particular operating conditions. To study the potential for using a number of different impaired waters as makeup water, a pilot-scale system capable of generating 27,000 kJ∕h heat load and maintaining recirculating water flow with a Reynolds number of 1.92 × 10(4) was designed to study these critical processes under conditions that are similar to full-scale systems. The pilot-scale cooling tower was equipped with an automatic makeup water control system, automatic blowdown control system, semi-automatic biocide feeding system, and corrosion, scaling, and biofouling monitoring systems. Observed operational data revealed that the major operating parameters, including temperature change (6.6 °C), cycles of concentration (N = 4.6), water flow velocity (0.66 m∕s), and air mass velocity (3660 kg∕h m(2)), were controlled quite well for an extended period of time (up to 2 months). Overall, the performance of the pilot-scale cooling towers using treated municipal wastewater was shown to be suitable to study critical processes (corrosion, scaling, biofouling) and evaluate cooling water management strategies for makeup waters of complex quality.

  2. Long-Term Water-Quality Changes in East Fork Poplar Creek, Tennessee: Background, Trends, and Potential Biological Consequences

    Science.gov (United States)

    Stewart, Arthur J.; Smith, John G.; Loar, James M.

    2011-06-01

    We review long-term changes that have occurred in factors affecting water quality in East Fork Poplar Creek (EFPC; in East Tennessee) over a nearly 25-year monitoring period. Historically, the stream has received wastewaters and pollutants from a major United States Department of Energy (DOE) facility on the headwaters of the stream. Early in the monitoring program, EFPC was perturbed chemically, especially within its headwaters; evidence of this perturbation extended downstream for many kilometers. The magnitude of this perturbation, and the concentrations of many biologically significant water-quality factors, has lessened substantially through time. The changes in water-quality factors resulted from a large number of operational changes and remedial actions implemented at the DOE facility. Chief among these were consolidation and elimination of many effluents, elimination of an unlined settling/flow equalization basin, reduction in amount of blow-down from cooling tower operations, dechlorination of effluents, and implementation of flow augmentation. Although many water-quality characteristics in upper EFPC have become more similar to those of reference streams, conditions remain far from pristine. Nutrient enrichment may be one of the more challenging problems remaining before further biological improvements occur.

  3. DMAIC makes solutions possible at Surmont

    Energy Technology Data Exchange (ETDEWEB)

    Petkau, R.

    2010-09-15

    This article discussed how the Lean Six Sigma business management practice was successfully applied to solve a scaling problem at a steam-assisted gravity drainage (SAGD) facility operated by ConocoPhillips at Surmont. Lean Six Sigma seeks to improve the quality of process outputs by identifying and removing the causes of defects and minimizing variability. The Six Sigma method for the existing facility was executed by the define, measure, analyse, improve, and control (DMAIC) problem-solving process. The scaling problem was causing the company to spend too much and lose too much production. Pigging every 2 years was identified as the goal. In the measure stage, it was determined that bitumen in water was staying mostly in the generator. Oil field culture was identified as a hindrance to solving the scaling problem. Several other contributing factors were identified, including dissolved organics reduction, online turbidity meters, removing pH flush, flocculant system reliability, and blowdown recycle. Work is ongoing to reach the 24-month target. The most challenging part of the DMAIC process was system control, notably maintaining operations regardless of changes in company personnel. The company will continue using the Lean Six Sigma methodology to solve problems. 2 figs.

  4. 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)

    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

  5. Silica scale technology and water conservation. [Recirculating evaporative cooling

    Energy Technology Data Exchange (ETDEWEB)

    Midkiff, W.S.; Foyt, H.P.

    1976-01-01

    Conservation of water at the Los Alamos Scientific Laboratory (LASL) has been accomplished by recirculating evaporative cooling waters. Because of high silica concentration (80 mg/l) in Los Alamos groundwater, the concentration of recirculating water must be carefully controlled to prevent scaling. The most troublesome scale at Los Alamos has been identified as colloidal silica bound in a crystalline matrix of calcium carbonate. Several approaches to controlling this scale are: (1) chemical treatment using a chelate, sequestrant, or threshold approach, (2) softening, or (3) pH control. Silica alone will form deposits when supersaturated. In LASL systems, where silica concentrations are 200 to 240 mg/l, no problems have been observed. However, there is evidence that deposits are forming at slightly higher concentrations. These amorphous silica deposits are not as hard and tenacious as the calcium carbonate--silica scale. Complete external treatment, which combines silica removal and water softening, may be an economically competitive process for scale control. The advantages of slightly reducing the quantity of makeup water and drastically reducing the amount of blowdown water have environmental and conservation implications that may encourage the selection of complete treatment.

  6. Measurement and control systems for steam boilers; Regel- und Steuertechnik fuer Dampfkessel. Betriebsmanagement

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, H. [Loos Deutschland GmbH, Loos International, Gunzenhausen (Germany)

    2002-09-01

    Loos International has been building its own switchgear for Loos boiler systems since the 1960s. An operating management system for Loos steam boilers, based on a self-contained automation unit, has also been developed. All boiler control functions are combined on a single SPC with one central operating unit. The use of controlled-speed feed pumps is new. Integrated pump protection functions for constant level control mean that there is no need for the usual constant control module with overflow recirculation. Other LBC functions are the desalination control and automatic blowdown systems. Additional measurement and control parameters can be added to be basic functions of a modern steam boiler. Unlike usual configurations, the basic LBC configuration offers substantially more steam boiler operating modes, operating data and measurands on a plain language display, facilitating rapid inspection and assessment of all operating contexts on site. A high level of planning and operating reliability is achieved. This measurement and control technology for steam boilers offers many advantages to planners, plant constructors and other clients. (orig.) [German] Das Unternehmen Loos International in Gunzenhausen bietet Heiz- und Dampfkessel, Kesselhauskomponenten, Feuerung und Steuerung aus einer Hand. Loos hat ein Betriebsmanagement mit integriertem Vorwarnsystem fuer Loos-Dampfkessel entwickelt, bei welchem Betriebsdaten und Messwerte des Dampfkessels im Klartextdisplay angezeigt werden koennen. Der Autor stellt das System und dessen Vorteile und Nutzen im folgenden Beitrag vor. (orig.)

  7. New design architecture decisions on water chemistry support systems at new VVER plants

    Energy Technology Data Exchange (ETDEWEB)

    Kumanina, V.E.; Yurmanova, A.V. [Joint Stock Company Atomenergoproekt, Moscow (Russian Federation)

    2010-07-01

    Major goals of nuclear power plant design upgrading are reduction of cost and construction time with unconditional safety assurance. Main ways of further improvement of nuclear power plant design are as follows: review of the results of research engineering and development and of new technologies; harmonization with international codes and standards; justified liberalization of conservatism based on operating experience and use of improved design codes. Operational experience of Russian and foreign NPPs has shown that the designs of new NPPs could be improved by upgrading water chemistry support systems. Some new design solutions for water chemistry support systems are currently implemented at new WWER plants such as Bushehr, Kudankulam, Belene, Balakovo Units 5 and 6, AES-2006 project. The paper highlights the improvements of the following systems and processes: low temperature high pressure primary coolant clean-up system; primary system surface preconditioning during pre-start hot functional testing; steam generator blowdown cleanup system; secondary water chemistry; phosphate water chemistry in intermediate cooling circuits and other auxiliary systems; alternator cooling system water chemistry; steam generator cleanup and decontamination systems. (author)

  8. New design architecture decisions on water chemistry support systems at new VVER plants

    International Nuclear Information System (INIS)

    Major goals of nuclear power plant design upgrading are reduction of cost and construction time with unconditional safety assurance. Main ways of further improvement of nuclear power plant design are as follows: review of the results of research engineering and development and of new technologies; harmonization with international codes and standards; justified liberalization of conservatism based on operating experience and use of improved design codes. Operational experience of Russian and foreign NPPs has shown that the designs of new NPPs could be improved by upgrading water chemistry support systems. Some new design solutions for water chemistry support systems are currently implemented at new WWER plants such as Bushehr, Kudankulam, Belene, Balakovo Units 5 and 6, AES-2006 project. The paper highlights the improvements of the following systems and processes: low temperature high pressure primary coolant clean-up system; primary system surface preconditioning during pre-start hot functional testing; steam generator blowdown cleanup system; secondary water chemistry; phosphate water chemistry in intermediate cooling circuits and other auxiliary systems; alternator cooling system water chemistry; steam generator cleanup and decontamination systems. (author)

  9. Study on modeling of pipe whipping by finite element method

    International Nuclear Information System (INIS)

    The general purpose finite element codes ADINA and MARC were used to make a preliminary analysis for the pipe whip tests performed with use of 4B, sch80 test pipes under the saturated water condition (pressure = 69 kg/cm2G, temperature = 284.50C). In the analysis the initial clearance between the test pipe and the restraints was taken to be equal to 30 mm, 50 mm and 100 mm for RUN No. 5405, 5406 and 5407, respectively. On the other hand, the overhang length was kept equal to 400 mm for the above three cases. The experimental result of the jet discharge test, RUN No. 5401, and the analytical results obtained from the PRTHRUST-J1 code for calculation of blowdown thrust force were used as the time history of loading of the pipe whip analyses. In the analyses, various models of the test pipe and the restraints were employed to study the effect of modeling on the pipe whip behavior and the guides to the analysis of pipe whip in the future are also presented. (author)

  10. Leaching of asbestos-cement cooling-tower fill. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, C.N.; Stone, R.W.

    1981-04-01

    Cooling-tower fill is sometimes made of asbestos cement. Asbestos-cement fill has frequently been damaged by leaching and mechanical problems. This leaching was investigated. Previous studies of asbestos-cement water pipe and cooling-tower fill are summarized. Five plants were visited, and 43 others were contacted by telephone. Water and fill samples were collected and analyzed. About half of the cooling towers with asbestos-cement fill have experienced significant deterioration. To control leaching, water should not be undersaturated with respect to calcium carbonate. The Langelier saturation index is a useful tool for controlling blowdown rates and chemical feed. However, because this index does not allow for all of the relevant factors, it is not possible to recommend values that are suitable for all plants. If no scale inhibitors are used, the index should be kept as high as possible without causing calcium carbonate scale. If scale inhibitors are used, overdosing should be avoided. Asbestos-cement fill should be used only if the cooling-water chemistry can be well controlled. Specifications for asbestos-cement fill can be improved. Other design features, operating practices, and research are suggested.

  11. Surge Pressure Mitigation in the Global Precipitation Measurement Mission Core Propulsion System

    Science.gov (United States)

    Scroggins, Ashley R.; Fiebig, Mark D.

    2014-01-01

    The Global Precipitation Measurement (GPM) mission is an international partnership between NASA and JAXA whose Core spacecraft performs cutting-edge measurements of rainfall and snowfall worldwide and unifies data gathered by a network of precipitation measurement satellites. The Core spacecraft's propulsion system is a blowdown monopropellant system with an initial hydrazine load of 545 kg in a single composite overwrapped propellant tank. At launch, the propulsion system contained propellant in the tank and manifold tubes upstream of the latch valves, with low-pressure helium gas in the manifold tubes downstream of the latch valves. The system had a relatively high beginning-of- life pressure and long downstream manifold lines; these factors created conditions that were conducive to high surge pressures. This paper discusses the GPM project's approach to surge mitigation in the propulsion system design. The paper describes the surge testing program and results, with discussions of specific difficulties encountered. Based on the results of surge testing and pressure drop analyses, a unique configuration of cavitating venturis was chosen to mitigate surge while minimizing pressure losses during thruster maneuvers. This paper concludes with a discussion of overall lessons learned with surge pressure testing for NASA Goddard spacecraft programs.

  12. The MAP Propulsion Subsystem

    Science.gov (United States)

    Davis, Gary T.; Bauer, Frank H. (Technical Monitor)

    2002-01-01

    This paper describes the requirements, design, integration, test, performance, and lessons learned of NASA's Microwave Anisotropy Probe (MAP) propulsion subsystem. MAP was launched on a Delta-II launch vehicle from NASA's Kennedy Space Center on June 30, 2001. Due to instrument thermal stability requirements, the Earth-Sun L2 Lagrange point was selected for the mission orbit. The L2 trajectory incorporated phasing loops and a lunar gravity assist. The propulsion subsystem's requirements are to manage momentum, perform maneuvers during the phasing loops to set up the lunar swingby, and perform stationkeeping at L2 for 2 years. MAP's propulsion subsystem uses 8 thrusters which are located and oriented to provide attitude control and momentum management about all axes, and delta-V in any direction without exposing the instrument to the sun. The propellant tank holds 72 kg of hydrazine, which is expelled by unregulated blowdown pressurization. Thermal management is complex because no heater cycling is allowed at L2. Several technical challenges presented themselves during I and T, such as in-situ weld repairs and in-situ bending of thruster tubes to accommodate late changes in the observatory CG. On-orbit performance has been nominal, and all phasing loop, mid-course correction, and stationkeeping maneuvers have been successfully performed to date.

