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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. Steam generator blowdown system upgrades

    International Nuclear Information System (INIS)

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

  4. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    There is about 80m3/h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  5. 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)

  6. 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...

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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.

  12. LOFT system structural response during subcooled blowdown

    International Nuclear Information System (INIS)

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

  13. PIV measurement at the blowdown pipe outlet

    International Nuclear Information System (INIS)

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

  14. The NRU blowdown test facility commissioning program

    International Nuclear Information System (INIS)

    A major experimental program has been established at the Chalk River Nuclear Laboratories (CRL) that will provide essential data on the thermal and mechanical behaviour of nuclear fuel under abnormal reactor operating conditions and on the transient release, transport and deposition of fission product activity from severely degraded fuel. A number of severe fuel damage (SFD) experiments will be conducted within the Blowdown Test Facility (BTF) at CRL. A series of experiments are being conducted to commission this new facility prior to the SFD program. This paper describes the features and the commissioning program for the BTF. A development and testing program is described for critical components used on the reactor test section. In-reactor commissioning with a fuel assembly simulator commenced in 1989 June and preliminary results are given. The paper also outlines plans for future all-effects, in-reactor tests of CANDU-designed fuel. (author). 11 refs., 3 tabs., 7 figs

  15. AP600 post blowdown containment transients

    International Nuclear Information System (INIS)

    The containment response in a hypothetical large loss of coolant accident can be divided into two major phases. A blowdown phase which lasts less than a minute and a long term phase, which is the topic of this paper, lasting for the remainder of the transient - on the order of hours. The long term containment pressure is controlled by the rate of heat exchange processes on the outer containment wall and by the driving force of heat and mass sources continuously released from the primary system. In order to estimate the magnitude of the primary energy sources, we have chosen to examine a hypothetical event in which the core remains partially uncovered for an extended period of time following a large loss of coolant accident. (authors). 3 refs., 2 figs

  16. 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)

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

    OpenAIRE

    Sunudas T

    2013-01-01

    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 ...

  18. Depressurization study of supercritical fluid blowdown from simple vessel

    International Nuclear Information System (INIS)

    Highlights: • A transient analysis code is developed to simulate the supercritical blowdown. • Models are established for the supercritical region and subcritical region. • The code is employed to calculate the supercritical blowdown for different fluids. • The code was verified by the blowdown experiment of supercritical CO2. • The blowdown of supercritical water was investigated in detail with the code. - Abstract: The loss of coolant accident (LOCA), particularly the depressurization process, is one of the difficulties in safety analysis of supercritical water-cooled reactor (SCWR). In this study, a comprehensive mathematic-physical model was established and a transient analysis code was developed to simulate the blowdown behaviors of SCWR in a large container. Three alternative phase separation models were adopted to calculate the stagnation enthalpy of the two-phase fluid in the container. Break flow rate models were established for different thermodynamic regions, including the supercritical region, the subcooled region, the overheating region and the two-phase region. The code was verified by comparison with blowdown experiment of supercritical CO2 which shows a good agreement. Then the blowdown of supercritical water from simple vessel was investigated in detail with the code. The effect of initial conditions on pressure transitions was discussed for different regions divided by the relationship between the initial temperature and the corresponding pseudo-critical temperature. Furthermore, both the depressurization speed and the void fraction increase with the increase of initial temperature and the decrease of the initial pressure, yet the fluid inventory has an opposite trend. Discharge speed varies directly with break area, and the pressure transition which turns up earlier remains a constant value. These investigations may lay a theoretical foundation for the accident analysis of SCWR

  19. 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

  20. 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

  1. 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

  2. Steam blowdown experiments with the condensation pool test rig

    International Nuclear Information System (INIS)

    During a possible loss-of-coolant accident (Local) a large amount of non-condensable (nitrogen) and condensable (steam) gas is blown from the upper drywell of the containment to the condensation pool through the blowdown pipes at the boiling water reactors (BWRs). The wet well pool serves as the major heat sink for condensation of steam. The blowdown causes both dynamic and structural loads to the condensation pool. There might also be a risk that the gas discharging to the pool could push its way to the emergency core cooling systems (ECCS) and undermine their performance. (author)

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. Blowdown wind tunnel control using an adaptive fuzzy PI controller

    Directory of Open Access Journals (Sweden)

    Corneliu Andrei NAE

    2013-09-01

    Full Text Available The paper presents an approach towards the control of a supersonic blowdown wind tunnel plant (as evidenced by experimental data collected from “INCAS Supersonic Blowdown Wind Tunnel” using a PI type controller. The key to maintain the imposed experimental conditions is the control of the air flow using the control valve of the plant. A proposed mathematical model based on the control valve will be analyzed using the PI controller. This control scheme will be validated using experimental data collected from real test cases. In order to improve the control performances an adaptive fuzzy PI controller will be implemented in SIMULINK in the present paper. The major objective is to reduce the transient regimes and the global reduction of the start-up loads on the models during this phase.

  9. 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.

  10. Nuclear commissioning of the NRU blowdown test facility

    International Nuclear Information System (INIS)

    The Blowdown Test Facility in the NRU reactor will be used to conduct all-effects experiments under postulated Loss-of-Coolant Accident and Severe Fuel Damage conditions. Experiments conducted in the BTF will provide information on the release, transport and deposition of fission products, and the thermal and mechanical behaviour of nuclear fuel under these conditions. This paper describes results from the nuclear commissioning experiment for the BTF. (2 refs., 4 figs.)

  11. Condensation pool experiments with steam using DN200 blowdown pipe

    International Nuclear Information System (INIS)

    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)

  12. 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.

  13. 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)

  14. BWR blowdown/emergency core cooling integral program

    International Nuclear Information System (INIS)

    The Program plan identifying the phased approach to testing has been completed and the first test phase is well underway. Test results to date show the expected lower peak cladding temperatures and slower system blowdown response for a BWR/6 LOCA simulation compared to the BWR/4. Counter-current flow limiting (CCFL) tests show that exact geometric replication was not necessary to simulate CCFL characteristics for a BWR prototype fuel bundle upper tie plate. Separate effects blowdown tests demonstrate the importance of flow length scaling in simulating break geometry for limiting the critical or blowdown flow rate. A new bundle thermal hydraulic method, MAYU04, has been completed. This method is shown to provide a substantial improvement in the prediction of bundle temperatures. The BD/ECC Program represents an important contribution to BWR safety research. Early results from the program have already provided a better understanding of the governing phenomena during hypothetical LOCA simulation tests. Future tests are expected to provide a basis for further improvements in BWR LOCA phenomena modeling

  15. 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

  16. BWR blowdown/emergency core cooling program: 64-rod bundle blowdown heat transfer (8 x 8 BDHT). Final report

    International Nuclear Information System (INIS)

    System performance and thermal response characteristics of BWR's, during the blowdown phase of the postulated loss-of-coolant accident (LOCA) conditions, were investigated in a test apparatus with different scaled configurations. Effects of the configuration changes on the system responses were identified. Test data obtained serve as baseline data for the BD/ECC interaction experiments and information to tie back to the previous 7 x 7 BDHT tests. Comparisons of the predictions with the tests were made

  17. 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

  18. 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

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

    International Nuclear Information System (INIS)

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

  20. Heat transfer considerations for the first nuclear blowdowns

    International Nuclear Information System (INIS)

    The first nuclear blowdowns were carried out in the Power Burst Facility at the Idaho National Engineering Laboratory as the LOC-11 series of experiments. This test series was designed to simulate a blowdown transient in a pressurized water reactor (PWR) so that nuclear fuel performance could be investigated under conditions representative of the PWR 15 x 15 fuel element design. Post-test calculations using the RELAP4 computer program were performed for the LOC-11B and LOC-11C tests. Comparisons between calculations and experimental data revealed that the ability to accurately model (1) critical heat flux (CHF) during low core flow conditions, (2) initial stored energy in the fuel rods, and (3) radiative heat transfer between fuel rods and shrouds, was required to adequately represent the fuel rod thermal behavior. Pre-test calculations performed using RELAP4 with licensing-type heat transfer and fuel rod models resulted in peak cladding temperatures several hundred K higher than measured, thus providing further evidence of the need to accurately model heat transfer and fuel rod behavior

  1. A pipeline depressurization model for fast decompression and slow blowdown

    International Nuclear Information System (INIS)

    The development and validation of CFD-DECOM, a pipeline depressurization model based on the arbitrary Lagrangian–Eulerian method (ALE) and the homogeneous equilibrium assumption is presented. In CFD-DECOM, the convection terms are separately solved from the other terms in a sub-cycled explicit manner using a sub-timestep that is only a fraction of the main computational timestep. This approach significantly simplifies the solution procedure and improves the computational efficiency. The model is validated against five release scenarios including one fast decompression of a rich gas pipeline and four slow blowdown cases of a liquefied petroleum gas pipeline. The predicted pressure, temperature and fluid inventory-time traces are found to be in good agreement with the measurements in all the cases. - Highlights: • CFD-DECOM, a CFD based pipeline depressurization model has been developed. • The model is computationally efficient. • The model can handle both pipeline decompression and slow blowdown. • The model is validated against 5 different release scenarios. • The predictions are in good agreement with the measurements in all the cases

  2. 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

  3. 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.

  4. 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.

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

    International Nuclear Information System (INIS)

    Objective is to give a best-estimate prediction of transient critical heat flux (CHF) during reactor transients and hypothetical accidents. A predictional method has been developed which 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. 234 figures, 13 tables

  6. Transient critical heat flux and blowdown heat transfer studies

    International Nuclear Information System (INIS)

    The objective of this study was 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 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. A summary of the prediction results is presented where 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 which occurs in a more tranquil state with slowly decaying mass velocity. In many instances, the early CHF can be well correlated by the chi = 1.0 criterion; this is certainly indicative of an annular-flow dryout type crisis. The delay CHF was found to occur at near or above 80% void fraction, and the success of the modified Zuber pool-boiling correlation suggests that this CHF is caused by flooding and pool-boiling type hydrodynamic crisis

  7. Redesign of steam generator blowdown system of Cernavoda Unit 2

    International Nuclear Information System (INIS)

    In 2005 the WANO changed the admissible values of the chemical parameters of the fluid in the steam generators to the following upper limits: - Na concentration: less than 2 μ/kg; - Cl concentration: less than 5 μ/kg; - SO4 concentration: less than 5 μ/kg. Current values from design manual of the chemical control system are: - Na concentration: less than 70 μ/kg; - Cl concentration: less than 100 μ/kg; - SO4 concentration: less than 100 μ/kg. In order to comply with the new requirements we had to increase the steam blowdown flow up to 3 times earlier operating value. We developed a system model in order to analyze system modifications and the implication to other components. One presents the required calculation steps concerning the redesign of the system that has to be operated at a higher nominal flow. The calculations requested various computational approaches such as: - Two phase flow model; - check and design calculation for regulating valves of three types: valve with liquid, gas or two phase flow inside the valve; - Flash tank; - Heat exchanger verification and design; - Flow element operating range checking. This paper presents actually the first approach of the system. At the moment we developed a stationary model of the system and the aim for the near future is to obtain a transient regime model. It will analyze the system behavior at flow changing from minimum to maximum load and especially the problems that occur at the start up of stand-by components. Currently, from operating experience of the system there are some very difficult transients during start up of stand-by heat exchanger in case that operating heat exchanger has to be isolated. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-15

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

  9. A study for the replacement of blowdown loads analysis code of Korea standard nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, B. T.; Hwang, S. T.; Choi, D. S.; Cho, C. S. [Korea Nuclear Fuel Company, Taejon (Korea, Republic of)

    2002-10-01

    The purpose of this study is to investigate the possibility to replace the CEFLASH-4B code used in the blowdown loads analysis of Korea Standard Nuclear Plants. Since the application of CEFLASH-4B is restricted after 2007 by U.S. Government, an alternative code to CEFLASH-4B is necessary. The SATAN-VI code was selected as an alternative choice to the CEFLASH-4B code since it was widely used in LOCA analyses to Westinghouse plants without any further charge. The SATAN-VI code was evaluated for the application to the blowdown loads analysis. With a few problems fixed and/or improved, SATAN-VI code is reasonably applicable to blowdown loads analysis in KSNP plants.

  10. A study for the replacement of blowdown loads analysis code of Korea standard nuclear plants

    International Nuclear Information System (INIS)

    The purpose of this study is to investigate the possibility to replace the CEFLASH-4B code used in the blowdown loads analysis of Korea Standard Nuclear Plants. Since the application of CEFLASH-4B is restricted after 2007 by U.S. Government, an alternative code to CEFLASH-4B is necessary. The SATAN-VI code was selected as an alternative choice to the CEFLASH-4B code since it was widely used in LOCA analyses to Westinghouse plants without any further charge. The SATAN-VI code was evaluated for the application to the blowdown loads analysis. With a few problems fixed and/or improved, SATAN-VI code is reasonably applicable to blowdown loads analysis in KSNP plants

  11. An assessment of RELAP5-3D using the Edwards-O'Brien Blowdown problem

    International Nuclear Information System (INIS)

    The RELAP5-3D (version bt) computer code was used to assess the United States Nuclear Regulatory Commission's Standard Problem 1 (Edwards-O'Brien Blowdown Test). The RELAP5-3D standard installation problem based on the Edwards-O'Brien Blowdown Test was modified to model the appropriate initial conditions and to represent the proper location of the instruments present in the experiment. The results obtained using the modified model are significantly different from the original calculation indicating the need to model accurately the experimental conditions if an accurate assessment of the calculational model is to be obtained

  12. A theoretical study of cyclon-effect in PWR downcomer during a LOCA blowdown

    Energy Technology Data Exchange (ETDEWEB)

    Arias, F.J. [Department of Physics and Nuclear Engineering, Technical University of Catalonia, (UPC) (Spain)], E-mail: frariasm7@fis.ub.edu; Reventos, F. [Department of Physics and Nuclear Engineering, Technical University of Catalonia, (UPC) (Spain); Institute of Energy Technologies, Technical University of Catalonia, (UPC) (Spain)

    2009-08-15

    The current strategy based on the use of so-called cross-junction, allows partial modeling during blowdown episodes driven by larger-scale (flow features such as helical profile) in downcomers nuclear reactors. However a subtle but significant effect may appear by the combined action of two factors: on the one hand high azimuthal flow, on the other hand the intrinsic curvature of downcomer, and additionally in presence of a two-phases (vapor-liquid) a cyclon-effect can manifest. The present paper is a theoretical analysis of a possible cyclon-effect during blowdown episodes that allows a qualitative estimate of the impact on the calculations.

  13. Assessment of TRAC-PF1 and RELAP5/MOD1 codes with GE large-vessel blowdown test

    International Nuclear Information System (INIS)

    The GE large vessel blowdown Test No. 5801-15 was simulated with the TRAC-PF1 (Version 7.0) and RELAP5/MOD1 (Cycle 14) codes. The test facility consisted of a pressure vessel, 49-in in diameter by 14-ft long, a 2.5-in diameter converging-diverging nozzle and a blowdown line connected to the center of the upper part of the vessel (elevation from the bottom of the vessel 10.5 ft). The vessel was filled with saturated water up to 5.5 ft at 1060 psia. The test was initiated by rupturing a disc attached at the end of the nozzle. The purpose of this experiment was to study blowdown phenomena such as critical blowdown flow and the level swell during blowdown from a partially water filled vessel. Understanding of these phenomena is essential for the analysis of Loss-of-Coolant (LOCA) and steam generator steam line break accidents

  14. Simulation of blowdown experiments with the TRAC-PD2 code

    International Nuclear Information System (INIS)

    The experiments CANON and EDWARD'S PIPE were intended to simulate the blowdown phase of a typical PWR loss-of-coolant accident by depressurizing horizontal tubes filled with water at different pressures and temperatures. In this work the computer code TRAC-PD2 was employed to model those experiments. The code results are in good agreement with the experimental data. (author)

  15. Simulation of the blowdown experiencies using the TRAC-PD2 computer code

    International Nuclear Information System (INIS)

    The experiments CANON and EDWARD'S PIPE were intented to simulate the blowdown phase of a typical PWR loss-of-coolant accident by depressurizing horizontal tubes filled with water at different pressures and temperatures. In this work the computer code TRAC-PD2 was employed to model those experiments. The code results are in good agreement with the experimental data. (Author)

  16. 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

  17. Critical air/water blow-down in safety valves at low qualities

    International Nuclear Information System (INIS)

    Critical air/water blow-downs in safety valves for qualities from 0.01 to 0.113 and mass flow rates from 1.5 up to 4.3 kg/s have been observed in our test facility. These critical blow-downs are characterized by a large void fraction and by an intense mixing of the phases both in the valve body and in the outlet pipe. A qualitative estimation of the flow pattern in the outlet pipe using the map of Taitel and Dukler suggests that these air/water flows are intermittent flows - presumably slug flows - evolving to annular flows for qualities above 0.1. Intermittent flows are also predicted for critical air/water and air/glycerine flows taken from the literature for the same safety valve at slightly larger relieving pressures.

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

    International Nuclear Information System (INIS)

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

  19. MELCOR 1.8.3 assessment: GE large vessel blowdown and level swell experiments

    International Nuclear Information System (INIS)

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a series of blowdown tests performed in the early 1980s at General Electric. The GE large vessel blowdown and level swell experiments are a set of primary system thermal/hydraulic separate effects tests studying the level swell phenomenon for BWR transients and LOCAS; analysis of these GE tests is intended to validate the new implicit bubble separation algorithm added since the release of MELCOR 1.8.2. Basecase MELCOR results are compared to test data, and a number of sensitivity studies on input modelling parameters and options have been done. MELCOR results for these experiments also are compared to MAAP and TRAC-B qualification analyses for the same tests. Time-step and machine-dependency calculations were done to identify whether any numeric effects exist in our GE large vessel blowdown and level swell assessment analyses

  20. RELAP4/MOD6 (blowdown) sensitivity study on LOFT L2-2 experiment

    International Nuclear Information System (INIS)

    LOFT L2-2 experiment is the first nuclear LOCE (Loss of Coolant Experiment) of large break. In the test, the core was widely rewetted during the early blowdown, which caused the depletion by approximate 60% of the stored energy in the fuel. The early rewet had not, however, been predicted in the pretest analysis performed with RELAP4/MOD6 computer code in INEL (Idaho National Engineering Laboratory). Therefore, in the present post test analysis for the L2-2 test, the focus was laid to clarify the important points of the analytical method with RELAP4/MOD6. Calculation results with two different system nodings, i.e. noding developed in INEL and in JAERI, were compared. Calculation results with different three nodings of downcomer were also compared. Sensitivity study was made on the effectiveness to the cladding temperature by various heat transfer correlations involved in the code. Through these studies, it was found that the flow resistances in the core and the downcomer is sensitive to the core inlet flow during early blowdown, while the effectiveness to it by the downcomer noding is small. It was also shown that the gap conductance model is mostly sensitive to cladding temperature during early blowdown. (author)

  1. 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.

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

    International Nuclear Information System (INIS)

    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.)

  3. 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

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

    International Nuclear Information System (INIS)

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

  5. 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

  6. 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

  7. Severe-fuel-damage experiments in the Canadian in-reactor Blowdown Test Facility

    International Nuclear Information System (INIS)

    The Blowdown Test Facility consists of an instrumented in-reactor irradiation site plus an out-reactor piping system. These are used to irradiate CANDU fuel under conditions representative of a loss-of-coolant accident (LOCA) or LOCA with Loss-Of-Emergency-Core-Cooling (LOECC) in order to study fuel performance, fission-product release from the fuel and the transport of fission products through the piping system. An overview of the facility and the experimental program is given in this paper. (author)

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

    International Nuclear Information System (INIS)

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

  9. Mixture levels models for the Toshiba and General Electric blowdown experiment analysis

    International Nuclear Information System (INIS)

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

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-02-14

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

  12. 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)

  13. 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.

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

    International Nuclear Information System (INIS)

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

  15. Pipe loads in the case of a safety valve blowdown with phase change

    International Nuclear Information System (INIS)

    For design of piping systems the dynamical response of the structure due to flow induced forces under normal and accident conditions has to be known. As an example in the present paper we discuss the safety- and relief valve systems of a KWU pressurized water reactor that functionally protects the primary reactorsystem against overpressure. This facility consists essentially of the pressurizer and the main blowdown pipes to the safety- and relief valves. The transient loadings for this piping system emerge from flow induced forces during valve operation under normal conditions (steam discharge) and under accident conditions (subcooled water and 2-phase mixture discharge). The latter situation may occur in a few cases of anticipated transients without reactor shutdown (ATWS). (orig./GL)

  16. 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.)

  17. 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)

  18. 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.

  19. 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

  20. A blowdown analysis on LPWR LOCA by ALARM-P1

    International Nuclear Information System (INIS)

    Presented in this report is the results of an LPWR LOCA blowdown analysis by the ALARM-P1 and their comparisons with the RELAP4-EM. Up to the present, the ALARM-P1 code has been improved and refined by solving a various type of calculational exercises given as the CSNI Standard Problems. As a result, confidence of the analytical models in it was proved to be sufficient through the international comparison. Based on such experiences accumulated, therefore, the analysis of a typical PWR plant was attempted here. The results of two codes agreed fairly well, thus showing that the ALARM-P1 could be applicable to actual power plants. With the capabilities as successfully demonstrated herein, this report concludes the development work of the ALARM-P1. (author)

  1. Fluid-structure-interaction of the pressurized water reactor core internals during blowdown - numerical simulation with a homogenization model

    International Nuclear Information System (INIS)

    A method for the numerical simulation of the Pressurized Water Reactor (PWR) core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. All these models have been implemented into the code Flux-4. For the solution of the very complex, coupled equations of motions for fluid and fuel rods an efficient numerical solution technique has been developed. With the new code-version Flux-5 the PWR-blowdown is parametically investigated. The calculated core barrel loadings are compared with Flux-4 results, simulating the core's inertia by a mass ring of HDR type. (orig.)

  2. 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.

  3. 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.)

  4. Analysis of Environmental Conditions for NPP Krsko DC Battery and Battery Charger Rooms Following SG Blowdown Processing System Line Break

    International Nuclear Information System (INIS)

    This paper describes analysis of thermal-hydraulic conditions in NEK DC Battery and Battery Charger rooms after a postulated break of the SG Blowdown Processing System (SGBD) line. The calculation was performed in frame of Equipment Qualification Parameters determination for NPP Krsko. Such break can result in release of the SG water in NEK Intermediate Building which can endanger the safety related electrical equipment. Since this break cannot happen in the mentioned rooms the presence of HVAC (heating, ventilation and air conditioning) ducts must be taken into account, i.e., the HVAC duct can potentially be damaged under the pressure conditions in the rooms where the SGBD break can occur leaving the free path to the DC Battery and Battery Charger rooms. This analysis covers mass and energy release calculation (MER) due to SG blowdown line break (A), calculation of ventilation duct structural integrity (B), and the pressure and temperature response of selected IB rooms (C). A. The SGBD break is postulated in the Intermediate Building downstream the Containment Isolation valves and upstream of the Blowdown Heat Exchanger to conservatively predict the mass and energy release. This calculation was performed with RELAP5/MOD3.2.2 computer code. B. FEM (Finite Element Method) stress calculation using NISA II and ALGOR computer programs was performed for representative section of the HVAC duct in order to estimate pressure difference which can open the duct and connect battery rooms with the rooms where SG blowdown break was simulated. C. In order to take into account influence of possible damage of HVAC duct on TH conditions in IB rooms 011, 012, 014 and 015 (DC Battery and Battery Charger rooms) the GOTHIC code was used considering SGBD break in rooms IB009 and IB010. (author)

  5. RELAP4 pre-test predictions for the LOFT transient (blowdown) DNB tests in the Columbia University test loop

    International Nuclear Information System (INIS)

    Rod bundles simulating the LOFT Core-1, both with and without rod external thermocouple simulators, will be tested to determine the effect of rod external thermocouples on time-to-DNB under blowdown conditions similar to those in LOFT. Pre-test predictions have been made using the RELAP4 computer code. The purposes of this analysis were (1) to predict blowdown orifice sizes which result in the closest simulation of coolant pressure, quality, and flow rate between the test section and the LOFT core for a LOFT 200 percent simulated cold leg break at a peak linear heat generation rate of 19 kw/ft, (2) to determine ranges for instrumentation, and (3) to estimate the time-to-DNB in the rod bundles. An exact simulation of the LOFT blowdown conditions, however, can not be obtained in the test section because the rod bundles have a uniform axial power profile and the Columbia test loop is not scaled to the LOFT configuration

  6. Blowdown hydraulic influence on core thermal response in LOFT nuclear experiment L2-3

    International Nuclear Information System (INIS)

    Experimental research into pressurized water reactor (PWR) loss-of-coolant phenomena conducted in the Loss-of-Fluid Test (LOFT) facility has given results indicating that for very large pipe breaks the core thermal response is tightly coupled to the fluid hydraulic phenomena during the blowdown phase of the loss-of-coolant transient. This summary presents and discusses data supporting this conclusion. LOFT Loss-of-Coolant Experiment (LOCE) L2-3 simulated a complete double-ended offset shear break of a primary coolant reactor vessel inlet pipe in a commercial PWR. The LOFT system conditions at experiment initiation were: fuel rod maximum linear heat generation rate (MLHGR) of 39.4 +- 3 kW/m, hot leg temperature of 593 +- 3 K, core ΔT of 32.2 +- 4 K, system pressure of 15.06 +- 0.03 MPa, and flow rate/system volume of 25.6 +- 0.8 kg/m3. These conditions are typical of those in commercial PWR systems at normal operating conditions

  7. 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)

  8. Integrity of the blowdown piping systems in a nuclear power plant - findings and consequences

    International Nuclear Information System (INIS)

    In a pressurized-water reactor (PWR) the blowdown pipes from the four steam generators (material 1.5415) were examined. All the quality reports of the piping materials and the welding seams as strength and toughness and all non-destructive test findings of the welding seams were listed. According to pipe design drawings, a static calculation of the system under operational loads was conducted. In some cross sections adjacent to elbows an overload was estimated. Selective non-destructive testing of affected sections and inspection of the existing pipe-support conditions were recommended. Non-destructive tests during the inspection in 1994 revealed some longitudinally oriented findings in the base material in the inside walls of two elbows; the two elbows were replaced. The causes of findings were 'in-plane bending overloading', corrosion processes in the two elbows with findings and, most probably, condensation hammers during the startup period of the plane (possibly the opening of the shutoff valves was too fast) which yielded a large displacement of the piping system. Based on findings, the following should be implemented on the 1996 inspection: (1) the two supports U 27 and U 28 are to be changed; (2) the six highly loaded elbows are to be replaced (wall-thickness is to be increased from 5.6 to 6.3 mm); (3) the whole pipework from the four steam generators to the protection barrier is to be replaced. (orig.)