  13. 284-E Powerplant Wastewater stream-specific report

    International Nuclear Information System (INIS)

    The proposed wastestream designation for the 284-E Powerplant Wastewater stream is that it is not a dangerous waste, pursuant to the Washington (State) Administrative Code (WAC) 173-303, Dangerous Waste Regulations. This proposed designation is based on applying both process knowledge and sample data to the WAC 173-303 requirements for the three types of dangerous waste: listed; criteria; and characteristic dangerous waste. The ''listed'' dangerous waste determination was made with process knowledge; the ''criteria'' and ''characteristic'' dangerous waste determinations were made with sampling data. Process knowledge was based on knowledge of 284-E Powerplant operations. Sample data are based on samples downstream of all process contributors. The proposed designation is made using ''validated'' data from routine operations samples taken from October 1989 through march 1990. Samples of the other two waste contributing activities, blowdown and softener regeneration, were taken prior to implementation of a data validation procedure. These data are included in Appendix A to further support the proposed designation. 8 refs., 5 figs., 7 tabs

  14. Evaluation of hideout return data from U.S. PWR steam generators

    International Nuclear Information System (INIS)

    Since the middle to late 1970's, dramatic reductions in the quantities of impurities in the bulkwater of PWR steam generators have been made by U.S. utilities. Today most utilities operate at full power with impurity concentrations in the steam generator blowdown in the low ppb range, well within existing industry guideline control limits. Despite these efforts, some of these same utilities have subsequently encountered secondary side stress corrosion cracking (SCC) and intergranular attack (IGA) of steam generator tubing within deep tubesheet crevices and more recently at tube support intersections. It must, therefore, be concluded that either continuous low level input of contaminants within existing guideline limits, or intermittent short duration input, undetected by either current sampling and analysis techniques or procedures, are permitting ingress of corrosive impurity species which subsequently concentrate in flow-occluded regions to produce localized tube corrosion. To better understand both the quantity and composition of accumulated impurity species, more and more utilities, even those who have not experienced any steam generator corrosion, have begun to perform rigorous sampling and analysis evaluations of returning chemical contaminants each time the units are brought off-line. This paper will show examples of how these data are being used by U.S. industry to gain valuable information about accumulated contaminant inventories, to make cycle-to-cycle and plant-to-plant comparisons, and to develop plant specific actions to promote maximum contaminant removal. (author)

  15. Probabilistic consequence assessment of hydrogen sulphide releases from a heavy water plant

    International Nuclear Information System (INIS)

    This report is concerned with the evaluation of the consequences to the public of an accidental release of hydrogen sulphide (H2S) 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 H2S from a rupture, the initial gas cloud behaviour, atmospheric dispersion and the toxicity of H2S and sulphur dioxide (SO2) 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

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

    International Nuclear Information System (INIS)

    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)

  17. Investigations on multicycle spray detonations

    Institute of Scientific and Technical Information of China (English)

    LI Mu; YAN Chuanjun; ZHENG Longxi; WANG Zhiwu; QIU Hua

    2007-01-01

    Experimental investigations were carried out on a 50-I.D. Multicycle pulse detonation engine (PDE) model, and liquid fuel (gasoline) was used. The average of pressure peak, as measured by piezoelectricity pressure transducer, increased versus distance to thrust wall before fully-developed detonation came into being. According to the pressure history, the pressure in detonation tube would not rise abruptly until the flame front advanced a certain distance downstream the spark. Just at that moment, two compression waves spreading to opposite direction were formed. One was enforced by combustion and became detonation rapidly. The other was weakened because of obstacles and insufficiency of fuel. Two methods were used to determine the induction length of two-phase detonation wave through the pressure history. Ignition delay time was found to be longer than deflagration-to-detonation transition (DDT) time, and the sum of the two would change little as cycle frequency increased. So they could be the most important factors controlling two-phase PDE frequency. Filling process and blowdown process were also analyzed.

  18. A cooling water system copper corrosion study

    Energy Technology Data Exchange (ETDEWEB)

    Pulkrabek, J.W.

    1998-07-01

    The plant has four units that have been operating normally for 12--33 years. Two of the units are 70 MW sister units that have copper alloy once-through condensers. The other two units are 350 MW and 500 MW units with copper alloy condensers and cooling towers. No cooling water related tube leaks had been experienced. Until 1993, the only chemicals used were sulfuric acid for pH control of the cooling tower systems and chlorine for biological control. The units were chlorinated for one hour per day per condenser. In early July 1992, their copper grab sample at the plant discharge to the river exceeded the weekly environmental limit. In fact, it was so high that there was a slim chance of coming in under their monthly average copper limit unless something was done quickly. The result of this incident was an extensive study of their plant wastewater and cooling systems. The study revealed that the elevated copper problem had existed sporadically for several years. Initially, copper control was achieved by altering the wastewater treatment processes and cooling tower blowdown flow path. Two extended trials, one with tolyltriazole (TTA) and one with a chemically modified benzotriazole (BZT) were performed. Optimal control of copper corrosion was eventually achieved by the application of a TTA treatment program in which the feed rates are adjusted based on on-line corrosion monitoring measurements. This report documents experiences and results over the past six years.

  19. Role of Nurse Logs in Forest Expansion at Timberline

    Science.gov (United States)

    Johnson, A. C.; Yeakley, A.

    2008-12-01

    Nurselogs, known to be key sites of forest regeneration in lower elevation temperate forests, may be important sites for seedling establishment at expanding timberline forests. To determine factors associated with seedling establishment and survival on nurselogs at timberline, fourteen sites, located across a precipitation gradient in the Washington North Cascades Mountains, were examined. Site attributes including seedling type and height, disturbance process introducing downed wood, wood decay type, shading, slope gradient, aspect, and temperature and water content of wood and adjacent soil were determined along 60 m long transects. Nurselogs were found at 13 out of 14 sites; sites typically associated with greater than 80% shade and downed wood having a high level of wood decay. Downed wood serving as nurselogs originated from blowdown, snow avalanches, and forest fires. In total, 46 of 136 downed wood pieces observed served as nurselogs. Seedlings on nurselogs included mountain hemlock (Tsuga mertensiana), Pacific silver fir (Abies amabilis), yellow cedar (Chamaecyparis nootkatensis), subalpine fir (Abies lasiocarpa), Engelmann spruce (Picea engelmannii), and western larch (Larix occidentalis). Nurselogs had significantly higher temperatures (p = 0.015) and higher moisture contents (p = 0.019) than the adjacent soil. Per equal volumes weighed, nurselogs had on average of 23.8 g more water than the adjacent soil. Given predictions of climate warming and associated summer drought conditions in Pacific Northwest forests, the moisture provided by nurselogs may be integral for conifer survival and subsequent timberline expansion in some landscapes.

  20. Mark III confirmatory test program: one-third scale pool swell impact tests, Test Series 5805

    International Nuclear Information System (INIS)

    A series of 51 blowdown tests was performed in support of the Mark III pressure suppression concept with particular emphasis on the effect of pool swell impact on structures located above the suppression pool. The integrated steam generator and drywell of the Pressure Suppression Test Facility was used to accelerate the water mass in the one-third scale suppression pool to velocities typical of Mark III containments, and the impact of this water on I-beams, pipes, and gratings was investigated. The loading mechanism was found to be high velocity pressure waves which traveled along the surface of impacted structures, with a wave velocity defined by the movement of the points of intersection between the horizontal target structures and the rising curved pool surface. The impulse associated with this loading was found to correlate as a function of pool approach velocity, target geometry, and water ligament thickness, the last variable being important only when the ligament thickness approached target dimensions. For pool surface velocities expected to occur in Mark III, the maximum measured impulses for all targets were 35 percent or less of those being used for Mark III design specifications. For targets of circular cross section, loads were one-half or less than the values for comparable flat surfaces. Both the factor of three and the pipe shape factor must be considered when evaluating the conservatism in the Mark III design specifications

  1. Cold leg injection reflood test results in the SCTF Core-I under constant system pressure

    International Nuclear Information System (INIS)

    The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a postulated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report describes the analytical results on the system behavior observed in the SCTF Core-I cold leg injection tests, S1-14 (Run 520), S1-15 (521), S1-16 (522), S1-17 (523), S1-20 (530), S1-21 (531), S1-23 (536) and S1-24 (537), performed under constant system pressure condition during transient. Major discussion items are: (1) steam binding, (2) U-tube oscillations, (3) bypass of ECC water (4) core cooling behavior, (5) effect of vent valve and (6) other parameter effects. These results give us very useful information and suggestion on reflood behavior. (author)

  2. Structural response of a rotating bladed disk to rotor whirl

    Science.gov (United States)

    Crawley, E. F.

    1985-01-01

    A set of high speed rotating whirl experiments were performed in the vacuum of the MIT Blowdown Compressor Facility on the MIT Aeroelastic Rotor, which is structurally typical of a modern high bypass ratio turbofan stage. These tests identified the natural frequencies of whirl of the rotor system by forcing its response using an electromagnetic shaker whirl excitation system. The excitation was slowly swept in frequency at constant amplitude for several constant rotor speeds in both a forward and backward whirl direction. The natural frequencies of whirl determined by these experiments were compared to those predicted by an analytical 6 DOF model of a flexible blade-rigid disk-flexible shaft rotor. The model is also presented in terms of nondimensional parameters in order to assess the importance of the interation between the bladed disk dynamics and the shaft-disk dynamics. The correlation between the experimental and predicted natural frequencies is reasonable, given the uncertainty involved in determining the stiffness parameters of the system.

  3. 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.

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

    International Nuclear Information System (INIS)

    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. 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.

  6. 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)

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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

  9. Uncertainties in modeling and scaling in the prediction of fuel stored energy and thermal response

    International Nuclear Information System (INIS)

    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

  10. SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR

    International Nuclear Information System (INIS)

    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)

  11. Simulation of noncondensable gases in SAGD steam chambers

    Energy Technology Data Exchange (ETDEWEB)

    Gittins, Simon; Gupta, Subodh; Zaman, Maliha [Cenovus Energy (Canada)

    2011-07-01

    Cenovus Energy has been successfully using the steam assisted gravity drainage (SAGD) process at various sites. As these and other wells mature, a greater understanding of non-condensable gasses is required to help to optimize other factors such as methane co-injection and the steam ramp-down and blow-down phases. It is very important to understand fully how non-condensable gasses operate in SAGD chambers in order to lower energy intensity, costs, and the environmental impact while increasing the yield from the reserves. Cenovus Energy also plans on reducing pressure in SAGD and solvent-aided processes in future projects by applying their acquired knowledge of non-condensable gasses. The paper shows results from recent simulations that improve understanding of this subject. Simulation has shown that if there are significant flow restrictions in SAGD injection wells, that would cause the steam to flow at a higher pressure axially along the steam chamber as opposed to axially along the liner and out. This accounts for the production of solution gas.

  12. Prediction of scales in boilers of thermal recovery projects

    Energy Technology Data Exchange (ETDEWEB)

    Thimm, H.F. [Thimm Engineering Inc., Calgary, AB (Canada); Kwasniewski, K. [EnCana Corp., Calgary, AB (Canada)

    2007-07-01

    Significant efforts at controlling silica in thermal petroleum recovery projects are a regular aspect of facilities engineering in such projects. However, there is more interest in iron, calcium, magnesium and sodium, in solutions of high pH, such as boiler feedwater and blowdown. Computer programs that rely on free energy minimizations enable the prediction of scaling. A large range of possible mineral deposits are frequently identified as potential scale deposits where instabilities for scaling are predicted in this manner. However, only one or two such minerals are ever found in the analysis of pigging solids. This paper presented the results of a study that derived a simple method, that permits the prediction by non-chemists of both type and quantity of preferential scales, and illustrated its use in steam assisted gravity drainage (SAGD) water management and recycling schemes. The study utilized the SOLMINEQ program developed by the Alberta Research Council. It was concluded that the effect of the presence of chelants in boiler feedwater may not prevent silicate scales, but simply shift the preferred scale. On the other hand, sample recovery and handling, may create a shift in scale preference under laboratory conditions, as opposed to facility conditions. 3 refs.

  13. 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.

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

    International Nuclear Information System (INIS)

    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

  15. A modelling study of the multiphase leakage flow from pressurised CO2 pipeline.

    Science.gov (United States)

    Zhou, Xuejin; Li, Kang; Tu, Ran; Yi, Jianxin; Xie, Qiyuan; Jiang, Xi

    2016-04-01

    The accidental leakage is one of the main risks during the pipeline transportation of high pressure CO2. The decompression process of high pressure CO2 involves complex phase transition and large variations of the pressure and temperature fields. A mathematical method based on the homogeneous equilibrium mixture assumption is presented for simulating the leakage flow through a nozzle in a pressurised CO2 pipeline. The decompression process is represented by two sub-models: the flow in the pipe is represented by the blowdown model, while the leakage flow through the nozzle is calculated with the capillary tube assumption. In the simulation, two kinds of real gas equations of state were employed in this model instead of the ideal gas equation of state. Moreover, results of the flow through the nozzle and measurement data obtained from laboratory experiments of pressurised CO2 pipeline leakage were compared for the purpose of validation. The thermodynamic processes of the fluid both in the pipeline and the nozzle were described and analysed.

  16. Evaluation of the effects of break nozzle configuration in the Semiscale Mod-1 system

    International Nuclear Information System (INIS)

    The Semiscale Mod-1 Program has utilized two different break nozzle configurations in the test system. An evaluation has been made to determine the effect these break nozzle configurations have on system thermal-hydraulic response during a 200 percent double-ended cold leg break loss-of-coolant accident simulation. The first nozzle was a convergent-divergent nozzle (Henry nozzle) and the second, an elongated constant area throat nozzle. Analysis is confined primarily to system response phenomena observed to be affected by the nozzle configuration and concentrates on the fluid response at the break and the resulting core behavior during subcooled and saturated blowdown. The evaluation shows that considerable difference in system response occurs as a result of the difference in break nozzle configuration. The elongated throat nozzle was scaled from the Loss-of-Fluid Test (LOFT) nozzle geometry and since the LOFT counterpart tests were designed to provide results for the LOFT Program, the elongated throat nozzle was used in the subsequent LOFT counterpart tests

  17. 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.

  18. 降低焦化废水总氰浓度的方法%Methods of reducing total cyanide concentration in coke plant wastewater

    Institute of Scientific and Technical Information of China (English)

    孙红艳; 刘武镛

    2015-01-01

    This paper introduces the three methods of reducing total cyanide concentration in coke plant wastewater. The test shows that the methods of coke absorption and flocculant dosing to oil re-moval tank are both effective. Coke absorption method will increase labor intensity,so it is not a good option;while the method of dosing flocculant to oil removal tank with self-pressure blowdown system could not only reduce the total cyanide concentration and also benefit the biochemical treatment process.%介绍了降低焦化废水总氰浓度的3种方法。试验表明,采用焦炭吸附和在除油池投加絮凝剂的方法都是有效的。焦炭吸附法会增加劳动强度,不宜采用;在有自压式排污系统的除油池内投加絮凝剂,不但降低了总氰浓度,而且有利于生化处理工序。

  19. Amazon old-growth forest wind disturbance and the regional carbon balance

    Science.gov (United States)

    Chambers, J. Q.; Negron Juarez, R. I.; Marra, D. M.; Roberts, D. A.; Hurtt, G. C.; Lima, A.; Higuchi, N.

    2010-12-01

    Estimating the carbon balance of a landscape is challenging. A key problem is determining whether or not measurements made in plots are representative of the carbon state of a larger region. A key parameter for calculating landscape carbon balance is the return frequency of episodic disturbances. If disturbances are clustered and occur more frequently than the time required for biomass recovery, a spatial mixture of patches in different stages of recovery occurs. Under these shifting steady-state mosaic conditions, quantifying the mean state of ecosystem attributes such as carbon balance or tree species diversity is difficult. In this study, satellite remote sensing (Landsat) was coupled with field investigations to create ~25 year landscape-scale disturbance chronosequence for old-growth forest in the Central Amazon. The detected disturbances were caused by strong storms which resulted in tree mortality events ranging from small clusters of 7-10 downed trees, to large contiguous blowdowns larger than 30 ha in size. Using the chronosequence, a cumulative probability distribution function was developed, which followed a power law, and was used to parameterize a forest carbon balance model. Results demonstrate that for power law exponents less than about 2.0, the spatial scale at which forest carbon balance establishes is much larger than generally expected. Ultimately, an increase in wind disturbance frequency and/or intensity with a warming climate has the potential to cause a net loss of carbon from Amazon forests to the atmosphere.