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

    International Nuclear Information System (INIS)

    The main function of the Steam Generator Blowdown System (SGBS) is to purify the secondary fluid from all kinds of pollutions: corrosion products from the secondary system, consequences of raw water pollutions through condenser's leakage, potential radiochemical pollutions resulting from Primary-to-Secondary leaks. The topic of this paper is to present the main SGBS dysfunctions linked to the degradation of the tubular heat exchangers, which sometimes can lead to integrity failure, through corrosion phenomenon. The degradation mechanisms have been characterized by various visual inspections and destructive examinations performed on pulled tubes and bundles. It appears that SGBS tubes suffer two main forms of corrosion. First, for the non-regenerative heat exchangers, where external surface of tubes is exposed to intermediate fluid, alkaline corrosion under tube sheet or shell-side baffles may occur. Caustic attack results from Na3PO4 decomposition by thermal or chemical process. Secondly, mainly for regenerative heat exchangers, pitting and under-deposits corrosion linked to lay-up conditions during outages. This kind of attack is the root cause of a potential 'domino effect': a steam jet from the leaking tube can induce mechanical and/or erosion on many tubes located in its vicinity. Concerning the external degradation by caustic corrosion, only design modifications and strong monitoring of the raw water inlet may able to limit the occurrence of tube perforation. The lay-up guidelines should be carefully followed to mitigate internal corrosion: a controlled atmosphere (limited humidity) and cleanliness of the tube (avoiding deposits formation on the bottom line) seem to be the main parameters

  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. An analysis on pressure and level change of Safety Injection Tank (SIT) during blowdown test for Yonggwang nuclear power plants units 5 and 6

    International Nuclear Information System (INIS)

    To meet the requirements of regulatory body, Safety Injection Tank (SIT) blowdown test is performed during plant startup test period to verify the performance of the tank. The test is sequentially conducted on each of four tanks and the pressure and level of SIT are measured to verify that the acceptance criteria are met. From the analysis of Yonggwang Units 5 and 6(YGN 5 and 6) test results, it is found that the polytropic index of the pressurized nitrogen gas expansion is 1.26. The TURTLE code, which has been modified in isolation valve model, simulates YGN 5 and 6 blowdown process adequately. Therefore, the modified TURTLE code will be used to generate a test acceptance criterion and evaluate test results for SIT blowdown tests

  12. 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

  13. 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

  14. The operation experience of SG blowdown demineralizers system in Korea's NPP and the best operating method in PWR

    International Nuclear Information System (INIS)

    The steam generator blowdown system(SGBDS)'s are installed to remove the concentrated impurities that are entered with the feedwater or the condensate polishing demineralizer's leakages, to minimize the effluent of radioactive materials in case of the SG's tube rupture and to maintain the lebel of ALARA's guideline in the 10CFR50 App.I. In Korea's NPPs, the SGBDS's design structure and operating procedures are various types although the installing purposes are same. The different designs are the each vessel volume, in series or not, and the resins(cation/anion) mixing ratio, etc. By the analysis results, the best method is the resins mixing ratio is determined by the impurities ratio and the new resin vessel is located after the other vessel to increase the chemical or radioactive material's decontamination factor

  15. The use of EDI to reduce the ammonia concentration in steam generators blowdown of PWR nuclear power plants

    International Nuclear Information System (INIS)

    To be recycled, PWR steam generator blowdown must be purified by mechanical filters, followed by ion exchangers (mixed bed preceded by a cationic ion exchange resin). The cationic ion exchange resin eliminates the conditioning agent ammonia in order to lengthen the cycles of the mixed bed. In the Doel nuclear power plant, Laborelec performed tests on a pilot plant for continuous electrodeionization that might replace the cation exchanger. The test campaign lasted six months. It is concluded that ammonia is removed well (1,000 μg/kg in the feed vs. 3 - 4 μg/kg in the product). The electrodeionization removes also other impurities; the conductivity of the treated water amounts to nearly 0.07 μs/cm

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

    International Nuclear Information System (INIS)

    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)

  17. 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)

  18. 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。

  19. 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

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

    International Nuclear Information System (INIS)

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

  1. Mass flow rate efflux resulting from the transient blowdown of two-phase fluids in a pipe connected to a large vessel

    International Nuclear Information System (INIS)

    The transient blowdown of two-phase fluids in a pipe connected to a large vessel is determined for both single component two-phase and two component, liquid dominant, two-phase fluids. A simple behavioral model is used to characterize these fluids and provides for the analytical evaluation of certain features of the decompression process. This paper also includes for purposes of comparison the blowdown solutions for gas filled systems. Emphasis is placed upon the transient mass flow rate efflux under the conditions of choked outflow, that is, for low backpressure. The vessel is assumed to be large enough that it can be treated as a constant state source. Adiabatic flow is assumed, and solutions with and without pipe friction effects are developed. The analysis is based upon a one-dimensional treatment, and homogeneous, thermodynamic equilibrium, no-slip flow is assumed for the two-phase fluid systems. A correlation is made between the friction and frictionless cases as a function of a length-to-diameter ratio parameter of the blowdown pipe. The results are applicable to the decompression of tubes in process system heat exchangers due to sudden tube failure and to a broad class of accidental releases of hazardous materials

  2. FIX-II. Loca-blowdown heat transfer and pump trip experiments. Summary report of phase 1: Design of experiments

    International Nuclear Information System (INIS)

    FIX-II is a loss of coolant blowdown heat transfer experiment, performed under contract for The Swedish Nuclear Power Inspectorate, SKI. The purpose of the experiments is to provide measurements from simulations of a pipe rupture on an external recirculation line in a Swedish BWR. Pump trips in BWRs with internal recirculation pumps will also be simulated. The existing FIX-loop at the Thermal Engineering Laboratory of Studsvik Energiteknik AB will be modified and used for the experiments. Components are included to simulate the steam dome, downcomer, two recirculation lines with one pump each, lower plenum, core (36-rod full length bundle), control rod guide tubes, core bypass, upper plenum and steam separators. The results of the first phase of the project are reported here. The following tasks are included in Phase 1: reactor reference analysis, scaling calculations of the FIX loop, development of fuel rod simulators, design of test section and test loop layout and proposal for test program. Further details of the work and results obtained for the different sub-projects are published in a number ofdetailed reports. (author)

  3. 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)

  4. 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

  5. UPTF/TEST5A/RUN063, Steam/Water Flow Phenom.Blowdown PWR Cold Leg Break LOCA

    International Nuclear Information System (INIS)

    1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: This separate effects test was performed to investigate the steam/water flow phenomena in the

  6. Statistical analysis of the blowdown phase of a loss-of-coolant accident in a pressurized water reactor as calculated by RELAP4/MOD6

    International Nuclear Information System (INIS)

    A statistical study is presented of the blowdown phase of a design basis accident (double-ended cold leg guillotine break) in the Zion pressurized water reactor. The response surface method was employed to generate a polynomial approximation of the peak clad temperatures calculated by RELAP4/MOD6. The nodalization was a modification of the RELAP model of Zion developed in the BE/EM study. Twenty one variables were initially selected for the study. These variables, their ranges and distributions resulted from the best engineering judgement of NRC, Sandia, INFL, and other interested and knowledgeable investigators

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  9. 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

  10. Developmental measurements of the optimum surface blowdown location, of temperature fields and strain in selected nodes of 1000 MW steam generator and pressurizer

    International Nuclear Information System (INIS)

    The reasons for and goals of developmental measurements on a WWER-1000 steam generator and pressurizer are discussed. For blowdown measurements, the steam generator of unit 1 of the Temelin nuclear power plant is equipped with 28 probes, out of which 26 are located under the water level whereas two probes are located before and after the separating equipment for separation efficiency evaluation. The concentrations of Na+, Fe2+, Fe3+ and Cl- ions and oxygen and the electrolytic conductance will be measured in water samples. At exposed sites of the steam generator and pressurizer, the operation load will be established by temperature and stress measurements. Temperatures will be measured with jacketed thermocouples and mechanical stress, with high-temperature tensometers. Active cracks in the lid-collector joint bolts and the state of the upper part of the collector, of the lid and of the dismantable joint will be monitored with acoustic emission sensors. (Z.M.). 2 figs., 4 refs

  11. 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

  12. Comparison of the RELAP4/MOD3 and RELAP4/MOD5 results to the loss-of-coolant accident blowdown phase simulation on the Angra-1 nuclear power station

    International Nuclear Information System (INIS)

    It's very important to obtain information of certain thermalhydraulic parameters of the Angra 1 Nuclear Power Plant during a large LOCA in three different points: in the cold leg, in the hot leg and between the steam generator and the pump. In this way, these paper describes the results comparison of these accidents with the RELAP4/MOD3 and RELAP4/MOD5 codes with the results in the Final Safety Analysis Report of Angra 1. It was used a 36 control volume model, with 48 junctions between the control volumes, 25 heat exchange structures and 8 valves. The transient analysis were made just during the blowdown phase of the LOCA. (author)

  13. 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

  14. 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.)

  15. 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.)

  16. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  18. 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)

  19. 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)

  20. Measurement of blowdown flow rates using load cells

    International Nuclear Information System (INIS)

    To establish a reliable method for measuring two-phase flow, experiments were planned for measurement of transient single phase flow rates from vessels using load cells. Suitability of lead-zirconate-titanate piezoelectric ceramic discs was examined. Discharge time constant of the disc used was low, leading to large measurement errors. Subsequently, experiments were carried out using strain gauge load cells and these were found satisfactory. The unsteady flow equation has been derived for the system under investigation. The equation has been solved numerically using the fourth order Runge-Kutta method and also by integrating it analytically. The experimental results are compared with the theoretical results and presented in this report. (auth.)

  1. Analytical modeling of bwr safety relief valve blowdown phenomenon

    International Nuclear Information System (INIS)

    An analytical, qualitative understanding of the pool pressures measured during safety relief valve discharge in boiling water reactors equipped with X-quenchers has been developed and compared to experimental data. A pressure trace typically consists of a brief 25-35 Hz. oscillation followed by longer 5-15 Hz. oscillation. In order to explain the pressure response, a discharge line vent clearing model has been coupled with a Rayleigh bubble dynamic model. The local conditions inside the safety relief valve discharge lines and inside of the X-quencher were simulated successfully with RELAP5. The simulation allows one to associate the peak pressure inside the quencher arm with the onset of air discharge into the suppression pool. Using the pressure and thermodynamic quality at quencher exit of RELAP5 calculation as input, a Rayleigh model of pool bubble dynamics has successfully explained both the higher and lower frequency pressure oscillations. The higher frequency oscillations are characteristic of an air bubble emanating from a single row of quencher holes. The lower frequency pressure oscillations are characteristic of a larger air bubble containing all the air expelled from one side of an X-quencher arm

  2. Subsonic Constant-Area MHD Generator Experiments with the CNEN Blow-Down Loop Facility

    International Nuclear Information System (INIS)

    The design of the facility, described at the Salzburg Symposium, was somewhat modified following the results of the commissioning tests; the changes were mainly concerned with the thermal insulation, duct materials and caesium recovery system. The facility went into full operation in March 1967 and since then two series of MHD experiments, a total of twenty-six runs, have been performed. During the MHD runs the facility has been working mostly under the following operating conditions: stagnation temperature 1500 to 1800°K; stagnation pressure-1 to 3 atm. abs.; mass How 50 to 150 g/sec; seeding 2 to 5 at.%- ; magnetic field 0 to 45 k G; Mach number 0.4 to 0.8; Hall parameter up to 6. The main purpose of the experiments was to study the performance of relatively small generators (cross-section 3 x 5 cm2, length 8-20 cm) both when the non-equilibrium ionization is expected to be negligible and when it should be, in a very idealized model, relevant. As a first step, efforts were made to ascertain whether any of the unsatisfactory results reported in Salzburg, both for equilibrium and non-equilibrium generators, stemmed not from the basic functioning principle of an MHD small-scale generator but rather from some inadequacy of the experimental apparatus. Therefore particular attention was paid to: ceasium vaporization and mixing with helium; plasma insulation from ground; electrical insulation from ground and from each other of those electrically conductive parts of the facility which may, during the functioning, come into contact with the plasma; temperature control of the duct; purity level; duct materials; measurement system and control. In the equilibrium regime the Faraday field measured is very close to the ideal value and it reaches 80 V/cm (400 volts between electrodes); the Hall field still remains below the ideal value uBβL (50% at β = 3). The maximum Hall field was about 35 V/cm for a corresponding voltage of 600 V. Preionization may especially influence the Hall field, confirming the presence of some axial leakages; some artificial changes in the gound-loop resistance have shown experimentally how drastically the Hall field may be reduced if a less than very careful approach is taken to the problems of electrical insulation. Current densities of a few amperes per square centimetre at the electrodes were measured and the electrical power generated in the duct was of the order of few watts per cubic centimetre. (author)

  3. 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)

  4. 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)

  5. The influence of back pressure on operation and blowdown of safety relief valves

    International Nuclear Information System (INIS)

    The article acquaints such researching procedure by the help of which the operating method of safety relief valve at back pressure is analysable and evaluable. There is a presentation of a new measuring apparatus supplied with electronic data processing device and computer on which at the safety relief valves the influence of the built-up and superimposed back pressure is examinable. The authors inform in the case of a selected valve type the correlations between the valve lift and the pressures acting on the valve. They analyse the relation between the flow force acting on the valve disc and the spring force. They introduce the relations among the discharge coefficient, the valve lift, and the back pressure. (orig.)

  6. 46 CFR 162.018-5 - Blow-down adjustment and popping tolerance.

    Science.gov (United States)

    2010-10-01

    ... constructed that no shocks detrimental to the valve or pressure vessel are produced when lifting or closing. Safety relief valves shall be designed to open sharply and reach full lift and capacity at the...

  7. The blowdown, refill and reflood phase during a LOCA. Survey of the main physical phenomena

    International Nuclear Information System (INIS)

    In this paper, the main physical phenomena occuring during a LOCA are reviewed. They are presented in a chronological order. For each phenomena, a detailed physical description is given followed by the review of the general modelling problems. For some of these phenomena, modelling details are given for critical flow, for two-phase flow and heat transfer, for critical heat flux and post critical heat flux heat transfer, for reflood and rewet heat transfer and in the survey on LOCA computation codes

  8. 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)

  9. 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)

  10. 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)

  11. 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)

  12. Progress in the simulation of transient events observed during the blowdown of a medium-scale pressurized-water loop

    International Nuclear Information System (INIS)

    The computer code RAMA has been developed to analyze postulated loss-of-coolant accident (LOCA) scenarios for a CANDU-PHW+ reactor circuit. The reactor circuit consists of a branched pipe network, with components such as fuel channels, steam generators, headers and pumps at discrete locations. The horizontal fuel channels distinguish this reactor from all other commercial reactors. As part of the code validation process, comparisons of predictions from the model are made with experiments conducted in the RD-12 loop, a facility that duplicates the important geometrical features of the reactor system. This paper focuses on two experiments conducted in RD-12 that illustrate the modelling requirements for different circuit flow conditions. Where high flow rates prevail, predictions of a homogeneous equilibrium model are in good agreement with experiment. At lower flow rates, the steam and water separate and may have different velocities and temperatures. Greatly improved predictions can be obtained for such conditions by using an unequal velocity, unequal-temperature (UVUT) model

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

    International Nuclear Information System (INIS)

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

  14. A new design of blowdown sampling and of an assembly for heat transfer tube surveillance specimens as a way to extending the lifetime of WWER-440 units

    International Nuclear Information System (INIS)

    A total of 16 boiler water sampling probes were installed at two steam generators of unit 4 of the Dukovany nuclear power plant. The blowing-down was modified by means of an internal structure so that it proceeds from areas with the highest concentrations of impurities. An asset of this approach is the feasibility of mounting the structure into operated steam generators. A surveillance specimen programme was developed to detect primary circuit leaks. Heat transfer tube surveillance specimens are approximately 600 mm long, are filled with water or air and pressure sealed at the two ends. The tube specimens are fastened in a special assembly and suspended on a support system in the steam generator. The specimens can be removed at a preselected time during scheduled outage without disturbing the compactness of the tube beam. (M.D.). 2 figs., 5 refs

  15. SCRELA, LOCA Analysis of Super-Critical Light-Water Reactors

    International Nuclear Information System (INIS)

    Description of program or function: LOCA Analysis Code for the Supercritical-Water Cooled Reactor. - Blowdown Module: Calculation of the Blowdown Phase and Refill Phase. - Reflood Module: Calculation of the Reflood Phase

  16. Dynamic loads of the RPV internals and primary coolant circuits of a pressurized water reactor for the case of a blowdown accident. DAISY computations for a 2A-break in the cold leg

    International Nuclear Information System (INIS)

    For the first time the complete 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. 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 influence of the simplifications and the assumptions are reviewed. The most important finding of this investigations in the fact that as consequence of a postulated 1A-break in the cold leg of a PWR reactor there is no risk for the structural integrity of the RPV internals. (orig./HP)

  17. 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.

  18. 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)

  19. 75 FR 64303 - Tennessee Gas Pipeline Company; Notice of Intent To Prepare an Environmental Assessment for the...

    Science.gov (United States)

    2010-10-19

    ... inlet gas filter-separator, a blowdown silencer, and a relief valve would be installed and unit piping... inlet gas filter-separator, a blowdown silencer, and a relief valve would also be installed. Compressor... relief valve would also be installed. Compressor Station 325--An inlet gas filter-separator, a...

  20. 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...

  1. 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.

  2. 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...

  3. Performance test of filtering system for controlling the turbidity of secondary cooling water in HANARO

    International Nuclear Information System (INIS)

    There is about 80 m3/h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30 MW research reactor. When the secondary cooling water is treated by high Ca-hardness treatment program for minimizing the blowdown loss, only the trubidity exceeds the limit. By adding filtering system it was confirned, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2 % of filtering rate without blowdown. And it was verified, through the field performace test of filtering system under normal operation condition, that the circulation pumps get proper capacity and that filter units reduce the turbidity below the limit. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  4. Short-term calculations to supplement the RS 16 B PWR experiments with internals (PWR1 to PWR5), using the LECK 4 computer code

    International Nuclear Information System (INIS)

    Within the framework of research project RS 16 B sponsored by the German BMFT a series of a blowdown experiments, DWR1 to DWR5, were performed using a vessel with dummy internals under conditions similar to those in a PWR. The prime objective of these experiments was the investigation of the highly transient blowdown phenomena in the discharge nozzle and the determination of the induced loads on the internals. As a partner in the project, KWU carried out both pre-test predictions and post-test analyses of these experiments using, among others, the computer code LECK 4. For the most severe blowdown test DWR5, the influence of the most important model parameters on the blowdown analysis was investigated in detail. These investigations suggest that, similar to the long-term analyses, calculations using the homogeneous critical flow model would improve agreement between calculation and experiment. (orig./RW)

  5. 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,...

  6. Secondary circuit water chemistry and related problems with SG

    International Nuclear Information System (INIS)

    Necessity for SG feed water and blowdown systems modernization Balakovo NPP steam generators PGV-1000M was identified at Units with VVER-1000 during commissioning separational, thermo-hydraulic and thermo-chemical testings. It was discovered, that in zone of 'hot' header coolant salt concentration (concentration of dissolved salts) was almost 2 times more, than salt concentration in blowdown water. A number of chemical testings was performed to investigate and optimize salts distribution in water volume of PGV-1000. (R.P.)

  7. Identification and analysis of noise sources in the Noor-Abad gas compressor station, Iran

    Directory of Open Access Journals (Sweden)

    Nader Mohammadi

    2015-01-01

    Full Text Available The objective of this study is to identify the sources of acoustic noise (noise pollution in the Noor-Abad gas compressor station and then to prioritize the station equipment based on noise pollution. First, the key locations inside the station as well as in the surrounding residential area, aka the study area, are determined for the measurement of sound pressure level. Then, the sound pressure level is measured at those points, and the related noise map is produced. Based on the noise map, the noise condition in the study area is evaluated by comparing the measured acoustic parameters with allowable standard values. Dangerous regions and critical points are thus identified. The major noise sources consist of main blowdown, units’ blowdowns, scrubbers, and turbo-compressors. The sound pressure level of main blowdown is measured at two intervals from its position: 80 m inside the station and 600 m outside the station (at the edge of the surrounding residential area. Also, the sound pressure level for a unit blowdown and a scrubber is measured at respectively 25 m and 40 m from their positions. Finally, the station equipment is prioritized based on noise pollution. The analysis of measurement results showed that the main noise sources are, respectively, the station main blowdown, units’ scrubbers, units’ blowdowns, turbo-compressors, and gas pipelines.

  8. Flotation process for removal of precipitates from electrochemical chromate reduction unit

    Science.gov (United States)

    DeMonbrun, James R.; Schmitt, Charles R.; Williams, Everett H.

    1976-01-01

    This invention is an improved form of a conventional electrochemical process for removing hexavalent chromium or other metal-ion contaminants from cooling-tower blowdown water. In the conventional process, the contaminant is reduced and precipitated at an iron anode, thus forming a mixed precipitate of iron and chromium hydroxides, while hydrogen being evolved copiously at a cathode is vented from the electrochemical cell. In the conventional process, subsequent separation of the fine precipitate has proved to be difficult and inefficient. In accordance with this invention, the electrochemical operation is conducted in a novel manner permitting a much more efficient and less expensive precipitate-recovery operation. That is, the electrochemical operation is conducted under an evolved-hydrogen partial pressure exceeding atmospheric pressure. As a result, most of the evolved hydrogen is entrained as bubbles in the blowdown in the cell. The resulting hydrogen-rich blowdown is introduced to a vented chamber, where the entrained hydrogen combines with the precipitate to form a froth which can be separated by conventional techniques. In addition to the hydrogen, two materials present in most blowdown act as flotation promoters for the precipitate. These are (1) air, with which the blowdown water becomes saturated in the course of normal cooling-tower operation, and (2) surfactants which commonly are added to cooling-tower recirculating-water systems to inhibit the growth of certain organisms or prevent the deposition of insoluble particulates.

  9. Analysis of Angra-1 fuel rod during the large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    The main objective of this work is to study the fuel element behavior of the Angra 1 Nuclear Reactor, during a large loss of coolant accident caused by as rupture of the cold leg. Only the blowdown phase was considered. For this study the steps discribed below were done: - analysis of the blowdown phase was performed with the computational code RELAP4/MOD5 (option EM); analysis of the hot channel during the blowdown was made using the computational code RELAP/MOD5 (option EM); analysis of the fuel element performance during the accident with the computational code FRAP-T6. The results obtained in the steps above were compared with data presented in the Angra 1 Final Safety Analysis Report. (author)

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

    International Nuclear Information System (INIS)

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

  11. Mark II containment 1/6-scale pressure suppression test program: data report no. 2

    International Nuclear Information System (INIS)

    This report documents experimental data from the first test phase of the Mark II Containment 1/6-Scale Pressure Suppression Test. The 1/6-Scale Test was initiated in December, 1976, to investigate the thermohydraulic responses of a BWR Mark II pressure suppression system to a postulated loss-of-coolant accident (LOCA), by means of scale model experiments. From January to June, 1977, a series of tests were performed for the Japanese BWR Owners' Group. These tests consisted of eight air-blowdown pool swell tests, three steam-blowdown pool swell tests, and twelve steam condensation tests. The dynamic responses of pressure and pool water level during the blowdown, pressure oscillation and chugging phenomena associated with unsteady condensation of steam were measured. (author)

  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. 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

  14. 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).

  15. Lessons learned at Calvert Cliffs

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, C.C. III (Baltimore Gas and Electric Co.); Graf, D.V.; Honey, J.A.

    1976-01-01

    Initial operation of the Calvert Cliffs-1 Reactor yielded significant reliability information in the following areas: circulating water system, primary system valve leakage, steam generator blowdown recovery, spare parts availability, breathing air systems, primary system water chemistry, secondary system water chemistry, RPS/ESFAS interactions, in-core detectors, and computer trending.

  16. 76 FR 22928 - Nextera Energy Point Beach, LLC; Point Beach Nuclear Plant, Units 1 and 2; Environmental...

    Science.gov (United States)

    2011-04-25

    ... the Federal Register on December 10, 2010 (75 FR 77010). Comments were received on the draft EA from... available online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html . From this site, the public... chemical and volume ] control system, steam generator blowdown, chemistry laboratory drains, laundry...

  17. 40 CFR 471.12 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ...) Shot-forming wet air pollution control scrubber blowdown. Subpart A—BAT Pollutant or pollutant property...). Except as provided in 40 CFR 125.30 through 125.32, any existing point source subject to this subpart... 0.030 Lead 0.010 0.005 (b) Rolling spent soap solutions. Subpart A—BAT Pollutant or...

  18. 40 CFR 471.11 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ...-forming wet air pollution control scrubber blowdown. Subpart A—BPT Pollutant or pollutant property Maximum... available (BPT). Except as provided in 40 CFR 125.30 through 125.32, any existing point source subject to... solutions. Subpart A—BPT Pollutant or pollutant property Maximum for any 1 day Maximum for monthly...

  19. 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

  20. 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...