  20. Widespread Amazon forest tree mortality from a single cross-basin squall line event

    Science.gov (United States)

    Negrón-Juárez, Robinson I.; Chambers, Jeffrey Q.; Guimaraes, Giuliano; Zeng, Hongcheng; Raupp, Carlos F. M.; Marra, Daniel M.; Ribeiro, Gabriel H. P. M.; Saatchi, Sassan S.; Nelson, Bruce W.; Higuchi, Niro

    2010-08-01

    Climate change is expected to increase the intensity of extreme precipitation events in Amazonia that in turn might produce more forest blowdowns associated with convective storms. Yet quantitative tree mortality associated with convective storms has never been reported across Amazonia, representing an important additional source of carbon to the atmosphere. Here we demonstrate that a single squall line (aligned cluster of convective storm cells) propagating across Amazonia in January, 2005, caused widespread forest tree mortality and may have contributed to the elevated mortality observed that year. Forest plot data demonstrated that the same year represented the second highest mortality rate over a 15-year annual monitoring interval. Over the Manaus region, disturbed forest patches generated by the squall followed a power-law distribution (scaling exponent α = 1.48) and produced a mortality of 0.3-0.5 million trees, equivalent to 30% of the observed annual deforestation reported in 2005 over the same area. Basin-wide, potential tree mortality from this one event was estimated at 542 ± 121 million trees, equivalent to 23% of the mean annual biomass accumulation estimated for these forests. Our results highlight the vulnerability of Amazon trees to wind-driven mortality associated with convective storms. Storm intensity is expected to increase with a warming climate, which would result in additional tree mortality and carbon release to the atmosphere, with the potential to further warm the climate system.

  1. Assessment study of RELAP5/MOD2 Cycle 36. 04 based on pressurizer safety and relief valve tests

    Energy Technology Data Exchange (ETDEWEB)

    Stubbe, E.J.; Vanhoenacker, L.

    1990-07-01

    This report presents a code assessment study based on full size relief and assisted safety valve (called SEBIM) tests performed on the CUMULUS valve test rig operated by Electricite de France (EdF). The increased awareness that the pressuriser safety and relief valves are not reliable under water blowdown conditions, has led to the design, testing and installation of so called assisted safety valves of which the SEBIM (TM) valves are an example. These valves, used in tandem, are gradually replacing the safety and relief valves on pressurisers in some European PWR's. Before installation at the plant, the Belgian safety authorities requested a thorough full scale testing of these valves on a test rig (CUMULUS) equipped with sufficient diagnostics to measure the characteristics of the valve. The Belgian architect-engineering firm TRACTEBEL was called upon the specify, order and test these valves for installation at the DOEL 1 and DOEL 2 power plants. These tests do provide sufficient data of high quality to justify an assessment study of the code RELAP-5 MOD-2 CYCLE 36 in the ICAP framework which is the subject of this report.

  2. ROSA-II test data report, 11

    International Nuclear Information System (INIS)

    Results of the ROSA-II tests simulating a loss-of-coolant accident (LOCA) and effects of an emergency core cooling system (ECCS) in a pressurized water reactor (PWR) are presented including the test conditions and interpretations of the data in test runs 327,328,329 and 330. Each test was performed with large double-ended hot leg break and effect of the break area distribution (break diameter are 25.0 mm at one end and 37.5 mm at the other end of break) and of pump circulation upon coolant flow in the core were studied. The following are the results: In the case of a smaller break on the steam generator side, core cooling was achieved due to upward coolant flow in the core and early reflooding by ACC water injected into the cold leg. In the case of a smaller break area on the vessel side, on the other hand, coolant flow in the core was stagnant and the heater rods were mostly exposed to steam, so that core cooling was not as good. Effect of the coolant circulation by acting pump on the core cooling during a blowdown was not significant except that in a steam generator side small break the core cooling was improved. (auth.)

  3. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    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{sup 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)

  4. Pilot RCM application to the Diablo Canyon main stream system

    International Nuclear Information System (INIS)

    In 1986 Pacific Gas ampersand Electric Company (PG ampersand E) became extremely interested in reliability-centered maintenance (RCM) after the initial review of two successful Electric Power Research Institute sponsored projects. RCM was visualized as a methodology to common sensitize the burgeoning preventive maintenance (PM) program at the Diablo Canyon plant. RCM could further the uses of predictive and condition-monitoring techniques, as well as eliminate maintenance on components whose failures were noncritical. An extensive review of maintenance and operation experience data, in conjunction with plant staff recommendations and a prioritization according to maintenance expenditures and operational/safety significance, produced the selected system: the turbine main steam supply system (main steam). The pilot project segmented the main steam system into eight subsystems to aid in analysis: (a) main steam isolation valves, (b) auxiliary feedwater pump turbine, (c) overpressure protection (steam dump), (d) main feedwater pump turbines, (e) main steam, (f) main turbine, (g) steam blowdown, and (h) moisture separator reheaters. System analysis activities, including the preparation of functional failure analyses, failure modes and effects analyses, and logic model analyses, were conducted in parallel with corrective and preventive maintenance data-gathering activities to maximize project team personnel participation during the project. Results and lessons learned are summarized

  5. Transient/structural analysis of a combustor under explosive loads

    Science.gov (United States)

    Gregory, Peyton B.; Holland, Anne D.

    1992-01-01

    The 8-Foot High Temperature Tunnel (HTT) at NASA Langley Research Center is a combustion-driven blow-down wind tunnel. A major potential failure mode that was considered during the combustor redesign was the possibility of a deflagration and/or detonation in the combustor. If a main burner flame-out were to occur, then unburned fuel gases could accumulate and, if reignited, an explosion could occur. An analysis has been performed to determine the safe operating limits of the combustor under transient explosive loads. The failure criteria was defined and the failure mechanisms were determined for both peak pressures and differential pressure loadings. An overview of the gas dynamics analysis was given. A finite element model was constructed to evaluate 13 transient load cases. The sensitivity of the structure to the frequency content of the transient loading was assessed. In addition, two closed form dynamic analyses were conducted to verify the finite element analysis. It was determined that the differential pressure load or thrust load was the critical load mechanism and that the nozzle is the weak link in the combustor system.

  6. Investigation of the performance of a variable area diffuser for gas dynamic lasers

    Energy Technology Data Exchange (ETDEWEB)

    Nuttbrock, D.L.

    1974-06-01

    An experimental study was performed to determine the performance of a variable area diffuser downstrem of an array of supersonic nozzles, and to determine the Mach number profile between the nozzle exit and the diffuser entrance. The study was conducted on a blowdown wind tunnel and the test section was designed to model a gas dynamic laser with an array of five nozzle blades, a constant area section, and a converging-diverging diffuser. Air at a temperature of 70/sup 0/F and at total pressures ranging from 100 to 210 psig was expanded through an area ratio of approximately 66. Using various pressure measurements the Mach number was found to decrease from M = 6.4 at the nozzle exit to approximately M = 4.0 at the diffuser entrance. The rapid decrease was attributed to the irreversible effects of friction, the nozzle blade wakes, and the nozzle throat shocks. The minimum starting area ratio of the diffuser was 0.59, which agrees well with one dimensional theory.

  7. 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)

  8. a Simplified Methodology for the Prediction of the Small Break Loss-Of Accident.

    Science.gov (United States)

    Ward, Leonard William

    1988-12-01

    This thesis describes a complete methodology which has allowed for the development of a faster than real time computer program designed to simulate a small break loss -of-coolant accident in the primary system of a pressurized water reactor. By developing an understanding of the major phenomenon governing the small break LOCA fluid response, the system model representation can be greatly simplified leading to a very fast executing transient system blowdown code. Because of the fast execution times, the CULSETS code, or Columbia University Loss-of-Coolant Accident and System Excursion Transient Simulator code, is ideal for performing parametric studies of Emergency Core Cooling system or assessing the consequences of the many operator actions performed to place the system in a long term cooling mode following a small break LOCA. While the methodology was designed with specific application to the small break loss-of-coolant accident, it can also be used to simulate loss-of-feedwater, steam line breaks, and steam generator tube rupture events. The code is easily adaptable to a personal computer and could also be modified to provide the primary and secondary system responses to supply the required inputs to a simulator for a pressurized water reactor.

  9. Multiple disturbances accelerate clonal growth in a potentially monodominant bamboo.

    Science.gov (United States)

    Gagnon, Paul R; Platt, William J

    2008-03-01

    Organisms capable of rapid clonal growth sometimes monopolize newly freed space and resources. We hypothesize that sequential disturbances might change short-term clonal demography of these organisms in ways that promote formation of monotypic stands. We examined this hypothesis by studying the clonal response of Arundinaria gigantea (giant cane, a bamboo) to windstorm and fire. We studied giant cane growing in both a large tornado-blowdown gap and under forest canopy, in burned and unburned plots, using a split-block design. We measured density of giant cane ramets (culms) and calculated finite rates of increase (lamda) for populations of ramets over three years. Ramet density nearly doubled in stands subjected to both windstorm and fire; the high ramet densities that resulted could inhibit growth in other plants. In comparison, ramet density increased more slowly after windstorm alone, decreased after fire alone, and remained in stasis in controls. We predict that small, sparse stands of giant cane could spread and amalgamate to form dense, monotypic stands (called "canebrakes") that might influence fire return intervals and act as an alternative state to bottomland forest. Other clonal species may similarly form monotypic stands following successive disturbances via rapid clonal growth. PMID:18459325

  10. Hydrogen Mixing Studies (HMS) assessment manual

    International Nuclear Information System (INIS)

    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 m3 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

  11. 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.

  12. A FORTRAN program for calculating three dimensional, inviscid and rotational flows with shock waves in axial compressor blade rows: User's manual

    Science.gov (United States)

    Thompkins, W. T., Jr.

    1982-01-01

    A FORTRAN-IV computer program was developed for the calculation of the inviscid transonic/supersonic flow field in a fully three dimensional blade passage of an axial compressor rotor or stator. Rotors may have dampers (part span shrouds). MacCormack's explicit time marching method is used to solve the unsteady Euler equations on a finite difference mesh. This technique captures shocks and smears them over several grid points. Input quantities are blade row geometry, operating conditions and thermodynamic quanities. Output quantities are three velocity components, density and internal energy at each mesh point. Other flow quanities are calculated from these variables. A short graphics package is included with the code, and may be used to display the finite difference grid, blade geometry and static pressure contour plots on blade to blade calculation surfaces or blade suction and pressure surfaces. The flow in a low aspect ratio transonic compressor was analyzed and compared with high response total pressure probe measurements and gas fluorescence static density measurements made in the MIT blowdown wind tunnel. These comparisons show that the computed flow fields accurately model the measured shock wave locations and overall aerodynamic performance.

  13. Periodical shedding of cloud cavitation from a single hydrofoil in high-speed cryogenic channel flow

    Institute of Scientific and Technical Information of China (English)

    Yutaka ITO; Koichi SETO; Takao NAGASAKI

    2009-01-01

    In order to explain criteria for periodical shedding of the cloud cavitation, flow patterns of cavitation around a piano-convex hydrofoil were observed using a cryogenic cavitation tunnel of a blowdown type. Two hydrofoils of similarity of 20 and 60 mm in chord length with two test sections of 20 and 60 mm in width were prepared. Working fluids were water at ambient temperature, hot water and liquid nitrogen. The parameter range was varied between 0.3 and 1.4 for cavitation number, 9 and 17 m/sec for inlet flow velocity, and -8° and 8° for the flow in-cidence angle, respectively. At incidence angle 8°, that is, the convex surface being suction surface, periodical shedding of the whole cloud cavitation was observed on the convex surface under the specific condition with cavitation number and inlet flow velocity, respectively, 0.5, 9 m/sec for liquid nitrogen at 192℃ and 1.4, 11 m/sec for water at 88℃, whereas under the supercavitation condition, it was not observable. Periodical shedding of cloud cavitation occurs only in the case that there are both the adverse pressure gradient and the slow flow region on the hydrofoil.

  14. Characterization and treatment options of solid residues from waste to energy plants

    International Nuclear Information System (INIS)

    Solid residues from waste to energy plants represent important byproducts of the thermal treatment process, with significant implications in all the procedures involved in the selection of alternative technological process options, in the achievement of the consensus of residents in the area and in decisions related to plant siting. Most recent restrictions broadly applied in the field of atmospheric emission limits have further increase their relative contribution to the environmental burden of the plant as a whole, particularly for certain toxic trace elements of interest removed with very high efficiencies from flue gas, most frequently through simple transfer rather than conversion and thus significantly enriched in the final residues of the removal process. Following a broad introduction on the main qualitative and quantitative characteristics of all the residues typically arising from waste to energy plants (furnace slag, flyash from particulate removal, ash from dry and semidry flue gas control operations, sludge from wet scrubbers blowdown treatment), the paper reports on the main technologies for their treatment and final disposal actually adopted in full scale applications, as well as on the alternatives that might be prospected in the near future for achieving further reductions in the total release of contaminants from the plant as a whole, in accordance with most recently proposed regulation strategies for industrial activities based on the IPPC approach (Integrated Pollution Prevention and Control)

  15. Blind Calculation of RD-14M Small Break LOCA Tests by CATHENA Code

    International Nuclear Information System (INIS)

    KAERI participated with the computer code CATHENA, which is used to analyze Pressurized Heavy Water Reactors (PHWRs), in an IAEA International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate thermal-hydraulic computer code against qualified data for Small Break Loss of Coolant Accident (SBLOCA) scenario generated on RD-14M Test Facility. Two specific SBLOCA tests selected for this ICSP titled 'Comparison of HWR Code Predictions with SBLOCA Experimental Data', are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow down. This report presents the blind calculation results for these tests conducted by CATHENA code before the test data are distributed to participants. For B9006 test, CATHENA code simulated all the phases of the transient such as blowdown, high-pressure ECI, secondary pressure ramp, refill, switch from high pressure ECI to low pressure ECI, exponential pump ramp, and natural circulation. For B9802 test, CATHENA calculation was intended to predict temperature rise of the FES sheath due to channel boiling, and power supply trip on high FES sheath temperature (600 .deg. C) process protection trip

  16. Containment vessel, its auxiliary system and plant air conditioning system of advanced thermal reactor Fugen

    International Nuclear Information System (INIS)

    The functional requirement for, the design and the construction of, and the functional test on the containment vessel, its auxiliary system, the plant air conditioning and ventilation system of the advanced thermal reactor, Fugen, are described in detail. The main specifications of the containment vessel are as follows: The type enclosed cylinder, the maximum operating pressure 1.35 kg/cm2g, the maximum operating temperature 100 deg C, the leak rate 0.4%/day, the inner diameter 36 m. The height 64 m, the volume 40,900 m3, and the material JIS G3118, SGV-49. The containment vessel is provided with an hatch of 5 m diameter for carrying equipments in two air locks, many high and low voltage cable penetrations, pipe penetrations, a transfer shoot and isolation values. The functions and the specifications of the containment vessel and its auxiliary equipments are explained. The relating auxiliary systems are composed of the containment vessel spray system, the pool facility for steam blow-down, the recirculation system for the air in the vessel, the annulus evacuation system and its pressure control devices, the pressure measuring instruments and pressure relief valves and the temperature measuring devices for the containment vessel, and the object, function, layout and installation of these systems are explained. Concerning the air conditioning system, each main building has the special subsystem, and they are introduced. The progress stage of construction works and the procedure and results of the functional test at the site are described. (Nakai, Y.)