  1. Condensate purification in PWR reactors

    International Nuclear Information System (INIS)

    The recommendations made by the VGB task group on 'condensate purification for PWR reactors' 1976 are discussed in detail. Techniques and circuiting possibilities of condensate purification for BBR steam generators (forced circulation) and KWU steam generators (U tube with blow-down) are mentioned. (HP)

  2. Annual meeting on nuclear technology 1981. Technical meeting: Structural stresses as a result of shock waves

    International Nuclear Information System (INIS)

    An examination is made of the influence of shock waves in connection with the safety considerations, the safeguarding of pipelines for dynamic loads with regard to elasticity and plasticity, the effects encountered in fluid-filled, plasticifying pipelines, the dynamic stresses placed on reactor pressure vessel internals in case of blowdown and the parts of the plant relevant to safety. (DG)

  3. 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)

  4. 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

  5. 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...

  6. BRAVO [test facility] puts PWR safety and relief valves to the test

    International Nuclear Information System (INIS)

    A new valve blowdown test facility is completing commissioning at the Marchwood Engineering Laboratories of Britain's Central Electricity Generating Board. BRAVO, as the facility is known, will prove the performance of the safety and relief valves to be used in Sizewell B, Britain's first PWR. (author)

  7. 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

  8. 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...

  9. 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.

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

    International Nuclear Information System (INIS)

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

  11. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  12. 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

  13. Environmental effects of cooling system alternatives at inland and coastal sites

    International Nuclear Information System (INIS)

    The environmental effects of alternative cooling systems for power plants in California were analyzed. At inland sites evaporative cooling systems must be used, with fresh water or waste water used as makeup. Because fresh water is scarce, most new plants would need to use agricultural or municipal waste waters. For agricultural waste water systems, disposing of the blowdown and dispersion of drift containing total dissolved solids are two significant problems requiring resolution. At coastal sites, once-through cooling systems or recirculating systems could be used. Once--through cooling causes fewer effects on the marine environment than do recirculating systems on the air and marine environment when oceans water makeup is used. In general, for a recirculating system, dispersing high-salinity blowdown in marine waters and the effects of salt water drift on the terrestrial ecology outweigh the effects of once-through warm water on marine life. (U.S.)

  14. An in-reactor loss-of-coolant test with flow blockage and rewet

    International Nuclear Information System (INIS)

    Three CANDU-type fuel elements were subjected to a blowdown transient in the Blowdown Test Facility of the NRU reactor. These elements, operating at 41 to 43 kW/m, went into sustained dryout about 22 seconds after depressurization. Sheath temperatures escalated rapidly following dryout, and subsequent damage to the elements caused a flow blockage below the fuel assembly. The high-temperature transient was terminated by automatic reactor shutdown and the initiation of cold water rewet. The first rewet water vapourized, but rewet eventually cooled the fuel. Fission products were released and measured during and following the transient. Post-irradiation examination has shown severe fuel damage at the bottom of the assembly

  15. 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.

  16. Fluid-structure-coupling algorithm

    International Nuclear Information System (INIS)

    A fluid-structure-interaction algorithm has been developed and incorporated into the two dimensional code PELE-IC. This code combines an Eulerian incompressible fluid algorithm with a Lagrangian finite element shell algorithm and incorporates the treatment of complex free surfaces. The fluid structure, and coupling algorithms have been verified by the calculation of solved problems from the literature and from air and steam blowdown experiments. The code has been used to calculate loads and structural response from air blowdown and the oscillatory condensation of steam bubbles in water suppression pools typical of boiling water reactors. The techniques developed here have been extended to three dimensions and implemented in the computer code PELE-3D

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

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. Application of the LIMIT code to the analysis of containment hydrogen transport

    International Nuclear Information System (INIS)

    The principal developmental focus of the LIMIT code is the ability to model hydrogen transport accurately in reactor containments. The program is capable of treating rapid two-phase dominated blowdown transients, slower mixing events in which diffusional transport is important, and lumped or nodal multicompartment analysis. The code's features include versatile multidimensional geometry options and models of ancillary equipment including solid heat sinks and mass and energy sources. The program is applied to a number of pertinent problems including continuum analysis of a hydrogen/water blowdown, simulation of experimental tests performed at the Battelle-Frankfurt Institute and the Hanford Engineering Development Laboratory, and lumped parameter studies of connected room problems. The code is shown to be capable of accurately treating a wide range of problems with reasonable computational efficiency. The need for even better efficiency, additional equipment submodels, and further validation are the code's principal limitations

  20. 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

  1. Thermo-hydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

    International Nuclear Information System (INIS)

    In nuclear reactors of the magnox or advanced gas cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accidents using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear electric . The tests were carried out on the thermal hydraulics experimental research assembly (THERA) loop at manchester university. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 k, 50 k, and 100 k was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. Thermal power plant wastes and their management: a case study

    Energy Technology Data Exchange (ETDEWEB)

    Ajmal, M.; Uddin, R.; Khan, M.A. (Aligarh Muslim University, Aligarh (India). Z.H. College of Engineering and Technology, Environmental Research Lab., Dept. of Applied Chemistry)

    1990-01-01

    The problem of pollution caused by coal-fired thermal power plants is still of importance because of the tremendous growth in use of electrical energy. The wastewaters of a typical coal fired power plant include cooling tower blowdown, wastewaters from ash handling system, boiler blow-down, regenerate wastes, floor and yard drains, coal pile run off, etc. The solid waste is the fly ash. The present communication deals with the physico-chemical characteristics of fly ash pond effluent and the fly ash and their effects on certain chemical properties of soil, seed germination pattern and the growth of some crop plants. The plants selected for these studies were as follows: for fly ash pond effluent - kidney bean ([ital Phaseolus aureus]) and lady's finger ([ital Abelmoschus esculentus]), whereas for the fly ash - wheat ([ital Triticum aestivum]) and pea ([ital Pisum sativum]). 20 refs., 7 tabs.

  5. Thermal power plant wastes and their management

    International Nuclear Information System (INIS)

    The problem of pollution caused by coal-fired thermal power plants is still of importance because of the tremendous growth in use of electrical energy. The wastewaters of a typical coal fired power plant include cooling tower blowdown, wastewaters from ash handling system, boiler blowdown, regenerate wastes, floor and yard drains, coal pile run off etc. The solid waste is the fly ash. The present communication deals with the physico-chemical characteristics of fly ash pond effluent and the fly ash and their effects on certain chemical properties of soil, seed germination pattern and the growth of some crop plants. The plants selected for these studies were as follows: for fly ash pond effluent - kidney bean (Phaseolus aureus) and lady's finger (Abelmoschus esculentus), whereas for the fly ash - wheat (Triticum aestivum) and pea (Pisum sativum). 7 tabs.; 20 refs. (author)

  6. Main steam system piping response under safety/relief valve opening events

    International Nuclear Information System (INIS)

    The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

  7. 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...

  8. 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.

  9. Combined effects experiments with the condensation pool test facility

    International Nuclear Information System (INIS)

    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

  10. 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

  11. 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

  12. 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...

  13. Investigation of the behaviour of main coolant pumps at LOCA conditions

    International Nuclear Information System (INIS)

    The LOCA analysis for a PWR requires a model of the primary coolant pump behaviour under single- and two-phase flow conditions. For model verification a one-quarter and a one-fifth scale model of an axial main coolant pump were tested under steady-state and transient conditions over ranges typical for a PWR-LOCA. Effects of the pump behaviour on LOCA's were studied by blowdown calculations, too. (orig.)

  14. Validation of computer code THYNAC for analysis of loss of coolant accident in pressurised heavy water reactor

    International Nuclear Information System (INIS)

    The computer code THYNAC has been validated against the available experimental results of blowdown from RD-4 loop, a Canadian test facility designed to simulate LOCA in PHWR. In this paper available experimental results are compared with predictions made with THYNAC. In general, predictions show consistent trends on pressure transient during LOCA and conservative trends with respect to fuel sheath peak temperatures. (author). 4 refs., 5 figs., 2 tab s

  15. Development of electromagnetic filtration in the feed water circuits

    International Nuclear Information System (INIS)

    Electromagnetic filtration in the feed water circuit of the steam generators in nuclear power plants is efficient towards insoluble corrosion products. The principle of electromagnetic filtration is shortly recalled and the results of corresponding development work are summarized. The magnitude of water volumes to be treated on the two priviledged parts of the circuit are estimated. These parts are on the feed water tank level and on the blow-down of the steam generator. The practical applications are discussed

  16. 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...

  17. Minimality of Symplectic Fiber Sums along Spheres

    CERN Document Server

    Dorfmeister, Josef G

    2010-01-01

    In this note we complete the discussion of minimality of symplectic fiber sums. We find, that for fiber sums along spheres the minimality of the sum is determined by the cases discussed by M. Usher and one additional case: If the sum is the result of the rational blow-down of a symplectic -4-sphere in X, then it is non-minimal if X contains a certain configuration of exceptional spheres in relation to this -4-sphere.

  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. 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

  20. 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.

  1. 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

  2. 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.)

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

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. 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

  6. 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

  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. Conceptual plan for 100-N Emergency Dump Basin (EDB) deactivation

    International Nuclear Information System (INIS)

    This document provides the conceptual plan for the 100-N Emergency Dump Basin (EDB) located at the Hanford Site in Richland, Washington. The EDB is an outdoor concrete retention pond with a carbon-steel liner underlain with fiberglass. The EDB was originally designed as a quenching pool for reactor blowdown in event of a primary coolant leak. However, the EDB only received routine steam-generator blowdowns from 1963 to 1987. The steam-generator blowdown and leaking isolation valves allowed radioactively contaminated water (from primary and secondary reactor coolant leaks) to enter the EDB. Over the years, wind-blown sand and dust have settled in the EDB, resulting in the present layer of sediments. As of February 1996, the EDB contained an estimated 260,000 gal of water and approximately 2,300 ft3 of sediment. The average sediment thickness is estimated at 2.5 ft and is covered with approximately 12 ft to 14 ft of water. Vegetation (mostly reeds and cattails) grows in the basin corners where the sediment is exposed. To minimize animal and bird intrusion, a kneeling net has been installed over the EDB

  9. Experimental data of ROSA-III integral test run 705

    International Nuclear Information System (INIS)

    Run 705 of the ROSA-III experimental program is an isothermal blowdown test without ECCS actuation in the BWR LOCA test series simulating a double-ended break on the inlet side of a recirculation pump. Purpose of the ROSA-III test program is to provide comprehensive experimental data of thermal hydraulic behavior in BWR LOCA to assess the system computer code. ROSA-III facility is a volume-scaled (1/424) system of a large (--1000MWe) BWR for an integral test on a BWR LOCA such as a 200% double-ended offset shear break on the inlet side of the pump in a recirculation loop. Power supply to main recirculation pumps and to core heater pins were stopped about 7 and 4 seconds respectively before initiation of the blowdown for an isothermal blowdown test. The primary initial conditions are steam dome pressure 7.11MPa, steam dome temperature 559K, lower plenum subcooling 9K, and core inlet flow 0.0 kg/s. The experiment in RUN 705 was successful; graphical data are presented in this report. (author)

  10. 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.)

  11. 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)

  12. Disposal of sediments from the 1300-N Emergency Dump Basin

    International Nuclear Information System (INIS)

    This report describes the characterization of the 1300-N Emergency Dump Basin (EDB) sediments, summarizes the data obtained, the resultant waste categorization, and the preferred disposal method. The EDB is an outdoor, concrete storage pond with a 3/16-in. carbon steel liner. The basin (completed in 1963) originally served as a quenching pool for reactor blowdown in the event of a primary coolant leak. Later, the basin received blowdown from the N Reactor steam generators. The steam generator blowdowns and leading isolation valves allowed radioactively contaminated water (from primary and secondary reactor coolant leaks) to enter the basin. Windblown dust and sand have settled in the basin over the years (because of its outdoor location), causing the present layer of sediments. To minimize potential airborne contamination, the water level was kept constant by adding water. However, the addition of water was stopped to minimize the amount of contaminated water needing disposal. To ensure that the surfaces exposed as a result of evaporation pose no immediate airborne contaminant problem, the contamination levels are monitored by Radiation Control Technicians. As part of the deactivation of N Reactor facilities, the EDB will be stabilized for long-term surveillance and maintenance prior to final decontamination and demolition

  13. 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

  14. 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)

  15. 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

  16. 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

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

    International Nuclear Information System (INIS)

    Corrosion products that make their way to the secondary side of pressurized water reactor (PWR) steam generators (SGs) via the feedwater can deposit on the SG tubes. These deposits can form an occluded region which inhibits heat transfer, leads to thermal hydraulic instabilities through blockage of tube supports and creates regions where corrosive species can concentrate along tubes and tube to tube support plate crevices. The performance of the SG is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. A promising new method for significantly reducing corrosion product deposition on the secondary side of recirculating steam generators is the use of online dispersant addition to help prevent the corrosion products from adhering to the steam generator surfaces. By inhibiting the deposition of the corrosion products, they are more effectively removed from the steam generator via blowdown. After completion of a significant and comprehensive qualification program, a short-term dispersant trial was performed at Arkansas Nuclear One Unit 2 (ANO-2) in Winter/Spring 2000, lasting approximately 3 months. A high purity, high molecular weight polyacrylic acid (PAA) dispersant produced by Betz-Dearborn was injected at low concentrations (0.5 μg x kg-1 to 12 μg x kg-1) into the final feedwater. The blowdown iron removal efficiency was observed to increase by an order to magnitude and more with use of PAA. Normal chemistry parameters, such as blowdown cation conductivity and TOC/TIC, were unaffected by PAA application. The results and conclusions from the trial are presented and discussed. (orig.)

  18. 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)

  19. Asbestos in cooling-tower waters

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.A.G.

    1977-12-01

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

  20. A Small-Scale Capsule Test for Investigating the Sodium-Carbon Dioxide Reaction

    International Nuclear Information System (INIS)

    The utilization of modular sodium-to-supercritical CO2 heat exchangers may yield significant improvements for an overall plant energy utilization. The consequences of a failure of the sodium CO2 heat exchanger boundary, however, would involve the blowdown and intermixing of high-pressure CO2 in a sodium pool, causing a pressurization which may threaten the structural integrity of the heat exchanger. Available data seems to indicate that the chemical reaction between sodium and CO2 would likely produce sodium oxides, sodium carbonate, carbon and carbon monoxide. Information on the kinetics of the sodium-CO2 reaction is virtually non-existent

  1. 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

  2. Short presentation of the activities of the Joint Research Center, Ispra establishment in the field of material research in reactor safety

    International Nuclear Information System (INIS)

    The Commission of the European Communities (CEC) disposes of a joint Research Center (JRC) composed of four establishments. In the ISPRA establishment, which is the largest of four, the largest project, Reactor Safety, includes the following: reliability analysis; blowdown; sodium thermohydraulics; fuel-coolant interaction and post accident heat removal; dynamic structural loading and response (LMFBR); structural failure prevention. The last is described in this paper. It deals with: code validation program for primary containment response in a LMFBR following core disruptive accident (COVA); dynamic material testing; fracture mechanics; creep fatigue; creep crack growth; creep damage evaluation; non-destructive testing

  3. 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.

  4. Dynamic analysis of piping systems - an approach to the experimental verification of computer programs (beam models)

    International Nuclear Information System (INIS)

    The structural analysis of dynamically excited piping systems as performed with the numerical method of Finite Elements (FE) is usually based on simple beam elements. However, in order to establish their applicability, the computer programs used for the structural analysis must be verified on the basis of large-scale experiments. The behaviour of the extremely endangered elbows is considered (blowdown accident in a NPP). The influences of form effects (cross-section ovalization) in the elbows are isolated from the measured total stresses by a new method and their importance for a comparison experiment and calculation is shown. (orig./HP)

  5. 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.)

  6. 1993 RCRA Part B permit renewal application, Savannah River Site: Volume 10, Consolidated Incineration Facility, Section C, Revision 1

    International Nuclear Information System (INIS)

    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

  7. 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)

  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. Containment atmospheric response (CAR) scoping study

    International Nuclear Information System (INIS)

    The primary objectives of the Containment Atmospheric Response (CAR) subtask are to identify and evaluate thermodynamic phenomena occurring in the containment atmosphere during postulated gas-cooled reactor depressurization events that may affect the release of fission products to the environment. The release of fission products to the environment can be altered if the containment function fails, if the amount of plateout in the containment is changed or if the performance of the containment atmospheric cleanup system is impaired. A knowledge of the blowdown phenomena can also provide guidelines for locating reactor shutdown equipment and establishing requirements for the CACS

  10. 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

  11. 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)

  12. Description of the THYDE-P code

    International Nuclear Information System (INIS)

    This paper is a preliminary report about the methods and the models applied to a computer code named THYDE-P which is concerned with thermal-hydraulic transients of a PWR plant following a large or small area break of a primary coolant system pipe, generally referred to as a loss-of-coolant accident (LOCA). The THYDE-P code deals not only with blowdown phase, but also with reflooding phase. What characterizes the THYDE-P code is its entirely new model for the primary loop network. The code user information and the programming detail are not included in this report, but in a future documentation. (auth.)

  13. 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.

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

    International Nuclear Information System (INIS)

    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

  15. 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.

  16. 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...

  17. TRACE validation against FIX-II test no. 3025

    International Nuclear Information System (INIS)

    The paper describes the results of the validation of the TRACE code against FIX-II LOCA Blowdown and Pump Trip Heat Experiment No. 3025. The FIX-II facility was a scaled down model of the Swedish type Boiling Water Reactor (BWR) with external recirculation pumps. The experiment simulated the 31% break in one of the recirculation lines. The experimental facility consisted of test section with model of one fuel assembly, spray condenser, bypass and downcomer, and two recirculation lines in which one simulated the broken loop. The results obtained with TRACE v5.0 Patch 2 are in general in a good agreement with the experimental measurements. (author)

  18. 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)

  19. 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.

  20. 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)

  1. COMPARE-MOD 1: a code for the transient analysis of volumes with heat sinks, flowing vents, and doors

    International Nuclear Information System (INIS)

    This report describes the COMPARE-MOD 1 computer program developed for performing transient containment subcompartment pressure response analyses of nuclear power plants, including those with ice condensers. The subcompartments are represented as volumes (less than or equal to 100) that are connected by junctions (less than or equal to 200) and may have blowdown (less than or equal to 5 sets). The volume thermodynamics and flow equations are for a homogeneous mixture, assumed to be in thermodynamic equilibrium, consisting of any one, or any combination, of (a) steam, (b) two-phase water to its triple point, and (c) any three perfect gases such as air, helium, etc

  2. 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).

  3. 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

  4. 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.

  5. Safety and relief valve testing

    International Nuclear Information System (INIS)

    Two major programmes of safety relief valve testing have been established in support of system design and the selection of valves for Sizewell 'B'. The Central Electricity Generating Board has constructed the BRAVO facility to reproduce PWR pressuriser fluid conditions for the full scale testing of pressuriser relief system valves. Current tests on BRAVO, on a full size tandem pair of pressuriser pilot operated safety relief valves, will be followed by the full flow testing of a pressuriser safety valve. Tests at secondary circuit conditions have been carried out at the Siemens Kraftwerk Union Test Facilities, W Germany, where candidate MSSV's have been tested at full scale blowdown conditions. (author)

  6. 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)

  7. 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

  8. Numerical Modeling of Pulse Detonation Rocket Engine Gasdynamics and Performance

    Science.gov (United States)

    Morris, C. I.

    2003-01-01

    Pulse detonation engines (PDB) have generated considerable research interest in recent years as a chemical propulsion system potentially offering improved performance and reduced complexity compared to conventional gas turbines and rocket engines. The detonative mode of combustion employed by these devices offers a theoretical thermodynamic advantage over the constant-pressure deflagrative combustion mode used in conventional engines. However, the unsteady blowdown process intrinsic to all pulse detonation devices has made realistic estimates of the actual propulsive performance of PDES problematic. The recent review article by Kailasanath highlights some of the progress that has been made in comparing the available experimental measurements with analytical and numerical models.

  9. Simplified Analysis of Pulse Detonation Rocket Engine B1owdown Gasdynamics and Performance

    Science.gov (United States)

    Morris, Christopher I.

    2001-01-01

    Pulsed detonation rocket engines (PDREs) have generated considerable research interest in recent years as a chemical propulsion system potentially offering improved performance and reduced complexity compared to conventional rocket engines. The detonative mode of combustion employed by these devices offers a thermodynamic advantage over the constant-pressure deflagrative combustion mode used in conventional rocket engines and gas turbines. However, while this theoretical advantage has spurred a great deal of interest in building PDRE devices, the unsteady blowdown process intrinsic to the PDRE has made realistic estimates of the actual propulsive performance problematic. The recent review article by Kailasanath highlights some of the difficulties in comparing the available experimental measurements with numerical models. The goal of this paper is to improve understanding of PDRE blowdown gasdynamics and performance issues through use of a simplified model that captures the essential features of the unsteady blowdown process, and yet remains computationally inexpensive. The PDRE system studied here is highly idealized, consisting of a constant-area detonation tube with one end closed and the other end open to the environment. The tube is prefilled with a gaseous propellant mixture with no initial velocity or outflow to the environment. The detonation is initiated instantaneously at the closed end of the device. Chapman-Jouguet (C-J) post-detonation gas conditions are calculated using the CET89 version of the NASA thermochemical code. The I-D, unsteady method of characteristics is used to calculate the flowfield following the detonation front. See the compressible flow texts by Thompson and Zucrow and Hoffman for details of this method. Parametric studies of the effect of mixture stoichiometry, fill temperature, and blowdown pressure ratio on performance are reported. A comparison of the performance of an idealized straight-tube PDRE with a conventional steady

  10. Depressurisation studies. Phase 3: results of Tests 142 and 143

    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 temperatures between 220 and 2500C has been concluded. Measurements were made of transient density, pressure, and temperature variations in a two-phase fluid in the system during the discharge. Phase 3 tests, which are reported, examined the blowdown of a straight 206m dia pipe, 4m long, with particular emphasis on the transient void distribution across a section of the pipe. The results of two tests are detailed. (UK)

  11. Experimental study of thermohydraulic processes in justification of passive systems operability of WWER NPP

    International Nuclear Information System (INIS)

    The operation of passive safety systems is based on the use of gravitation, natural circulation processes, compressed gases energy. The passive systems ensure shutdown, reactor shutdown cooling and continuous after-heat removal. The results of investigations of thermohydraulic processes, during which the operability of WWER NPP passive safety systems have been justified, are considered. The processes are blowdown of subcooled liquid into opposing steam flow at GE-2 system starting-up, undeveloped boiling of subcooled liquid on horizontal tubes under condensation operation of WWER steam generator, heat transfer in air-air heat exchanger of WWER passive filtration system

  12. Passive containment system

    International Nuclear Information System (INIS)

    Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

  13. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    A containment system is described that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means

  14. 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.

  15. Full-scale Mark II CRT program: dynamic response evaluation test of pressure transducers

    International Nuclear Information System (INIS)

    A dynamic response evaluation test of pressure transducers was conducted in support of the JAERI Full-Scale Mark II CRT (Containment Response Test) Program. The test results indicated that certain of the cavity-type transducers used in the early blowdown test had undesirable response characteristics. The transducer mounting scheme was modified to avoid trapping of air bubbles in the pressure transmission tubing attached to the transducers. The dynamic response of the modified transducers was acceptable within the frequency range of 200 Hz. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Meservey, R.H.

    1976-04-01

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

  17. Toward the classification of higher-dimensional toric Fano varieties

    OpenAIRE

    Sato, Hiroshi

    2000-01-01

    The purpose of this paper is to give basic tools for the classification of nonsingular toric Fano varieties by means of the notions of primitive collections and primitive relations due to Batyrev. By using them we can easily deal with equivariant blow-ups and blow-downs, and get an easy criterion to determine whether a given nonsingular toric variety is a Fano variety or not. As applications of these results, we get a toric version of a theorem of Mori, and can classify, in principle, all non...

  18. 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.

  19. Review of the GOTHIC code and trial application

    International Nuclear Information System (INIS)

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

  20. Lower plenum voiding

    International Nuclear Information System (INIS)

    One of the phenomena involved in a Loss-of-Coolant-Accident (LOCA) in a Pressurized Water Reactor (PWR) may be Lower Plenum Voiding (LPV). This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with theories and data obtained by other researchers. 5 refs

  1. Concentrations of plastic strains in the clamping region of cylindrical shells under excessive loading

    International Nuclear Information System (INIS)

    Different clamping conditions and variations of the material behaviour are considered. Especially the plastic strain concentrations at the shell clamping are determined. The analyses have been carried out with a finite-element code. Standard stress analysis procedures like the ASME-code are critically assessed. The results show that only minor plastic strain concentrations have to be expected for a rigidly clamped shell loaded by internal pressure. The work hardening characteristic of the material has little influence on these strain concentrations. On the other hand considerable plastic strain concentrations have to be expected at a flange type clamping loaded by axial stresses in the shell. These concentrations depend on the work hardening characteristic of the material. Comparison of these analyses with standard stress analysis procedures shows that these procedures are conservative, provided the classification of the different types of stresses has been done in a correct way which, however, may be difficult for some problems. For instance, for the rigidly clamped shell bending stresses can be classified as secondary stresses, but for a flange type clamping bending stresses must be treated as primary stresses with lower limits. With these results the structural integrity of the core barrel clamping of a PWR under blowdown loading has been assessed. Even in case of a sudden and complete pipe break the structure is able to withstand blowdown loading. (orig./HP)

  2. 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.