  17. 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)

  18. 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

  19. 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

  20. The January 21, 1951 Blast of Mount Lamington in Papua New Guinea: Sequence of Events and Characteristics of the Deposits

    Science.gov (United States)

    Belousova, M.; Belousov, A.; Patia, H.; Hoblitt, R. P.

    2011-12-01

    We present the results of a detailed reinvestigation of deposits of the famous 1951 eruption of Mount Lamington which was originally studied by T. Taylor (1958). We found that the climactic phase of the eruption was triggered by a relatively small gravitational collapse of the old intracrater lava dome (debris avalanche V=0.02-0.04 cub. km; L=8.5 km; H/L=0.14). The collapse was followed by vertical explosive fountain which was not buoyant and formed a pyroclastic density current (PDC). This PDC completely devastated an area of 230 sq. km, traveling maximum distance of 15 km in N direction; 3500 people were killed by the eruption. The PDC deposit, which is still well-preserved, was studied in 2 profiles, which are parallel to the longest axis of the surge propagation. The deposit consists of mostly juvenile rock fragments (80-85%) represented by poorly vesicular (4 - 40%) highly crystalline dacite; bombs with poorly developed bread crust surfaces are common in proximal areas. The deposit is in general normally graded and consists of lapilli and coarse ash fining upward into fine ash. The base of the deposit is mixed with soil in proximal areas. Stratigraphic characteristics of the deposit demonstrate strong local fluctuations, but have clear trends with distance from the volcano. At distances from 3 to 12 km from the volcano the maximum deposit thickness decreases from 55 to 5 cm, and the average size of the 10 largest clasts decreases from 4.5 cm to 0.5 cm; Md diameter decreases from -1.5 to 4.5 phi; sorting improves from 3 to 0.7 phi. The surge produced spectacular tree blow-down in the devastated area. Aerial photographs taken one month after eruption show that the PDC was strongly channelized even by small (tens meters) topographic features; the front of the moving PDC was frequently split into multiple small tongues which were variously deflected by topography. The deposit and the tree blow-down features demonstrate many similarities with those of blast

  1. Ethanolamine experience at Koeberg nuclear power station, South Africa

    International Nuclear Information System (INIS)

    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)

  2. 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)

    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

  3. Fulton Generating Station Units 1 and 2 (Docket Nos. 50-463 and 50-464): Final environmental statement

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. Component Testing of the J-2X Augmented Spark Igniter (ASI)

    Science.gov (United States)

    Osborne, Robin J.; Peters, Warren T.; Gaspar, Kenny C.; Hauger, Katherine; Kwapisz, Mike J.

    2013-01-01

    In support of the development of the J-2X engine, 201 low pressure, liquid oxygen / liquid hydrogen (LOX/LH2) J-2X Augmented Spark Igniter (ASI) subsystem ignition tests were conducted at Marshall Space Flight Center (MSFC). The main objective of these tests was to start the ASI within the anticipated J-2X engine start box, as well as outside of it, to check for ignition margin. The setup for the J-2X ASI component testing simulated, as much as possible, the tank-head start-up configuration of the ASI within the J-2X Engine. The ignition tests were divided into 124 vacuum start tests to simulate altitude start on a flight engine, and 77 sea-level start tests to simulate the first set of ground tests for the J-2X Engine at Stennis Space Center (SSC). Other ignition parameters that were varied included propellant tank pressures, oxidizer temperature entering the ASI oxidizer feedline, oxidizer valve timing, spark igniter condition (new versus damaged), and oxidizer and fuel feedline orifice sizes. Propellant blowdowns using venturis sized to simulate the ASI resistance allowed calculation of transient propellant mass flow rates as well as global mixture ratio for all ignition tests. Global mixture ratio within the ASI at the time of ignition varied from 0.2 to 1.2. Detailed electronics data obtained from an instrumented ignition lead allowed characterization of the breakdown voltage, sustaining voltage and energy contained in each spark as the ASI propellants ignited. Results indicated that ignition always occurred within the first five sparks when both propellants were present in the ASI chamber.

  6. Biphase turbine bottoming cycle for a diesel engine

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, S.; Hays, L.

    1977-02-15

    Application of a two-phase turbine system to waste heat recovery was examined. Bottoming cycle efficiencies ranging from 15 to 30% were calculated for a 720/sup 0/F diesel exhaust temperature. A single stage demonstration unit, designed for non-toxic fluids (water and DowTherm A) and for atmospheric seals and bearings, had a cycle efficiency of 23%. The net output power was 276 hp at 8,100 rpm, increasing the total shaft power from 1,800 hp for the diesel alone, to 2,076 hp for the combined system. A four stage organic turbine, for the same application, had a rotational speed of 14,700 rpm while a four stage steam turbine had 26,000 rpm. Fabrication drawings were prepared for the turbine and nozzle. The major improvement leading to higher cycle efficiency and lower turbine rpm was found to be the use of a liquid component with lower sensible heat. A reduction in capital cost was found to result from the use of a contact heat exchanger instead of tube-fin construction. The cost for a contact heat exchanger was only $35-52/kWe compared to $98/kWe for a tube-fin heat exchanger. Design drawings and materials list were prepared. A program resulting in the demonstration of a two-phase bottoming system was planned and the required cost estimated. The program would result in a feasibility test of the nozzle and turbine at the end of the first year, a laboratory performance test of the bottoming system by the end of the second year and a field demonstration test and laboratory endurance test of the bottoming system during the third year. The blowdown test rig for the first year's program and test turbine were designed.

  7. Forward osmosis applied to evaporative cooling make-up water

    Energy Technology Data Exchange (ETDEWEB)

    Nicoll, Peter; Thompson, Neil; Gray, Victoria [Modern Water plc, Guildford (United Kingdom)

    2012-11-15

    Modern Water is in the process of developing a number of forward osmosis based technologies, ranging from desalination to power generation. This paper outlines the progress made to date on the development and commercial deployment of a forward osmosis based process for the production of evaporative cooling tower make-up water from impaired water sources, including seawater. Evaporative cooling requires significant amounts of good quality water to replace the water lost by evaporation, drift and blowdown. This water can be provided by conventional desalination processes or by the use of tertiary treated sewage effluent. The conventional processes are well documented and understood in terms of operation and power consumption. A new process has been successfully developed and demonstrated that provides make-up water directly, using a core platform 'forward osmosis' technology. This new technology shows significant promise in allowing various raw water sources, such as seawater, to be used directly in the forward osmosis step, thus releasing the use of scarce and valuable high grade water for other more important uses. The paper presents theoretical and operational results for the process, where it is shown that the process can produce make-up water at a fraction of the operational expenditure when compared to conventional processes, in particular regarding power consumption, which in some cases may be as low as 15 % compared to competing processes. Chemical additives to the cooling water (osmotic agent) are retained within the process, thus reducing their overall consumption. Furthermore the chemistry of the cooling water does not support the growth of Legionella pneumophila. Corrosion results are also reported. (orig.)

  8. Dynamic fracture analysis of HDR-E31 pipe system experiments

    International Nuclear Information System (INIS)

    This paper presents analyses of circumferentially surface-cracked pipe system experiments that were conducted under blowdown (water-hammer) loading as part of the German HDR program. Two analysis procedures are compared to the experimental results: (1) typical engineering pipe fracture analyses using the uncracked pipe elastic stress analysis, and (2) nonlinear cracked-pipe time-history analyses. The nonlinear time-history analysis uses a nonlinear spring element at the crack location to simulate the rotations due to the crack at the crack section, with the rest of the pipe system modeled using typical pipe and elbow beam elements. Such a FEM approach was very economical to run compared with three-dimensional modeling of a surface crack in a pipe system. The results show that using the elastic uncracked pipe dynamic analysis, all the pipe fracture analyses predict the pipe will reach the maximum load. Because such calculations consider the dynamic stresses to be statically applied, these analyses would all predict a double-ended guillotine break (DEGB). The most conservative analyses was the ASME Section 11 IWB-3650 analysis for ferritic pipe (without the Code applied safety factors) which predicted failure loads less than half of the other analysis methods. The nonlinear cracked-pipe element time-history FEM analysis predictions of surface-crack penetration were much closer to the experiments but still slightly conservative. Using this method, the extent of the crack growth after the surface crack penetration is also predicted. The nonlinear time-history calculations predicted that a DEGB would not occur, consistent with experimental behavior. The extent of the predicted through-wall crack penetration was slightly conservative compared to the experimental results

  9. 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.

  10. Modeling of conjugate two-phase heat transfer during depressurization of pipelines

    International Nuclear Information System (INIS)

    A growing number of multiphase technology applications in the petroleum, chemical, aerospace, power, and process industries stimulate the development of reliable methods for analyzing transient processes in two-phase systems in which the temperature field in the moving fluid and the temperature field in the bounding walls are directly dependent on each other. Examples of such types of processes include: controlled or accidental blowdowns of subsea oil/gas pipelines, the loss of coolant accident in a nuclear power plant, and release of highly volatile multicomponent liquids during accidents in chemical plants. Here, transient conjugate two-phase heat transfer during depressurization of pipelines containing flashing liquids is examined in this paper. A numerical model for transient flashing liquid flow in a pipe is formulated. The model takes into account the transient radial heat conduction and the forced convection effects. Numerical simulation of flashing two-component (propane and butane) flow is performed in order to investigate the effect of wall friction on the heat transfer conditions in the pipe. The simulation results are compared with predictions of the model that are based on a new formulation of energy equation proposed by the author in an early study. A comparison of the results obtained allows one to determine the range of applicability of the new energy equation formulation. A procedure is proposed for choosing an appropriate model for predicting transient conjugate two-phase heat transfer during releases of flashing liquids from pipes. A criterion of thermodynamic similarity for flashing liquids flows in pipes or channels is formulated. The proposed criterion provides the basis for selecting model fluids and for constructing experimental models of systems containing flashing (volatile) liquids with scaled thermodynamic conditions. An example of its use is given

  11. Testing HEPA filter response to high flow velocity and overpressure

    International Nuclear Information System (INIS)

    In cooperation with the Los Alamos National Laboratory (LANL) HEPA filters presently on the market and modified versions were tested using the LANL blowdown facility. The filter response during the test was documented on high speed films which indicated, among other information, the load when first failure occurred and the failure mode. Differential pressure through the filter and velocity were measured with variable reluctance pressure transducers and an oscilloscope recorder. New HEPA filters show first failures already at differential pressures between some 4 kPa and 17 kPa. The better results are obtained with conventionally pleated filter paper packs sealed in a wooden frame with a polyurethane material. However, these filters cannot be used under accident conditions. Metal frame filters considered as potential filters for accident situations showed only poor resistance against overpressures. The weakest point in such filters proved to be the attachment of the filter pack in the metal frame, which limited the loadability to about 10 kPa. As soon as proper attachment is assured, the integrity of the filter pack limits the admissible differential pressure. Through some modifications of wood frame filters, physical integrity was improved to a failure pressure of 24 kPa. During operation, the filter elements are loaded with dust. Dust loading up to 1000 Pa pressure drop at rated flow, which was simulated with polystyrene latex aerosols, was found to reduce the mechanical stability by up to 40%. Pressure drop characteristics of HEPA filters influence the occurrence of high differential pressures and high mechanical loads. Pressure drops (δp) of unloaded and preloaded filters were investigated up to high flow velocities (v). Measured data can be presented in the form of δp=av+bv2, where a and b are constants. Theory, in contrast, predicts a linear relationship. The deviation is due to deformations of the pleated filter paper pack with increasing flow velocity and

  12. International Space Station (ISS) Gas Logistics Planning in the Post Shuttle Era

    Science.gov (United States)

    Leonard, Daniel J.; Cook, Anthony J.; Lehman, Daniel A.

    2011-01-01

    Over its life the International Space Station (ISS) has received gas (nitrogen, oxygen, and air) from various sources. Nitrogen and oxygen are used in the cabin to maintain total pressure and oxygen partial pressures within the cabin. Plumbed nitrogen is also required to support on-board experiments and medical equipment. Additionally, plumbed oxygen is required to support medical equipment as well as emergency masks and most importantly EVA support. Gas are supplied to ISS with various methods and vehicles. Vehicles like the Progress and ATV deliver nitrogen (both as a pure gas and as air) and oxygen via direct releases into the cabin. An additional source of nitrogen and oxygen is via tanks on the ISS Airlock. The Airlock nitrogen and oxygen tanks can deliver to various users via pressurized systems that run throughout the ISS except for the Russian segment. Metabolic oxygen is mainly supplied via cabin release from the Elektron and Oxygen Generator Assembly (OGA), which are water electrolyzers. As a backup system, oxygen candles (Solid Fuel Oxygen Generators-SFOGs) supply oxygen to the cabin as well. In the past, a major source of nitrogen and oxygen has come from the Shuttle via both direct delivery to the cabin as well as to recharge the ISS Airlock tanks. To replace the Shuttle capability to recharge the ISS Airlock tanks, a new system was developed called Nitrogen/Oxygen Recharge System (NORS). NIORS consists of high pressure (7000 psi) tanks which recharge the ISS Airlock tanks via a blowdown fill for both nitrogen and oxygen. NORS tanks can be brought up on most logistics vehicles such as the HTV, COTS, and ATV. A proper balance must be maintained to insure sufficient gas resources are available on-orbit so that all users have the required gases via the proper delivery method (cabin and/or plumbed).

  13. Propulsion engineering study for small-scale Mars missions

    Energy Technology Data Exchange (ETDEWEB)

    Whitehead, J.

    1995-09-12

    Rocket propulsion options for small-scale Mars missions are presented and compared, particularly for the terminal landing maneuver and for sample return. Mars landing has a low propulsive {Delta}v requirement on a {approximately}1-minute time scale, but at a high acceleration. High thrust/weight liquid rocket technologies, or advanced pulse-capable solids, developed during the past decade for missile defense, are therefore more appropriate for small Mars landers than are conventional space propulsion technologies. The advanced liquid systems are characterize by compact lightweight thrusters having high chamber pressures and short lifetimes. Blowdown or regulated pressure-fed operation can satisfy the Mars landing requirement, but hardware mass can be reduced by using pumps. Aggressive terminal landing propulsion designs can enable post-landing hop maneuvers for some surface mobility. The Mars sample return mission requires a small high performance launcher having either solid motors or miniature pump-fed engines. Terminal propulsion for 100 kg Mars landers is within the realm of flight-proven thruster designs, but custom tankage is desirable. Landers on a 10 kg scale also are feasible, using technology that has been demonstrated but not previously flown in space. The number of sources and the selection of components are extremely limited on this smallest scale, so some customized hardware is required. A key characteristic of kilogram-scale propulsion is that gas jets are much lighter than liquid thrusters for reaction control. The mass and volume of tanks for inert gas can be eliminated by systems which generate gas as needed from a liquid or a solid, but these have virtually no space flight history. Mars return propulsion is a major engineering challenge; earth launch is the only previously-solved propulsion problem requiring similar or greater performance.

  14. Hydrodynamics of AHWR gravity driven water pool under simulated LOCA conditions

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) employs a double containment concept with a large inventory of water within the Gravity Driven Water Pool (GDWP) located at a high elevation within the primary containment building. GDWP performs several important safety functions in a passive manner, and hence it is essential to understand the hydrodynamics that this pool will be subjected to in case of an accident such as LOCA. In this paper, a detailed thermal hydraulic analysis for AHWR containment transients is presented for postulated LOCA scenarios involving RIH break sizes ranging from 2% to 50%. The analysis is carried out using in-house containment thermal hydraulics code 'CONTRAN'. The blowdown mass and energy discharge data for each break size, along with the geometrical details of the AHWR containment forms the main input for the analysis. Apart from obtaining the pressure and temperature transients within the containment building, the focus of this work is on simulating the hydrodynamic phenomena of vent clearing and pool swell occurring in the GDWP. The variation of several key parameters such as primary containment V1 and V2 volume pressure, temperature and V1-V2 differential pressure with time, BOP rupture time, vent clearing velocity, effect of pool swell on the V2 air-space pressure, GDWP water level etc. are discussed in detail and important findings are highlighted. Further, the effect of neglecting the pool swell phenomenon on the containment transients is also clearly brought out by a comparative study. The numerical studies presented in this paper give insight into containment transients that would be useful to both the system designer as well as the regulator. (author)

  15. An improved gate valve for critical applications in nuclear power plants

    International Nuclear Information System (INIS)

    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

  16. Laguna Verde annulus pressurization loads evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Castaneda, M. A.; Cruz, M. A.; Cardenas, J. B.; Vargas, A.; Cruz, H. J.; Mercado, J. J., E-mail: miguel.castaneda01@cfe.gob.m [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Km. 42.5 Carretera Cardel-Nautla, Veracruz (Mexico)

    2010-10-15

    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

  17. Steam separator modeling for various nuclear reactor transients

    International Nuclear Information System (INIS)

    In a pressurized water reactor steam generator, a moisture separator is used to separate steam and liquid and to insure that essentially dry steam is supplied to the turbine. During a steam line break or combined steam line break plus tube rupture, a number of phenomena can occur in the separator which have no counterparts during steady-state operation. How the separator will perform under these circumstances is important for two reasons, it affects the carry-over of radioactive iodine and the water inventory in the secondary side. This study has as its goal the development of a simple separator model which can be applied to a variety of steam generator for off-design conditions. Experiments were performed using air and water on three different types of centrifugal separators: a cyclone as a generic separator, a Combustion Engineering type stationary swirl vane separator, and a Westinghouse type separator. The cyclone separator system has three stages of separation: first the cyclone, then a gravity separator, and finally a chevron plate separator. The other systems have only a centrifugal separator to isolate the effect of the primary separator. Experiments were also done in MIT blowdown rig, with and without a separator, using steam and water. The separators appear to perform well at flow rates well above the design values as long as the downcomer water level is not high. High downcomer water level rather than high flow rates appear to be the primary cause of degraded performance. Appreciable carry-over from the separator section of a steam generator occurs when the drain lines from three stages of separation are unable to carry off the liquid flow. Failure scenarios of the separator for extreme range of conditions from the quasi-steady state transient to the fast transients are presented. A general model structure and simple separator models are provided

  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 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.

  20. 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.

  1. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    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

  2. Final environmental statement: Related to the operation of Davis-Besse Nuclear Power Station, Unit 1 (Docket No. 50-346)

    International Nuclear Information System (INIS)

    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

  3. The probability of containment failure by direct containment heating in Zion

    International Nuclear Information System (INIS)

    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 (UO2 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 10-3 for all scenarios considered. Sensitivity analyses did not show any areas of large sensitivity

  4. US nuclear safety. Review and experience

    International Nuclear Information System (INIS)

    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)

  5. Orbit transfer rocket engine technology program

    Science.gov (United States)

    Gustafson, N. B.; Harmon, T. J.

    1993-01-01

    An advanced near term (1990's) space-based Orbit Transfer Vehicle Engine (OTVE) system was designed, and the technologies applicable to its construction, maintenance, and operations were developed under Tasks A through F of the Orbit Transfer Rocket Engine Technology Program. Task A was a reporting task. In Task B, promising OTV turbomachinery technologies were explored: two stage partial admission turbines, high velocity ratio diffusing crossovers, soft wear ring seals, advanced bearing concepts, and a rotordynamic analysis. In Task C, a ribbed combustor design was developed. Possible rib and channel geometries were chosen analytically. Rib candidates were hot air tested and laser velocimeter boundary layer analyses were conducted. A channel geometry was also chosen on the basis of laser velocimeter data. To verify the predicted heat enhancement effects, a ribbed calorimeter spool was hot fire tested. Under Task D, the optimum expander cycle engine thrust, performance and envelope were established for a set of OTV missions. Optimal nozzle contours and quick disconnects for modularity were developed. Failure Modes and Effects Analyses, maintenance and reliability studies and component study results were incorporated into the engine system. Parametric trades on engine thrust, mixture ratio, and area ratio were also generated. A control system and the health monitoring and maintenance operations necessary for a space-based engine were outlined in Task E. In addition, combustor wall thickness measuring devices and a fiberoptic shaft monitor were developed. These monitoring devices were incorporated into preflight engine readiness checkout procedures. In Task F, the Integrated Component Evaluator (I.C.E.) was used to demonstrate performance and operational characteristics of an advanced expander cycle engine system and its component technologies. Sub-system checkouts and a system blowdown were performed. Short transitions were then made into main combustor ignition and

  6. Fluorescence Visualization of Hypersonic Flow Past Triangular and Rectangular Boundary-layer Trips

    Science.gov (United States)

    Danehy, Paul M.; Garcia, A. P.; Borg, Stephen E.; Dyakonov, Artem A.; Berry, Scott A.; Inman, Jennifer A.; Alderfer, David W.

    2007-01-01

    Planar laser-induced fluorescence (PLIF) flow visualization has been used to investigate the hypersonic flow of air over surface protrusions that are sized to force laminar-to-turbulent boundary layer transition. These trips were selected to simulate protruding Space Shuttle Orbiter heat shield gap-filler material. Experiments were performed in the NASA Langley Research Center 31-Inch Mach 10 Air Wind Tunnel, which is an electrically-heated, blowdown facility. Two-mm high by 8-mm wide triangular and rectangular trips were attached to a flat plate and were oriented at an angle of 45 degrees with respect to the oncoming flow. Upstream of these trips, nitric oxide (NO) was seeded into the boundary layer. PLIF visualization of this NO allowed observation of both laminar and turbulent boundary layer flow downstream of the trips for varying flow conditions as the flat plate angle of attack was varied. By varying the angle of attack, the Mach number above the boundary layer was varied between 4.2 and 9.8, according to analytical oblique-shock calculations. Computational Fluid Dynamics (CFD) simulations of the flowfield with a laminar boundary layer were also performed to better understand the flow environment. The PLIF images of the tripped boundary layer flow were compared to a case with no trip for which the flow remained laminar over the entire angle-of-attack range studied. Qualitative agreement is found between the present observed transition measurements and a previous experimental roughness-induced transition database determined by other means, which is used by the shuttle return-to-flight program.

  7. The development and application of overheating failure model of FBR steam generator tubes. 2

    International Nuclear Information System (INIS)

    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)

  8. 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

  9. AP1000 Features Prevent Potential Containment Recirculation Screen Plugging

    International Nuclear Information System (INIS)

    This paper presents the results of plant design development and evaluations that demonstrate that the AP1000 plant is not subject to potential containment recirculation screen plugging following a loss-of-coolant-accident (LOCA). Following a LOCA in a pressurized water reactor, it is necessary to recirculate water from the containment back into the reactor to maintain long term core cooling. The AP1000 utilizes passive safety systems to provide containment recirculation for long term core cooling following a LOCA. The AP1000 also has non-safety pumps which provide a backup means of providing recirculation. Screens are provided around the recirculation pipes to prevent debris from blocking recirculation flow and core cooling passages. Debris may be generated by the LOCA blowdown from insulation and coatings used inside containment. Even with effective cleanliness programs, there may be some resident debris such as dust and dirt. The potential for plugging the recirculation screens is a current PWR licensing issue. The AP1000 design provides inherent advantages with respect to the potential plugging of containment recirculation screens. These characteristics include prevention of fibrous debris generation, improved debris settling and improved recirculation screen design. Debris settling analysis demonstrates that failure of coatings does not result in debris being transported to the screens before it settles to the floor. Additional analysis also shows that the plant can tolerate conservative amounts of resident debris being transported to the screens. The AP1000 significantly reduces the probability of plugging the containment recirculation screens and significantly reduces inspection and maintenance of coatings used inside containment. (authors)

  10. ChemAND - a system health monitor for plant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Turner, C.W.; Mitchel, G.R.; Tosello, G.; Balakrishnan, P.V.; McKay, G.; Thompson, M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Dundar, Y.; Bergeron, M.; Laporte, R. [Hydro-Quebec, Groupe Chimie, Centrale Nucleaire Gentilly-2, Gentilly, Quebec (Canada)

    2001-03-01

    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. These parameters can be used as inputs 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 includes two analytical models developed for the balance-of-plant. The first model, 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 can be used by plant staff to evaluate the susceptibility of the SG tubes to crevice corrosion. 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. The second model, SLUDGE, calculates the deposit loading and distribution in the SG as a function of time, based on concentrations of corrosion product in the final feedwater for both normal and start-up conditions. Operations personnel can use this information to predict where to inspect and when to clean. (author)

  11. 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.

  12. Fluiddynamic effects in the fuel element top nozzle area during refilling and reflooding

    International Nuclear Information System (INIS)

    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

  13. 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)

    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

  14. Experiments of ECCS strainer blockage and debris settling in suppression pools

    Energy Technology Data Exchange (ETDEWEB)

    Hecker, G.E.; Johnson, A.B.; Murthy, P.; Padmanabhan

    1996-03-01

    If a rupture occurs in a nuclear power station pipe that leads to or from the reactor pressure vessel, the resultant Loss of Coolant Accident (LOCA) would initiate a chain of events involving complex flow phenomena. In a Boiling Water Reactor (BWR), the steam or liquid pipe break pressurizes the dry well, forcing the inert containment gases and steam through downcomers into the suppression pool, thoroughly mixing any particulates and pipe insulation debris carried with the gas flow to the pool. As the steam flow decreases, its unsteady condensation at the end of the downcomers (Condensation Oscillation and Chugging) produces continued water motion in the suppression pool and downcomers. During the blowdown event, high pressure and then low pressure pumps automatically start injecting water from the suppression pool into the reactor to keep its temperature under control. Proper functioning of this Emergency Core Cooling System (ECCS) is critical for the first 30 minutes or so, before operators have time to consider and align alternative sources of cooling water. A major concern for proper operation of the ECCS is the effect of fragmented insulation and plant particulates on the head loss at pump suction strainers. Sufficient loss could exceed the NPSH margin, causing cavitation with a resultant loss of pump capacity and longevity. The bead loss increases with the mass of debris accumulated on the pump strainers, which in turn is dependent on the debris concentration versus time in the suppression pool. This paper describes two sets of experiments that quantified the strainer head loss. One set of experiments considered the mixing and settling of fibrous insulation debris and fine iron oxide particles in the suppression pool during and after chugging. These tests used a reduced scale facility which duplicated the kinetic energy per unit water volume to define the concentration of the actual materials in the pool versus time.