  3. A critical flow model for flow through cracks in pipes

    International Nuclear Information System (INIS)

    The ability to determine leakage from cracks or flaws is of vital importance in demonstrating the leak-before-break condition. This paper describes a fully continuous analytical model that was developed to predict flow rates through tight cracks in pipes and tubes. The initial blowdown can be subcooled liquid, saturated liquid, two-phase mixture, saturated steam or superheated steam. The model is based on Henry's homogeneous non-equilibrium critical flow model (Henry 1970) for blowdown of initially subcooled or saturated liquids in frictionless pipes. Extensive modifications were made to account for different initial fluid conditions, surface roughness of the walls, and curved flow paths. The model can be used for calculating flow rates through fatigue or intergranular stress corrosion cracks (IGSCC). Guidelines for selecting the number of crack paths changes in direction are provided. Experimental data for flow through tight cracks is scarce. The model has been assessed using several open literature data (includes both fatigue and IGSCC cracks) for all fluid conditions. The work reported here has been used to help resolve leak-before-break issues for PWR and BWR plants

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

    International Nuclear Information System (INIS)

    With ROSA-I (Rig of Safety Assessment - I), 61 runs of the LWR blowdown experiment have been carried out under the conditions: model reactor type, BWR and PWR; reactor core, none, no-heating and heating; rupture position, upper and lower pressure vessel nozzle; initial discharge pressure, 40, 70 and 100 kg/cm2G; and rupture diameter, 25, 50, 70, 100 and 125 mm. The purpose was to obtain the data of thermal and hydrodynamic behavior in the reactor pressure vessel during a blowdown, including in-vessel pressure, coolant temperature, discharge flow rate, model fuel rod surface temperature and shock wave. Analysis was also made with the codes RELAP-2 and -3 developed by NRTS of the United States, to verify the calculation model used. In addition, the results of calculation with the shockwave analysis code DEPCO developed in JAERI were compared with those by experiment. The experimental facility ROSA-I and the results obtained with it and also the analyses made in this connection, are described in detail. (auth.)

  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. 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)

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

    International Nuclear Information System (INIS)

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

  8. 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

  9. 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.

  10. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

  11. Thermohydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

    International Nuclear Information System (INIS)

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 K, 50 K, and 100 K was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. Pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s. (author)

  12. A model for radionuclide transport in the Cooling Water System

    International Nuclear Information System (INIS)

    A radionuclide transport model developed to assess radiological levels in the K-reactor Cooling Water System (CWS) in the event of an inadvertent process water (PW) leakage to the cooling water (CW) in the heat exchangers (HX) is described. During and following a process water leak, the radionuclide transport model determines the time-dependent release rates of radionuclide from the cooling water system to the environment via evaporation to the atmosphere and blow-down to the Savannah River. The developed model allows for delay times associated with the transport of the cooling water radioactivity through cooling water system components. Additionally, this model simulates the time-dependent behavior of radionuclides levels in various CWS components. The developed model is incorporated into the K-reactor Cooling Tower Activity (KCTA) code. KCTA allows the accident (heat exchanger leak rate) and the cooling tower blow-down and evaporation rates to be described as time-dependent functions. Thus, the postulated leak and the consequence of the assumed leak can be modelled realistically. This model is the first of three models to be ultimately assembled to form a comprehensive Liquid Pathway Activity System (LPAS). LPAS will offer integrated formation, transport, deposition, and release estimates for radionuclides formed in a SRS facility. Process water and river water modules are forthcoming as input and downstream components, respectively, for KCTA

  13. 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

  14. 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.

  15. 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

  16. Analysis of Semiscale Mod-1 integral test with asymmetrical break (Test S-29-1)

    International Nuclear Information System (INIS)

    Selected experimental data obtained from Semiscale Mod-1 cold leg break Test S-29-1 and results obtained from analytical codes are analyzed. This test was the first integral blowdown reflood test conducted with the Mod-1 system and was a special test designed specifically to evaluate the sensitivity of the early Mod-1 core thermal response (0 to 5 sec after rupture) to the magnitude and direction of the core flow. To achieve this specific objective in Test S-29-1, the vessel side break area was reduced to approximately one-half the scaled break area associated with a 200 percent cold leg break test. The reduction in break area significantly reduced the core flow reversal that took place immediately after rupture and resulted in periods of positive core flow in the early portion of the test. The results obtained from this test are compared with results obtained from a 200 percent cold leg break test and the effect of core flow on early core thermal response is evaluated. Since Test S-29-1 was the first integral blowdown reflood test conducted with the Mod-1 system, data are also presented through the reflood stage of the test and the results are analyzed. The test data and the core thermal response calculated with the RELAP4 code are also compared

  17. A BWR pump suction-line 200% break test at ROSA-III Program (RUN 903)

    International Nuclear Information System (INIS)

    The Rig-of-Safety Assessment (ROSA)-III Program conducted a 200% recirculation pump-suction line break test, RUN 903, simulating a loss-of-coolant accident (LOCA) in a boiling water reactor (BWR). In this test, the main recirculation pumps (MRPs) were continuously operated during the transient following the break to study the influence of increased core flow rate on the system responses. This report describes major thermal-hydraulic phenomena observed in this test and presents all the experiment data. The effects of prolonged pump operation on system responses are described in comparison with the results of standard 200% break test, RUN 926, in which the core flow coasted down after the break faster than a scaled BWR LOCA condition. It is shown that the significant core heatup observed during the early blow-down phase in RUN 926 was not observed in RUN 903 due to an additional mass transport (approximately 6% of the initial system mass) from the downcomer into the core shroud as a result of the prolonged pump operation. It is clear that the lower-than-scaled transient core flow rate in the ROSA-III tests significantly affected the core thermal conditions especially during the early blowdown phase. (author)

  18. Post-test simulations of BTF-107: an in-reactor loss-of-coolant test with flow blockage and rewet

    International Nuclear Information System (INIS)

    The Blowdown Test Facility (BTF) located in the NRU reactor at Chalk River Laboratories is the principal experimental tool for the Canadian in-reactor safety research program. This dedicated facility was designed for performing integrated 'all effects' tests on CANDU-type fuel to generate data for verifying and assessing Canadian safety analysis codes and models. This paper briefly describes the first BTF experiment, designated BTF-107, and presents the results of post-test thermalhydraulics and fuel behaviour simulations of this experiment. The thermalhydraulics simulations, performed using the CATHENA computer code, focus on analyzing the response of the BTF test section following blowdown, during dryout, and during the final rewet phase of the experiment. The fuel behaviour simulations, performed using the ELOCA Mk5 code, give estimates of the thermo-mechanical and fission-product release behaviour of the fuel during the course of the transient. The results of these simulations illustrate the capabilities of the CATHENA and ELOCA codes to model the processes involved in this severe high-temperature transient, and indicate possible areas for future improvement of these codes. (author). 7 refs., 2 tabs., 7 figs

  19. 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.)

  20. 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

  1. 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)

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

    International Nuclear Information System (INIS)

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

  3. 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)

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

    Science.gov (United States)

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

    2012-12-01

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

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

    International Nuclear Information System (INIS)

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

  6. 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.)

  7. Quasi-One-Dimensional Modeling of Pulse Detonation Rocket Engines

    Science.gov (United States)

    Morris, Christopher I.

    2002-01-01

    Pulsed detonation rocket engines (PDREs) have generated considerable research interest in recent years as a chemical propulsion system potentially offering improved performance and reduced complexity compared to conventional rocket engines. The detonative mode of combustion employed by these devices offers a thermodynamic advantage over the constant-pressure deflagrative combustion mode used in conventional rocket engines and gas turbines. However, while this theoretical advantage has spurred a great deal of interest in building PDRE devices, the unsteady blowdown process intrinsic to the PDRE has made realistic estimates of the actual propulsive performance problematic. The recent review article by Kailasanath highlights some of the difficulties in comparing the available experimental measurements with numerical models. In a previous paper by the author, parametric studies of the performance of a single, straight-tube PDRE were reported. A 1-D, unsteady method of characteristics code, employing a constant-gamma assumption behind the detonation front, was developed for that study. Models of this type are computationally inexpensive, and are particularly useful for parametric performance comparisons. For example, a plot showing the specific impulse of various PDRE and steady-state rocket engine (SSRE) configurations as a function of blowdown pressure ratio. The performance curves clearly indicate that a straight-tube PDRE is superior in specific impulse to a SSRE with a sonic nozzle over the entire range of pressure ratios. Note, however, that a straight-tube PDRE in general does not compare favorably to a SSRE fitted with an optimized de Laval supersonic nozzle, particularly at the high pressure ratios typical for boost or in-space rocket applications. However, the calculations also show that if a dynamically optimized, supersonic de Laval nozzle could be could be fitted to a PDRE, then the specific impulse of the device would exceed that of a comparable SSRE

  8. 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

  9. 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 .

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-15

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

  11. 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

  12. Demonstration of a steam jet scrubber off-gas system and the burner efficiency of a mixed incinerator facility

    International Nuclear Information System (INIS)

    A full-scale incinerator system, the Consolidated Incineration Facility (CIF), is being designed to process solid and liquid low-level radioactive, mixed, and RCRA hazardous waste. This facility will consist of a rotary kiln, secondary combustion chamber (SCC), and a wet of-gas system. A prototype steam jet scrubber wastewater will be immobilized in a cement matrix after assumptions for the CIF. The scrubber wastewater will be immobilized in a cement matrix after the blowdown has been concentrated to a maximum solids concentration in a cross-flow filtration system. A sintered metal inertial filter system has been successfully tested. Burner efficiency was tested in a high intensity vortex burner, which destroyed the hazardous waste streams tested. These tests are detailed by the authors

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

    International Nuclear Information System (INIS)

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

  14. 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.

  15. Fuel rod mechanical deformation during the PBF/LOFT lead rod loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Results of four PBF/LOFT Lead Rod (LLR) sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. Each test employed four separately shrouded fuel rods. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods when subjected to a series of double ended cold leg break loss-of-coolant accident (LOCA) tests, and to determine whether subjecting these deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature and system pressure measurements with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements

  16. Thermal-hydraulic model verification calculation of PWR tests

    International Nuclear Information System (INIS)

    Test PWR 5 determined forces and pressure differences across the reactor pressure vessel and the internals in the test vessel during the first 80 ms by means of the present test data tapes. Furthermore a comparison between the measured data and those determined with the aid of the LECK program system is carried out. The following results were obtained in this connection: The qualitative pattern as compared between calculation and measurement shows a good agreement. Higher pressure differences resulted across the components due to the higher pressure gradients in the initial phase of the blowdown verification in the calculations. The best agreement of the pressure gradients was obtained with the verification calculations for a rupture opening time of 6 ms. Since there was no fluid/structural-dynamic coupling it was not possible to simulate the premature pressure reduction within the core barrel. The distribution of the initial temperature in the calculation did not always agree with that during the test. (orig.)

  17. 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

  18. 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.

  19. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

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

  20. Critical flow in small nozzles for saturated and subcooled water at high pressure

    International Nuclear Information System (INIS)

    Critical flow rate measurements of 4 mm and 16 mm nozzles have been performed with saturated and subcooled water at high pressure. The steady state and transient critical flow tests were conducted by discharging the fluid from a pressurized vessel through a blowdown leg. The fluid stagnation conditions upstream of the nozzle were measured by a gamma densitometer, thermocouple, and pressure transducer. The pressure and temperature of the tests range from 4.5 MPa to 15.0 MPa and from 530 K to 560 K, respectively. The results show that the flow upstream of the nozzle is stratified. The discharge mass flux obtained by this experiment is in good agreement with General Electric (GE) critical flow test data and Henry-Fauske and Burnell critical flow model predictions using a multiplier of 1.0 +- 0.3

  1. 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

  2. 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.

  3. Research and development work in the field of materials and strength

    International Nuclear Information System (INIS)

    The objective of research and development in the field of material and strength of light water reactor technology is the quantification and demonstration of the safety margin against failure as well as the possible further enhancement of safety. The conception of research covers the boundaries of possible material, law, and loading conditions on one hand, on the other hand, there could be introduced a basis safety by consequent optimization of production, service and safety technology. This conception is illustrated by the current BMFT-research programmes regarding light water reactor safety. The HDR-Programme is mainly concerned with accident conditions (blowdown, earthquake), whereas the research programme integrity of components deals with the integral behavior and long time properties of the reactor pressure vessel. (orig.)