  15. 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

  16. Uncertainty Quantification of LBLOCA PCTR for a pressurized water reactor by AAEC Algorithm and Gaussian Process Model

    International Nuclear Information System (INIS)

    Two generalized regression methods, the Alternating Conditional Expectation (ACE) algorithm and a Gaussian process model (GPM), are presented for constructing response surface models used for uncertainty analysis of pressurized water reactor accident simulations by best-estimate thermal-hydraulic codes. Conventional regression techniques are limited by the requirement for a priori estimates of the functional form of the regression model. The ACE algorithm yields an optimal relationship between the dependent variable and multiple independent variables by obtaining one-dimensional transformations of each variable through an iterative procedure that maximizes the statistical correlation between the transformations. The GPM defines predictive distributions of the dependent variable over the multi-dimensional input variable space by taking LOCAy weighted averages of data points with weights determined by a parameterized covariance function. Both methods have been successfully implemented to obtain a probability density function (PDF) for the peak clad temperature (PCT) during the blowdown phase of a large break loss of coolant accident (LBLOCA) in an OPR1000 reactor as a function of 20 uncertain input parameters. The ACE and GPM response surfaces were generated from 400 PCT values obtained from simulations of the LBLOCA by the MARS code. The MOSAIQUE software was used to draw samples from the input parameter distributions, generate MARS input decks, and automate the execution of MARS runs over multiple processors. PCT estimates from both response surfaces agreed well with additional MARS simulations left out of the training data set indicating the ACE algorithm and GPM are satisfactory code surrogates. Furthermore, sensitivity or importance information can be extracted from the models giving physical insight into the input-output relationships, a significant advantage over conventional regression techniques

  17. An organic profile of a pressurised water reactor secondary plant

    International Nuclear Information System (INIS)

    Make-up water addition to the steam/water cycle at Koeberg Nuclear Power Station usually results in a corresponding increase of the chloride concentration in the steam generator blowdown system. During plant transients, when higher than normal make-up is required to the secondary plant, the concentration of chloride occasionally exceeds the limiting value for the station chemistry performance indicator. Irrespective of this, the demineralised water make-up supply tanks, which are routinely analysed for chloride, are within all recognised acceptable standards for secondary water make-up and therefore these tanks do not initially appear to be the source of chloride contamination. Water treatment at the plant relies essentially on ion exchange, which has been proven to be very effective in removing inorganic ionic species such as chloride. Organic compounds are less effectively removed by ion exchange and may pass through the treatment system, and these organics can reside undetected in the make-up water tanks. Historically, the elevated chloride concentration following high system make-up has been attributed to chlorinated organic compounds known as trihalomethanes being present in the make-up water tanks, but no rigorous study had been undertaken. As it has been assumed that the majority of chloride in the secondary system originates from the make-up water organic impurities, it was considered important to confirm this by compiling an organic profile of the secondary plant. The use of organic additives was also taken into account in the profile. This work has confirmed the contribution from trihalomethanes and has also found that other organochlorides contribute even more significantly to the overall chloride inventory of the secondary plant. (orig.)

  18. Predicting Ares I Reaction Control System Performance by Utilizing Analysis Anchored with Development Test Data

    Science.gov (United States)

    Stein, William B.; Holt, K.; Holton, M.; Williams, J. H.; Butt, A.; Dervan, M.; Sharp, D.

    2010-01-01

    The Ares I launch vehicle is an integral part of NASA s Constellation Program, providing a foundation for a new era of space access. The Ares I is designed to lift the Orion Crew Module and will enable humans to return to the Moon as well as explore Mars.1 The Ares I is comprised of two inline stages: a Space Shuttle-derived five-segment Solid Rocket Booster (SRB) First Stage (FS) and an Upper Stage (US) powered by a Saturn V-derived J-2X engine. A dedicated Roll Control System (RoCS) located on the connecting interstage provides roll control prior to FS separation. Induced yaw and pitch moments are handled by the SRB nozzle vectoring. The FS SRB operates for approximately two minutes after which the US separates from the vehicle and the US Reaction Control System (ReCS) continues to provide reaction control for the remainder of the mission. A representation of the Ares I launch vehicle in the stacked configuration and including the Orion Crew Exploration Vehicle (CEV) is shown in Figure 1. Each Reaction Control System (RCS) design incorporates a Gaseous Helium (GHe) pressurization system combined with a monopropellant Hydrazine (N2H4) propulsion system. Both systems have two diametrically opposed thruster modules. This architecture provides one failure tolerance for function and prevention of catastrophic hazards such as inadvertent thruster firing, bulk propellant leakage, and over-pressurization. The pressurization system on the RoCS includes two ambient pressure-referenced regulators on parallel strings in order to attain the required system level single Fault Tolerant (FT) design for function while the ReCS utilizes a blow-down approach. A single burst disk and relief valve assembly is also included on the RoCS to ensure single failure tolerance for must-not-occur catastrophic hazards. The Reaction Control Systems are designed to support simultaneously firing multiple thrusters as required

  19. 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)

    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

  20. 小型贯流锅炉给水处理中的反渗透技术应用%Application of Reverse Osmosis Technology in Feedwater Treatment for Small Tubular Boiler

    Institute of Scientific and Technical Information of China (English)

    贾旭亮; 杨洪东

    2012-01-01

    THe cost of ion exchange water, treatment for small tubular boiler feedwater is about 2.83 RMB/t. Compared with ion exchange water treatment process, the cost of reverse osmosis water treatment is about 5.04 RMB/t. However, the hardness, alkalinity and conductivity of reverse osmosis water is lower than that of ion exchange water, the boiler's continuous water volume of blowdown and discharge are reduced, which saves a lot of fuel. The results show that reverse osmosis water treatment process for boiler feedwater has more superiority in economy with simple operation and wide application prospect.%小型贯流锅炉具有效率高、节能、自动化程度高的特点,但对于给水水质要求也较高.与离子交换法相比,反渗透制水用于小型贯流锅炉给水时制水成本约5.04元/t,而离子交换的制水成本约2.83元/t,但反渗透的产水硬度、碱度、电导率都较离子交换水低,使锅炉的连续排污量和全排污量降低,从而节省了大量燃料费用,符合节能减排的要求.结果表明反渗透制水用于锅炉给水时,比离子交换制水更经济、操作更简单,应用前景广阔.

  1. Cause analysis and treatment of boiler water abnormity in a 300 MW unit%300 MW热电机组锅炉水质异常分析及处理

    Institute of Scientific and Technical Information of China (English)

    刘玮; 马知敬; 贾蓉

    2014-01-01

    新疆米东热电厂1号机组锅炉炉水电导率和 pH 值较高,超出正常范围,对水冷壁管存在腐蚀的风险。检测发现炉水中出现了大量游离NaOH,分析发现其由作为给水补水的热网疏水受热网循环水的污染所致。由于连续排污扩容器的分离蒸汽至除氧器的逆止阀存在缺陷,造成锅炉未真正有效排污,炉水长期处于过度浓缩状态,最终产生大量游离NaOH,对锅炉加强排污后,炉水水质恢复正常。%The conductivity and pH value of boiler water in No.1 unit of Midong Thermal Power Plant were over standard.So the water wall tubes suffered the risk of alkaline corrosion and boiler water concentrate corrosion.Detection finds a larger number of free sodium hydroxide appears in the boiler water.This is because the heat-supply network drain which is used as the feed water supplement water was polluted by leakage of circulating water of district heating system heater. The check valve between the steam separated from continuous blow-down flash vessel and the de-aerator had some defects,leading to invalid blow down,so the boiler water were over concentrat-ed,a larger number of free sodium hydroxide appeared in the boiler water,resulting in boiler wa-ter abnormality eventually.Finally,after improving continuous blow down,the boiler water turned back to normal.

  2. 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).

  3. Flue gas injection control of silica in cooling towers.

    Energy Technology Data Exchange (ETDEWEB)

    Brady, Patrick Vane; Anderson, Howard L., Jr.; Altman, Susan Jeanne

    2011-06-01

    Injection of CO{sub 2}-laden flue gas can decrease the potential for silica and calcite scale formation in cooling tower blowdown by lowering solution pH to decrease equilibrium calcite solubility and kinetic rates of silica polymerization. Flue gas injection might best inhibit scale formation in power plant cooling towers that use impaired makeup waters - for example, groundwaters that contain relatively high levels of calcium, alkalinity, and silica. Groundwaters brought to the surface for cooling will degas CO{sub 2} and increase their pH by 1-2 units, possibly precipitating calcite in the process. Recarbonation with flue gas can lower the pHs of these fluids back to roughly their initial pH. Flue gas carbonation probably cannot lower pHs to much below pH 6 because the pHs of impaired waters, once outgassed at the surface, are likely to be relatively alkaline. Silica polymerization to form scale occurs most rapidly at pH {approx} 8.3 at 25 C; polymerization is slower at higher and lower pH. pH 7 fluids containing {approx}220 ppm SiO{sub 2} require > 180 hours equilibration to begin forming scale whereas at pH 8.3 scale formation is complete within 36 hours. Flue gas injection that lowers pHs to {approx} 7 should allow substantially higher concentration factors. Periodic cycling to lower recoveries - hence lower silica concentrations - might be required though. Higher concentration factors enabled by flue gas injection should decrease concentrate volumes and disposal costs by roughly half.

  4. CFX-10 Analysis of the High Temperature Thermal- Chemical Experiment (CS28-2)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

    2008-02-15

    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.

  5. 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

  6. Characterization Investigation Study: Volume 3, Radiological survey of surface soils

    International Nuclear Information System (INIS)

    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

  7. 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

  8. NASA Ares I Launch Vehicle First Stage Roll Control System Cold Flow Development Test Program Overview

    Science.gov (United States)

    Butt, Adam; Popp, Christopher G.; Holt, Kimberly A.; Pitts, Hank M.

    2010-01-01

    pressurization system, including regulator blowdown and propellant ullage performance, measure system pressure drops for comparison to analysis of tubing and components, and validate system activation and re-activation procedures for the helium pressurant system. Secondary objectives included: validating system processes for loading, unloading, and purging, validating procedures and system response for multiple failure scenarios, including relief valve operation, and evaluating system performance for contingency scenarios. The test results of the cold flow development test program are essential in validating the performance and interaction of the Roll Control System and anchoring analysis tools and results to a Critical Design Review level of fidelity.

  9. Dispersant trial at ANO-2: Qualification for a short-term trial prior to SG replacement

    Energy Technology Data Exchange (ETDEWEB)

    Fruzzetti, K.; Frattini, P. [Electric Power Research Inst., Palo Alto, CA (United States); Robbins, P. [Entergy Operations, Arkansas Nuclear One, Russellville, AR (United States); Miller, A. [Pedro Point Technology, Inc., Pacifica, CA (United States); Varrin, R.; Kreider, M. [Dominion Engineering Inc., McLean, VA (United States)

    2002-07-01

    Corrosion products in the secondary side of pressurized water reactor (PWR) steam generators (SGs) primarily deposit on the SG tubes. These deposits can inhibit heat transfer, lead to thermal-hydraulic instabilities through blockage of tube supports, and create occluded regions where corrosive species can concentrate along tubes and in tube-to-tube support plate crevices. The performance of the SGs is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. Currently, there are two strategies employed by utilities for minimizing deposit formation on steam generator internal surfaces. The first is to minimize the source term, i.e., reduce the amount of corrosion products in the feedwater. Two methods are commonly used to accomplish this goal: chemistry optimization and plant modifications. The first method uses alternate amines to control the at-temperature pH (pH{sub T}) in specific locations of the secondary system, thereby minimizing the corrosion of balance of plant (BOP) metals. The second method requires removal of metals from the secondary system that are a significant source of corrosion products (e.g., replace 90/10 Cu/Ni condenser tubes with titanium). The second strategy for lowering deposit loads utilizes chemical or mechanical means to remove existing deposits from the SGs (e.g., chemical cleaning or sludge lancing). Many utilities have opted for a combination of these two strategies. A third potential strategy for minimizing deposition of corrosion products on SG internal surfaces is to use online dispersant addition to help prevent the corrosion products from adhering to the steam generator surfaces. By inhibiting the deposition of the corrosion products, the dispersant can facilitate more effective removal from the SGs via blowdown. This type of strategy has been employed at fossil boilers for many decades. However, due to the use of inorganic (sulfur and other impurities) polymerization

  10. Passive Safety Optimization in Liquid Sodium-Cooled Reactors

    International Nuclear Information System (INIS)

    structure, and intermixing of high-pressure CO2 and sodium. For the test for determination of the liquidus/solidus temperature, the standard differential thermal analysis (DTA) method will be employed. And for the test for determination of the mobilization temperature, a thermal gravimetric method will be employed. Concerned with the transient freezing and plugging behavior of molten fuel flowing in coolant channels, the overall test apparatus consists of a fuel melt furnace vessel, melt delivery system, test section, catch pan and containment vessel. To investigate the safety issues and key phenomena associated with the CO2 blowdown and pool mixing, three different types of experiment, Bubble Column experiment, Free Sonic-Jet experiment, Impinging Sonic-Jet experiment are designed

  11. SULFUR POLYMER ENCAPSULATION

    International Nuclear Information System (INIS)

    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

  12. 基于VSP2的放热反应失控紧急泄放特性%Characteristics of emergency venting on exothermic reaction runaway based on VSP2

    Institute of Scientific and Technical Information of China (English)

    喻健良; 闫兴清; 孟庭宇; 谢传欣

    2013-01-01

    基于VSP2绝热量热仪,通过增加泄放物收集罐、快速响应气动阀及泄放孔板,开展了放热反应失控泄放实验,详细探讨了输入热功率、初始填充率、泄放压力、泄放直径以及反应物发泡性对泄放能力及泄放物质量的影响.结果表明:二相流泄放能力随输入热功率增大而降低,随初始填充率增大先增大后减小,随泄放压力增大先快速降低后缓慢增加,随泄放直径增大而增大.泄放物质量随输入热功率、初始填充率、泄放压力和泄放直径的增大而增大.发泡性材料能显著降低泄放装置的泄放能力,但能增大泄放物质量.%Through the addition of effluent container,rapid response pneumatic valve and orifice based on VSP2 (Vent Sizing Package 2),the blowdown tests of exothermic reaction runaway were conducted.The impact of heating power,initial fill rate,venting pressure,venting diameter and the reactant foaming property on the venting capacity and the effluent mass were investigated.The results indicate that the venting capacity decreases with the increase of heating power,and increases then falls with the rise of fill rate.In addition,the vent capacity decreases rapidly then increases slowly with the increase of venting pressure.Larger venting diameter leads to larger venting capacity.The effluent mass increases with the rise of heating power,fill rate,venting pressure and venting diameter.Foamy reactants can greatly decrease the venting capacity whereas increase the effluent mass.

  13. Numerical Stability Analysis of New Hyperbolic Equations for Two-Phase Flow

    International Nuclear Information System (INIS)

    capability for the transient analysis and improvement for stability, VERTICAL-CANNON blowdown assessments has been performed. The new model gave similar pressure and void transient to that obtained by the RELAP5/MOD3 model which uses only virtual mass force and showed more stable tendency in relatively high phasic velocities. In conclusion, the new hyperbolic model can be easily adapted to existing nuclear safety codes which use semi-implicit scheme to enhance numerical stability and to hold computational efficiency

  14. PPOOLEX experiments on wall condensation

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

    This report summarizes the results of the wall condensation experiments carried out in December 2008 and January 2009 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 a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows, were carried out. The main purpose of the experiment series was to study wall condensation phenomenon inside the dry well compartment while steam is discharged through it into the condensation pool and to produce comparison data for CFD calculations at VTT. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. For the wall condensation experiments the test facility was equipped with a system for collecting and measuring the amount of condensate from four different wall segments of the dry well compartment. A thermo graphic camera was used in a couple of experiments for filming the outside surface of the dry well wall. The effect of the initial temperature level of the dry well structures and of the steam flow rate for the accumulation of condensate was studied. The initial temperature level of the dry well structures varied from 23 to 99 deg. C. The steam flow rate varied from 90 to 690 g/s and the temperature of incoming steam from 115 to 160 deg. C. During the initial phase of steam discharge the accumulation of condensate was strongly controlled by the temperature level of the dry well structures; the lower the initial temperature level was the more condensate was accumulated. As the dry well structural temperatures increased the condensation process slowed down. Most of the condensate usually accumulated during the first 200 seconds of the discharge. However, the condensation process never completely stopped because a small temperature difference remained between the dry well atmosphere and inner wall

  15. Supercritical water: On a road from CFD to NPP simulations

    International Nuclear Information System (INIS)

    The Fission and Radiation Physics Group at the Aalto University is contributing to the Finnish SCWR activities within the GEN4FIN-network. Our research involves reactor core thermal hydraulics, and in particular, heat transfer phenomena in supercritical water including both theoretical studies and simulations with APROS and OpenFOAM. APROS is a software applicable to full-scale power plant simulations and OpenFOAM an open source CFD code. The complicated heat transfer in the supercritical region is a very challenging problem for the design of SCWRs and their safety assessment. The steam tables of APROS have been extended to the supercritical region and their functionality has been tested with, e.g. blowdown simulations where the transient is rapid, hence mainly challenging for numerical stability whereas heat transfer has negligible effects. Numerous different heat correlations for supercritical water have been suggested , but simulations of benchmark experiments have shown that for instance fuel clad temperatures generally cannot be described sufficiently accurately. This discrepancy has been encountered in several process simulation codes. The largest errors occur near the pseudo critical line, during the heat transfer deterioration. It turns out that the physics in supercritical water is clearly more intricate than in ordinary boiling heat transfer where rather satisfactory heat transfer correlations are available. Full 3D CFD calculations allow a better description of various aspects of heat transfer in the supercritical region, i.e., effects arising from turbulence , buoyancy , varying material properties etc. On the other hand, CFD calculations are not feasible for plant-scale simulations. We have selected some simplified geometries and parameter ranges to study SCW heat transfer in a reactor. Old experiments have been calculated with satisfactory results with OpenFOAM to check its validity. A steady state case of heat transfer in a circular pipe with upward

  16. 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.

  17. CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

    International Nuclear Information System (INIS)

    CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

  18. 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

  19. 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

  20. Water Quality of NPP Secondary Side with Combined Water Chemistry of Ammonia and Ethanolamine

    International Nuclear Information System (INIS)

    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 SO42- 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)

  1. Obtaining a more realistic hydrogen distribution in the containment by coupling MELCOR with GASFLOW

    Energy Technology Data Exchange (ETDEWEB)

    Szabó, Tobias, E-mail: tobias.szabo@kit.edu; Kretzschmar, Frank, E-mail: frank.kretzschmar@kit.edu; Schulenberg, Thomas, E-mail: thomas.schulenberg@kit.edu

    2014-04-01

    Highlights: • Subsequent MELCOR and GASFLOW simulations are inconsistent. This inconsistency can be overcome by coupling both codes. • We tested the existing MELCOR coupling interface. • We developed and verified a coupling of MELCOR and GASFLOW. • We successfully applied the coupling to calculate a THAI experiment. • We compared the coupling to the common methodology. We found that the results from coupling were more realistic. - Abstract: The system code MELCOR provides an integral analysis capability for severe accidents in nuclear power plants. However, its Lumped Parameter model provides less accurate information about the thermal hydraulics in the containment during a loss of coolant accident. GASFLOW is a 3D CFD code that simulates the containment thermal hydraulics and the local hydrogen distribution more realistically. Currently, the common procedure is to use a source term from a previous MELCOR calculation in GASFLOW. Yet, the effect of the more realistic GASFLOW pressure to the mass flow through the leak cannot be taken into account in this approach. This inconsistency can be overcome by coupling both codes. First, the coupling interface existing in MELCOR 1.8.6 was tested by calculating a postulated accident in a simplified BWR using two coupled instances of MELCOR. The results agreed perfectly with the ones from a similar stand-alone calculation. Hence, MELCOR could be coupled to GASFLOW. A GASFLOW interface for an external, explicit, and asynchronous coupling to MELCOR was developed. It enabled to receive the source term from MELCOR and to send back the containment pressure during the run time. The correct functioning of this data exchange was verified for a representative blow-down problem. Moreover, we tested the coupling by calculating the TH7 experiment in the THAI facility with the coupled code system and with GASFLOW in stand-alone mode for comparison. The calculation results agreed very well to one another. Accordingly, the coupling

  2. Landscape-scale effects of the 2008 Chaiten (Chile) eruption on vegetation disturbance and regeneration from satellite image analysis

    Science.gov (United States)

    Moore, K. M.; Jones, J. A.; Swanson, F. J.; Crisafulli, C.

    2010-12-01

    The Chaiten eruption affected all components of old-growth temperate evergreen forest -- trees, canopy epiphytes, vines, and understory - over ~ 400 km2. Vegetation responses differed depending on initial vegetation type, phenologic stage of the vegetation, landscape position, and the character and chronology of geophysical processes. Four distinct phases of the eruption and associated disturbances affected vegetation greenness in satellite imagery: (1) the initial eruption in early May 2008 produced a lateral blast-like pyroclastic density current (PDC) to N and NE and deposited tephra to NE, E, and SE, (2) >650 mm of precipitation May 13-20 caused flooding and redistributed deposits from hillslopes into channels and downstream, (3) collapse of dome in February 2009 caused PDC into Rio Chaiten to S, (4) floods in 2009 (second winter after eruption) rearranged deposits and wood in channels, and exhumed areas buried in earlier phases of the eruption. Vegetation effects included (a) ash deposition on foliage, (b) foliage/branch removal, (c) tree blowdown and scorching of foliage, and (d) burial, removal, and redeposition of wood and sediment. Three trajectories of vegetation change are apparent in satellite imagery: (1) loss of greenness from Dec 2007 to Dec 2008, followed by recovery of greenness in second growing season (Dec 2008-Apr 2009). Field observations indicated (a) tephra washed off trees whose underlying canopies were still intact (thin deposition zone W of volcano), (b) regeneration/sprouting of vegetation whose leaves had been removed or killed (Rio Rayas bridge), (c) regeneration/sprouting of vegetation exhumed by removal of overlying deposits such as in some river channels (Rio Amarillo), or (d) sprouting of understory (Rio Michinmahuida). (2) Initial loss of greenness that persisted through first two growing seasons. Field observations indicated that (a) trees were blown down and little sprouting has occurred (N flank), (b) vegetation was buried in

  3. 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.

  4. Chemical preventive remedies for steam generators fouling and tube support plate blockages

    International Nuclear Information System (INIS)

    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 pH25oC (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

  5. Preparation of an organochlorine transfer agent for crude oil and performance evaluation%原油有机氯转移剂的制备与效果评价

    Institute of Scientific and Technical Information of China (English)

    刘公召; 李雪姣

    2012-01-01

    以合成的三乙基苄基氢氧化铵为相转移催化剂,与乙醇钠、溶剂复配后制得一种原油有机氯转移剂,在原油电脱盐、脱水条件下将原油中的有机氯转化为无机氯而将其脱除,减少氯对后续工序的危害.在实验室分别对四种原油进行了脱有机氯的效果评价,结果表明:随着加剂量的增加,有机氯的脱除率增加.当加剂量为20~40μg/g时,原油中有机氯质量分数可降低到3μg/g以下.该有机氯转移剂在500 Mt/a的常减压蒸馏装置上进行了工业试验,结果表明:加剂量为10μg/g时,电脱盐罐排水中氯离子含量增加,常顶及初顶冷凝水中氯离子质量分数分别减少69.71%和18.92%,脱除有机氯的效果比较明显.该原油有机氯转移剂无毒无害,加注方便,加剂费用低,不影响装置的正常生产.%The triethyl benzyl ammonium hydroxide, as a phase transfer catalyst, is complexed with sodium ethylate and solvent to prepare an organochlorine transfer agent for crude oil, which can convert the or-ganochlorines in the crude oil to inorganochlorines under the process conditions of electric desalting and dehydration to remove organiochlorines and to reduce their detrimental effects in the downstream units. The results of laboratory evaluations of organochlorine transfer agent on 4 crude oils show that the organochlorine removal rate rises with increased dosage of agent. When it is added at a dosage of 20 to 40 μg/g, the organochlorine in crude oil is lowered to less 3 μg/g. The application of the organochlorine transfer agent in a 5. 0 MM TPY commercial atmospheric-vacuum distillation unit demonstrates that, when it is added at a dosage of 10 μg/g, the chlorine ions in blowdown water of desalter drum are increased, und chlorine ions in the overhead condensates of atmospheric tower and preliminary tower are reduced by 66.68% and 18.92% respectively. The de-chlorination effect of the agent is obvious. The organochlorine

  6. Technical meeting on heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors. Book of abstracts

    International Nuclear Information System (INIS)

    There is high interest internationally in both developing and industrialized countries in the design of innovative supercritical water cooled reactors (SCWRs). This interest arises from the high thermal efficiencies (44-45%) and improved economic competitiveness promised by this concept, which utilizes and builds upon the recent developments of highly efficient fossil power plants. The SCWR is one of the six concepts included in the Generation-IV International Forum (GIF). In support of Member States' efforts in the area of SCWRs, the IAEA started in 2008 a Coordinated Research Project (CRP) on 'Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs'. The two key objectives of this CRP are: 1) To establish accurate databases for heat transfer, pressure drop, blowdown, natural circulation, and stability for conditions relevant to supercritical pressure fluids, and 2) To test analysis methods for SCWR thermohydraulic behaviour, and identify code development needs. Annual Research Coordination Meetings take place under the framework of this CRP once a year to assess the progress of the project. These meetings are mainly focused on programmatic issues associated with the project and very little time is available for discussion on details and specific technical areas. This is why during the 2nd Research Coordination Meeting for this CRP held in Vienna in August 24-27, 2009, Member States expressed an interest in organizing a technical meeting in which specialists in the areas of heat transfer and thermal-hydraulics, thermodynamics and systems design for supercritical water cooled reactors would have the opportunity of participating in extended technical discussions on the details associated to the science and engineering of supercritical water cooled reactor concepts. The University of Pisa kindly offered to host such a technical meeting. The purpose of the meeting was to provide a platform for detailed presentations and technical discussions leading, to

  7. Online recognition of the multiphase flow regime and study of slug flow in pipeline

    Science.gov (United States)

    Liejin, Guo; Bofeng, Bai; Liang, Zhao; Xin, Wang; Hanyang, Gu

    2009-02-01

    single sensor performance. Among various flow patterns of gas-liquid flow, slug flow occurs frequently in the petroleum, chemical, civil and nuclear industries. In the offshore oil and gas field, the maximum slug length and its statistical distribution are very important for the design of separator and downstream processing facility at steady state operations. However transient conditions may be encountered in the production, such as operational upsets, start-up, shut-down, pigging and blowdown, which are key operational and safety issues related to oil field development. So it is necessary to have an understanding the flow parameters under transient conditions. In this paper, the evolution of slug length along a horizontal pipe in gas-liquid flow is also studied in details and then an experimental study of flowrate transients in slug flow is provided. Also, the special gas-liquid flow phenomena easily encountered in the life span of offshore oil fields, called severe slugging, is studied experimentally and some results are presented.

  8. 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

  9. Management of injected nitrogen into a gas condensate reservoir

    Directory of Open Access Journals (Sweden)

    Hadi Belhaj

    2016-04-01

    Full Text Available This study investigates the means of deferring the breakthrough of injected N2 and alleviating the impact of such on production rates and specifications as well as minimizing the required changes to the gas processing facilities. This aimed at assisting the ongoing efforts to transfer the Cantarell experience to Abu Dhabi, where large amounts of N2 gas will be generated and injected into a large gas condensate reservoir to partially substitute the recycling of lean gas. This will bring forward the opportunity to exploit lean gas by securing base load supplies before the start of reservoir blowdown, compared to the peak shaving approach currently practiced. Managing N2 breakthrough starts by better understanding the pattern at which N2 injection spreads into the gas accumulation. Based on the findings of initial subsurface and plant simulations carried out in 2008, N2 breakthrough in Abu Dhabi might be possibly deferred by segmenting the reservoir into a rich N2 region and lean N2 region. The approach assumes no thief zones will be faced and no channeling of N2 injected between the two regions is taking place. N2 is injected in the north region of the reservoir. The production of that region will be segregated and fed to a gas processing plant of lower NGL (natural gas liquid recovery, which essentially takes longer time to start suffering the deterioration of residue gas (gas mixture resulted after separating NGL quality. The residue gas use can be limited to re-injection where the effect of below specification LHV (Low Heat Value would not be an issue. The rest of the reservoir feeds another gas processing plant of higher NGL recovery level from which an amount of residue gas equivalent to that of the injected N2 will be rerouted to the sales network. This scenario will significantly delay as well as downsize the requirement of a N2 rejection plant. There is technical and certainly economical advantage of deferring the

  10. 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.

  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; Gershenzon, Mike; Ghosh, Rajat

    2012-07-31

    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 a refinery location. For both cases the in-duct scrubber technology was compared to traditional amine- based capture

  12. 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

    International Nuclear Information System (INIS)

    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

  13. Implications of Air Ingress Induced by Density-Difference Driven Stratified Flow

    International Nuclear Information System (INIS)

    One of the design basis accidents for the Next Generation Nuclear Plant (NGNP), a high temperature gas-cooled reactor, is air ingress subsequent to a pipe break. Following a postulated double-ended guillotine break in the hot duct, and the subsequent depressurization to nearly reactor cavity pressure levels, air present in the reactor cavity will enter the reactor vessel via density-gradient-driven-stratified flow. Because of the significantly higher molecular weight and lower initial temperature of the reactor cavity air-helium mixture, in contrast to the helium in the reactor vessel, the air-helium mixture in the cavity always has a larger density than the helium discharging from the reactor vessel through the break into the reactor cavity. In the later stages of the helium blowdown, the momentum of the helium flow decreases sufficiently for the heavier cavity air-helium mixture to intrude into the reactor vessel lower plenum through the lower portion of the break. Once it has entered, the heavier gas will pool at the bottom of the lower plenum. From there it will move upwards into the core via diffusion and density-gradient effects that stem from heating the air-helium mixture and from the pressure differences between the reactor cavity and the reactor vessel. This scenario (considering density-gradient-driven stratified flow) is considerably different from the heretofore commonly used scenario that attributes movement of air into the reactor vessel and from thence to the core region via diffusion. When density-gradient-driven stratified flow is considered as a contributing phenomena for air ingress into the reactor vessel, the following factors contribute to a much earlier natural circulation-phase in the reactor vessel: (a) density-gradient-driven stratified flow is a much more rapid mechanism (at least one order of magnitude) for moving air into the reactor vessel lower plenum than diffusion, and consequently, (b) the diffusion dominated phase begins with a

  14. Chemical treatment of deposits of junctions 'collector-tube' of horizontal steam generators

    International Nuclear Information System (INIS)

    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)

  15. ForWarn: A Cross-Cutting Forest Resource Management and Decision Support System Providing the Capacity to Identify and Track Forest Disturbances Nationally

    Science.gov (United States)

    Hargrove, W. W.; Spruce, J.; Norman, S.; Christie, W.; 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. ForWarn is the first national-scale system of its kind based on remote sensing developed specifically for forest disturbances. It has operated as a prototype since January 2010 and has provided useful information about the location and extent of disturbances detected during the 2011 growing season, including tornadoes, wildfires, and extreme drought. The ForWarn system had an official unveiling and rollout in March 2012, initiated by a joint NASA and USDA press release. The ForWarn home page has had 2,632 unique visitors since rollout in March 2012, with 39% returning visits. ForWarn was used to map tornado scars from the historic April 27, 2011 tornado outbreak, and detected timber damage within more than a dozen tornado tracks across northern Mississippi, Alabama, and Georgia. ForWarn is the result of an ongoing, substantive cooperation among four different government agencies: USDA, NASA, USGS, and DOE. Disturbance maps are available on the

  16. 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

  17. Online recognition of the multiphase flow regime and study of slug flow in pipeline

    Energy Technology Data Exchange (ETDEWEB)

    Guo Liejin; Bai Bofeng; Zhao Liang; Wang Xin; Gu Hanyang, E-mail: lj-guo@mail.xjtu.edu.c [Xi' an Jiaotong University, Xi' an 710049 (China)

    2009-02-01

    sensor performance. Among various flow patterns of gas-liquid flow, slug flow occurs frequently in the petroleum, chemical, civil and nuclear industries. In the offshore oil and gas field, the maximum slug length and its statistical distribution are very important for the design of separator and downstream processing facility at steady state operations. However transient conditions may be encountered in the production, such as operational upsets, start-up, shut-down, pigging and blowdown, which are key operational and safety issues related to oil field development. So it is necessary to have an understanding the flow parameters under transient conditions. In this paper, the evolution of slug length along a horizontal pipe in gas-liquid flow is also studied in details and then an experimental study of flowrate transients in slug flow is provided. Also, the special gas-liquid flow phenomena easily encountered in the life span of offshore oil fields, called severe slugging, is studied experimentally and some results are presented.

  18. 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

  19. Implications of Air Ingress Induced by Density-Difference Driven Stratified Flow

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Soo Kim; Richard Schultz; David Petti; C. P. Liou

    2008-06-01

    One of the design basis accidents for the Next Generation Nuclear Plant (NGNP), a high temperature gas-cooled reactor, is air ingress subsequent to a pipe break. Following a postulated double-ended guillotine break in the hot duct, and the subsequent depressurization to nearly reactor cavity pressure levels, air present in the reactor cavity will enter the reactor vessel via density-gradient-driven-stratified flow. Because of the significantly higher molecular weight and lower initial temperature of the reactor cavity air-helium mixture, in contrast to the helium in the reactor vessel, the air-helium mixture in the cavity always has a larger density than the helium discharging from the reactor vessel through the break into the reactor cavity. In the later stages of the helium blowdown, the momentum of the helium flow decreases sufficiently for the heavier cavity air-helium mixture to intrude into the reactor vessel lower plenum through the lower portion of the break. Once it has entered, the heavier gas will pool at the bottom of the lower plenum. From there it will move upwards into the core via diffusion and density-gradient effects that stem from heating the air-helium mixture and from the pressure differences between the reactor cavity and the reactor vessel. This scenario (considering density-gradient-driven stratified flow) is considerably different from the heretofore commonly used scenario that attributes movement of air into the reactor vessel and from thence to the core region via diffusion. When density-gradient-driven stratified flow is considered as a contributing phenomena for air ingress into the reactor vessel, the following factors contribute to a much earlier natural circulation-phase in the reactor vessel: (a) density-gradient-driven stratified flow is a much more rapid mechanism (at least one order of magnitude) for moving air into the reactor vessel lower plenum than diffusion, and consequently, (b) the diffusion dominated phase begins with a

  20. Isotope exchange kinetics in metal hydrides I : TPLUG model.

    Energy Technology Data Exchange (ETDEWEB)

    Larson, Rich; James, Scott Carlton; Nilson, Robert H.

    2011-05-01

    pronounced deviations at long times. These discrepancies can be overcome by postulating the presence of a surface poison such as carbon monoxide, but this explanation is highly speculative. When the method is applied to D {yields} H exchanges intentionally poisoned by known amounts of CO, the fitting results are noticeably degraded from those for the nominally CO-free system but are still tolerable. When TPLUG is used to simulate a blowdown-type experiment, which is characterized by large and rapid changes in both pressure and temperature, discrepancies are even more apparent. Thus, it can be concluded that the best use of TPLUG is not in simulating realistic exchange scenarios, but in extracting preliminary estimates for the kinetic parameters from experiments in which variations in temperature and pressure are intentionally minimized.

  1. 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

  2. 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.

  3. The development of a non-equilibrium dispersed flow film boiling heat transfer modeling package

    Science.gov (United States)

    Meholic, Michael J.

    The dispersed flow film boiling (DFFB) heat transfer regime is important to several applications including cryogenics, rocket engines, steam generators, and in the safety analysis of nuclear reactors. Most notably, DFFB is responsible for the heat transfer during the blowdown and reflood portions of the postulated loss-of-coolant-accident (LOCA). Such analyses require the accurate predictions of the heat transfer resulting from the non-equilibrium conditions present in DFFB. A total of six, interrelated heat transfer paths need to be modeled accurately in order to quantify DFFB heat transfer. Within the nuclear industry, transient safety analysis codes, such as COBRA-TF, are used to ensure the safety of the reactor under various transient and accident scenarios. An extensive literature review of DFFB heat transfer highlighted a number of correlative, phenomenological, and mechanistic models. The Forslund-Rohsenow model is most commonly implemented throughout the nuclear industry. However, several of the models suggested by Forslund and Rohsenow to model DFFB phenomena are either inapplicable for nuclear reactors or do not provide an accurate physical representation of the true situation. Deficiencies among other DFFB heat transfer models in their applicability to nuclear reactors or in their computational expenses motivated the development of a mechanistically based DFFB model which accounted for each heat transfer mechanism explicitly. The heat transfer resulting from dispersed droplets contacting the heated wall in DFFB was often neglected in previous models. In this work, a first-principles approach was implemented to quantify the heat transfer attributed to direct contact. Lagrangian droplet trajectory calculations incorporating realistic radial vapor velocity and temperature profiles were performed to determine if droplets could contact the heated wall based upon the local conditions. These calculations were performed over a droplet size spectrum accounting

  4. A shallow water equation solver and particle tracking method to evaluate the debris transport

    International Nuclear Information System (INIS)

    Debris generated by loss-of-coolant accident (LOCA) may run all over the containment floor, block the sump screen (or strainer), increase the hydraulic head loss across the screen, and eventually, have an adverse effect to long term recirculation cooling operation in pressurized water reactor (PWR). To resolve the problem from the issue, the replacement of containment recirculation sump strainer is being performed for the most of operating nuclear power plants (NPP) having limited strainer areas. The screen area required to incorporate the potential debris loading has been determined using the transport fraction (TF) defined by a ratio of amount of debris accumulated on screen to one generated by LOCA. For the most conventional NPP, the debris transport to the sump screen is initiated by the recirculation actuation. Therefore, evaluation of TF was based on the separated analyses on how the debris generated by LOCA is distributed to the containment floor before recirculation and on how much debris is transported by the flow in the containment pool after recirculation, respectively. This led to an approach to obtain TF values for blowdown phase, wash-down phase, pool recirculation phase, separately. Especially, the TF during recirculation phase has been calculated by steady state CFD analysis based that the break flow and recirculation safety injection flow are balanced, thus steady state flow field over the containment is established. However, such a phase separation cannot be applied to some NPP like the Advanced Power Reactor 1400 (APR1400) having no recirculation operation. Transport of debris to sump in the APR1400 is initiated from the early phase of a LOCA in fully transient manner. The present study is to calculate debris transport on the containment floor to sump in APR1400. For this purpose, a hydraulic model to calculate the transient flow field and a particle tracking model to trace the debris particle within the calculation domain are discussed

  5. Experimental study of choking flow of water at supercritical conditions

    Science.gov (United States)

    Muftuoglu, Altan

    Future nuclear reactors will operate at a coolant pressure close to 25 MPa and at outlet temperatures ranging from 500°C to 625°C. As a result, the outlet flow enthalpy in future Supercritical Water-Cooled Reactors (SCWR) will be much higher than those of actual ones which can increase overall nuclear plant efficiencies up to 48%. However, under such flow conditions, the thermal-hydraulic behavior of supercritical water is not fully known, e.g., pressure drop, forced convection and heat transfer deterioration, critical and blowdown flow rate, etc. Up to now, only a very limited number of studies have been performed under supercritical conditions. Moreover, these studies are conducted at conditions that are not representative of future SCWRs. In addition, existing choked flow data have been collected from experiments at atmospheric discharge pressure conditions and in most cases by using working fluids different than water which constrain researchers to analyze the data correctly. In particular, the knowledge of critical (choked) discharge of supercritical fluids is mandatory to perform nuclear reactor safety analyses and to design key mechanical components (e.g., control and safety relief valves, etc.). Hence, an experimental supercritical water facility has been built at Ecole Polytechnique de Montreal which allows researchers to perform choking flow experiments under supercritical conditions. The facility can also be used to carry out heat transfer and pressure drop experiments under supercritical conditions. In this thesis, we present the results obtained at this facility using a test section that contains a 1 mm inside diameter, 3.17 mm long orifice plate with sharp edges. Thus, 545 choking flow of water data points are obtained under supercritical conditions for flow pressures ranging from 22.1 MPa to 32.1 MPa, flow temperatures ranging from 50°C to 502°C and for discharge pressures from 0.1 MPa to 3.6 MPa. Obtained data are compared with the data given in

  6. 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.

  7. 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

  8. Results of MACE tests M0 and M1

    International Nuclear Information System (INIS)

    -water interaction stage of M0 was very vigorous, removed heat from the melt pool at 3.5 MW/m2 for about three minutes, and extracted an amount of energy greater than the heat of solidification of the entire melt mass. There was no steam explosion. 2) M0 data suggests that basemat erosion was temporarily halted following the initial aggressive quench period. 3) A bridge crust was formed which anchored to the test section sidewalls and prevented water ingression into the corium zone. This would not be expected at reactor scale. The crust was sufficiently porous to permit upward passage of concrete decomposition gasses. 4) Both tests M0 and M1 showed evidence of periodic occurrence of large pool swelling. The pool swelled to contact the underside of the crust and additionally caused extrusion of melt above the crust (eruption when accompanied by gas release/blowdown effects). This significantly augmented the corium quench process and depleted the remaining corium mass interacting with the concrete. The dispersed debris ranged from particles (which formed a particle bed atop the crust) to a large extruded mass. The corium pool void fraction increased to as high as 56% in M0 to account for this extrusion. 5) The upward heat flux after crust formation in M0 amounted to 600 kW/m2 which gradually diminished to ∼150 kW/m2 as the extruded/dispersed mass grew in depth atop the crust. 6) Test M1 showed evidence of 'ablation bursts' early in the test while appreciable metal remained in the corium composition. 7) Test M1 demonstrated the effectiveness of the overlying water to trap released aerosol. It is premature to attempt to draw conclusions on melt coolability pertaining to the reactor system from the M0 and M1 tests. Their value lies first in guiding model development as regards the stages of cooling, crust formation, and pool swell effects. Secondly, these tests are stepping stones to improved tests addressing melt cooling phenomena

  9. 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

  10. 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

  11. 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

  12. 台风对森林的影响%A review of the effect of typhoon on forests

    Institute of Scientific and Technical Information of China (English)

    刘斌; 潘澜; 薛立

    2012-01-01

    Typhoons are one of the most important natural disturbances, which affect forest stand structures and dynamics in various ways such as by snapping branches, stripping leaves and fruits, uprooting trunks and breaking stem. Many factors, such as position in canopy, soil conditions, stage of growth, physiognomy, affect wind resistance of tree species. Moreover, there is a wind velocity beyond which all species are affected by the sheer magnitude of the wind's kinetic energy. Therefore, forest damage varies as a function of tree species, tree age, forest type, tree height and topographic location. High density forests often suffer great damage and mortality due to their relatively weak roots and a high ratio of tree height to diameter at breast height. In forest ecosystems, typhoon disturbance is one of the major agents generating a mosaic of heterogeneous habitat patches at various spatial scales. Canopy gaps caused by typhoon disturbance can occur in a variety of sizes, from single fallen trees to large-scale blowdowns, which is important factor affecting species composition and some critical ecosystem patterns and processes such as understory light environments, and nutrient cycling, and have major effects on patterns of forest regeneration through differential effects on tree species and impacts on resource availability. Light increases due to canopy gaps and soil nutrient availability in the understory also increases due to a decrease in uptake by disturbed canopy trees, which can promote the immigration of early successional species. Moreover, species richness is affected by coarse woody debris, litter layer, pits, and mounds, and the availability of propagules. Coarse woody debris and litterfall caused by typhoon promote return of forest carbon to soil, and affect distribution of soil nutrients. Typhoon disturbances can return large amounts of plant material into the forest floor. Litterfall, particularly green leaves resultingfrom the typhoons have higher

  13. 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

  14. 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