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

    International Nuclear Information System (INIS)

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

  5. Motor-operated valves - French experience

    International Nuclear Information System (INIS)

    During the startup of French 900 and 1,300 MW plants, recurrent failures occurred on pressurizer, atmospheric steam dump, and safety injection isolation motor operated gate valves. The pressurizer original wedge gate valves were not able to isolate completely the blowdown flow. After extensive testing, two improved designs with increased disc guiding length and stellite overlay on friction surfaces were found to be adequate. On parallel-slide, double-disc, atmospheric steam dump block valves, it was necessary to increase the seating surfaces of the discs to avoid sticking in the closed position as a result of the disc-seat galling. The failures of several remote control couplings for the safety injection valves were the consequence of over-powered operators and inadequate testing procedures. Therefore current French research and development programs are aimed to optimize operators thrust relative to valve functions and designs

  6. Study of safety injection and reflooding

    International Nuclear Information System (INIS)

    To assess the safety of nuclear reactors, it is necessary to evaluate the performance of the emergency core cooling system which is designed to protect the reactor core from becoming overheated after a loss-of-coolant accident. The ob cetives of this project are: 1) To modify and develop ECCS evaluation code; 2) To investigate thermal hydraulic phenomena during reflood phase with various test loops. Best estimate calculations with RELAP 4/006 for SEMISCALE blowdown experiment (S-02-9) and reflooding analysis with REFLUX for the single rod experiments were carried out. 7 rod bundle experiments in vertical and horizontal flow channel were carried out to investigate the effects of channel orientation on the reflood characteristics. Observations were made of the precursory cooling effects on the reflooding using a single rod in annulus. (Author)

  7. Radiological impact of nuclear power stations in India on their environment

    International Nuclear Information System (INIS)

    There are five sites in India where nuclear power plants (NPPs) are operating. Two of them (Tarapur and Kalpakkam) are situated on the coast and the other three (Rawatbhata, Narora and Kakrapar) are situated inland. Except for the first power station at Tarapur, which has two BWR units, all other stations have PHW type reactors. Under normal operation, the low level radioactive wastes produced in the plants are diluted and dispersed in the environment after monitoring and treatment, if necessary. Gaseous wastes are discharged through a high stack and liquid wastes are diluted with condenser cooling water or coolant blowdown water (using cooling towers) and discharged to a water body (sea or reservoir). 3 refs

  8. 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

  9. 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)

  10. MASFLO: a computer code to calculate mass flow rates in the Thermal-Hydraulic Test Facility (THTF). Technical report

    International Nuclear Information System (INIS)

    This report documents a modular data interpretation computer code. The MASFLO code is a Fortran code used in the Oak Ridge National Laboratory Blowdown Heat Transfer Program to convert measured quantities of density, volumetric flow, and momentum flux into a calculated quantity: mass flow rate. The code performs both homogeneous and two-velocity calculations. The homogeneous models incorporate various combinations of the Thermal-Hydraulic Test Facility instrumented spool piece turbine flow meter, gamma densitometer, and drag disk readings. The two-velocity calculations also incorporate these instruments, but in models developed by Aya, Rouhani, and Popper. Each subroutine is described briefly, and input instructions are provided in the appendix along with a sample of the code output

  11. 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.-

  12. An analysis of CSNI standard problem, No. 8

    International Nuclear Information System (INIS)

    The CSNI International Standard Problem (ISP8), based on the Semiscale S-06-3 Test, was analyzed in the course of verification work of the computer code ALARM-P1. In this report, described was the result of the initial trial, which had been submitted to the CSNI. Due to the limitations of ALARM-P1 capability, only the blowdown portion of the transient was calculated. Though the hydraulic behavior before ECCS injection agreed with the test data, the ALARM-P1 could not continue calculation after 26 seconds due to severe predicted instability following the ECCS injection. The prediction of surface temperature of the heater rods was also unsatisfactory. Several problems to be improved have been identified both in the analytical model and the input data. (author)

  13. Trial evaluations in comparison with the 1983 safety goals

    International Nuclear Information System (INIS)

    This report provides retrospective comparisons of selected generic regulatory actions to the 1983 NRC safety goals, which had been issued for evaluation during a two-year period. The issues covered are those analyzed by the Office of Nuclear Reactor Regulation (NRR) (assisted in some cases by the Battelle Pacific Northwest Laboratory). The issues include auxiliary feedwater reliability, pressurized thermal shock, power-operated relief valve isolation, asymmetric blowdown loads on PWR primary systems, pool dynamic loads for BWR containments, and steam generator tube rupture. Calculated core-melt frequencies, mortality risks, and cost-benefit ratios are compared with the corresponding safety-goal quantitative design objectives. Considerations that should influence interpretation of the comparisons are discussed. Comments are included on whether and how the safety goals may have helped in the regulatory decision process and on problems encountered

  14. MAAP comparison to separate effects tests

    International Nuclear Information System (INIS)

    As part of the Modular Accident Analysis Program (MAAP) benchmarking efforts, data from several separate effects tests were identified as candidates for qualification of MAAP models. These included two critical flow tests, to check the critical flow models in MAAP, and a series of heat transfer tests, to check the heat transfer modeling in MAAP. The critical flow tests were selected because critical flow modeling was identified as important to the MAAP predictions in the phenomena identification study. The same basis was used to select fuel heat transfer tests for comparison to MAAP predictions. For these comparisons, data from two different vessel blowdown tests at General Electric and critical flow data for four different power operated relief valves from the Electric Power Research institute Valve Testing Program was used

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

    International Nuclear Information System (INIS)

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

  16. Application of frequency domain analysis to transient response of nuclear containment structures

    International Nuclear Information System (INIS)

    A combination of frequency domain and time domain analyses is proposed to obtain the dynamic responses of nuclear power plant containment structures. A soil-structure model of a boiling water reactor containment subjected to an assumed safety relief valve blowdown load is used as illustration. Linear time-invariant systems are analysed using input forcing functions with varying frequency contents. Time domain analysis is performed using a synthesized input forcing function. The system characteristic function is generated in the frequency domain through Fourier transforms of the response time history and the synthesized input time history. The frequency response due to any other forcing function is obtained in frequency domain by using the system characteristic function, and the response time history is obtained by inverse Fourier transforms of the frequency response. The results obtained by the proposed method are in close agreement with the conventional time domain dynamic finite element analysis. (Auth.)

  17. 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

  18. Two-phase critical flow models: a technical addendum to the CSNI state of the art report on critical flow modelling

    International Nuclear Information System (INIS)

    The purpose of this work was to obtain a comprehensive survey on the two-phase flow dynamics during accidental situations in nuclear reactors. About sixty theories regarding the two-phase flow calculation have been reviewed in this report with particular reference to their physical basis and assumptions; the aim is to control their applicability to nuclear safety problems. The main conclusions may be drawn as follows: the examined theories (perfect fluid, theories assuming thermodynamical equilibrium between liquid and vapor phases, non equilibrium models, etc.) are very different both for formulation and results; general validity of most theories is troublesome to check for the use of empirical coefficients. Moreover, according to the author's opinion, it is necessary to set up an organic program to obtain reliable experimental results in this field and to develop a model considering the whole blowdown transient

  19. 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)

  20. 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

  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. 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

  3. 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)

  4. Testing of the AP600 automatic depressurization system

    International Nuclear Information System (INIS)

    The Automatic Depressurization System (ADS) of the Westinghouse AP600 reactor will be used to provide controlled depressurization of the reactor coolant system (RCS). This will, in turn allow the initiation and long term operation of gravity driven cooling flow in the RCS. ADS tests were conducted at the VAPORE test facility in Casaccia, Italy through a Technical Cooperation Agreement between Westinghouse, ENEA, SOPREN/ANSALDO, and ENEL to produce data for the development and verification of computer codes to simulate the system. The test program also provided insights about the operation of valves supplied from various vendors that could be used in the AP600 ADS. The data gathered from the tests showed the ability of the ADS design to fulfill its function over the range of conditions expected in the AP600. The tests also demonstrated the abilities of gate and globe valves from several vendors to initiate and terminate an ADS blowdown as could be required in the AP600

  5. 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

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

    International Nuclear Information System (INIS)

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

  7. Environmental assessment of the projected uses for geopressured waters

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-16

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

  8. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  10. X-616 Chromium Sludge Lagoons pictorial overview, Piketon, Ohio

    International Nuclear Information System (INIS)

    The Portsmouth Gaseous Diffusion Plant uses large quantities of water for process cooling. The X-616 Liquid Effluent Control Facility was placed in operation in December 1976 to treat recirculation cooling water blowdown from the process cooling system. A chromium-based corrosion inhibitor was used in the cooling water system. A chromium sludge was produced in a clarifier to control chromium levels in the water. Chromium sludge produced by this process was stored in two surface impoundments called the X-616 Chromium Sludge Lagoons. The sludge was toxic due to its chromium concentration and therefore required treatment. The sludge was treated, turning it into a sanitary waste, and buried in an Ohio EPA approved landfill. The plant's process cooling water system has changed to a more environmentally acceptable phosphate-based inhibitor. Closure activities at X-616 began in August 1990, with all construction activities completed in June 1991, at a total cost of $8.0 million

  11. Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

    International Nuclear Information System (INIS)

    The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

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

    International Nuclear Information System (INIS)

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

  13. The LP-FP-2 severe fuel damage scenario and discussion of the relative influence of the transient and reflood phases in affecting the final condition of the bundle

    International Nuclear Information System (INIS)

    The purpose of this paper is to review the evidence from the OECD LP-FP-2 experiment that a high temperature excursion occurred within the center fuel module (CFM) during the reflood portion of the test, was caused by rapid metal-water reaction. It is shown that this reflood scenario explains many perplexing observations from the experiment, in particular, the small amount of fission products and hydrogen transported to the blowdown suppression tank (BST) as compared with the larger quantities trapped within the primary coolant system (PCS). The timing and destruction of the CFM upper tie plate, as well as the transport of fuel debris to the top of this plate, are also explained. In general, all measurements, observations, and analyses of the LP-FP-2 data indicate that most of the CFM damage occurred during a relatively short period of time coincident with the reflood portion of the experiment. 4 refs., 6 figs

  14. 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)

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

    International Nuclear Information System (INIS)

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

  16. Sample calculations on fuel rod behaviour during a LOCA with the code system SSYST-MOD 1

    International Nuclear Information System (INIS)

    The present paper shows results generated with SSYST, a program system developed for the analysis of the LWR fuel rod behaviour during a LOCA. A blowdown experiment in an out-of-pile test facility is analysed. The aim of the calculations is to demonstrate the influence of the various separate models, each describing a particular phenomenon such as rod internal pressure or rod mechanics on the behaviour of a hot rod by switching on these models sequentially. Calculations showed that the models presently included in the SSYST-system are able to describe the thermal and mechanical rod behaviour qualitatively in a correct way and that they may well be used to analyse the rod behaviour in a LWR during a LOCA. (Auth.)

  17. Characteristics of the ROSA-III test facility

    International Nuclear Information System (INIS)

    The ROSA-III facility is an integral test facility designed to simulate a loss-of-coolant accident (LOCA) and the performance of ECCS of a BWR. One of the important differences between ROSA-III and BWR is that in ROSA-III the jet pumps are installed outside the pressure vessel. It is important, therefore, to know characteristics of the jet pumps in ROSA-III to understand the fluid behavior during LOCA. Characteristics test results are described in Normal (M-N curve) and reverse flow conditions. As the reverse flow resistance in the jet pumps affects the core flow significantly during blowdown, the reverse flow characteristics are measured for the connecting pipings as well as the jet pumps. (author)

  18. Considerations for realistic ECCS evaluation methodology for LWRs

    International Nuclear Information System (INIS)

    This paper identifies the various phenomena which govern the course of large and small break LOCAs in LWRs, and affect the key parameters such as Peak Clad Temperature (PCT) and timing of the end of blowdown, beginning of reflood, PCT, and complete quench. A review of the best-estimate models and correlations for these phenomena in the current literature has been presented. Finally, a set of models have been recommended which may be incorporated in a present best-estimate code such as TRAC or RELAP5 in order to develop a realistic ECCS evaluation methodology for future LWRs and have also been compared with the requirements of current ECCS evaluation methodology as outlined in Appendix K of 10CFR50. 58 refs

  19. An operational experience with cooling tower water system in chilling plant

    International Nuclear Information System (INIS)

    Cooling towers are popular in industries as a very effective evaporative cooling technology for air conditioning. Supply of chilled water to air conditioning equipments of various plant buildings and cooling tower water to important equipments for heat removal is the purpose of chilling plant at PRPD. The cooling medium used is raw water available at site. Water chemistry is maintained by make-up and blowdown. In this paper, various observations made during plant operation and equipment maintenance are discussed. The issues observed was scaling and algal growth affecting the heat transfer and availability of the equipment. Corrosion related issues were observed to be less significant. Scaling indices were calculated to predict the behavior. (author)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. 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.

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

    International Nuclear Information System (INIS)

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

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. Lower plenum voiding

    Energy Technology Data Exchange (ETDEWEB)

    Bharathan, D.; Wallis, G.B.; Richter, H.J.

    1982-08-01

    One of the phenomena involved in a loss-of-coolant accident in a pressurized water reactor may be lower plenum voiding. This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. Steam generated in the reactor core may flow out of the bottom of the reactor core, turn in the lower plenum of the vessel, in a direction countercurrent to the emergency core coolant flow, and escape via the break. If its velocity is high enough, this steam may sweep water from the bottom (lower plenum) of the reactor vessel. Emergency coolant added to the vessel may also be carried out by the escaping steam and thus reflooding of the core would be delayed. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with data obtained by other researchers.

  12. 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)

  13. 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.

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

    International Nuclear Information System (INIS)

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

  15. Wake Management Strategies for Reduction of Turbomachinery Fan Noise

    Science.gov (United States)

    Waitz, Ian A.

    1998-01-01

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

  16. Calculated thermal-hydraulic response for Semiscale Mod-3 Test S-07-6 using RELAP5: a new LWR system analysis code

    International Nuclear Information System (INIS)

    The newly developed, advanced, light water reactor (LWR) simulation code, RELAP5, is used to analyze the response of Semiscale Mod-3 Test S-07-6. The objective of Test S-07-6 was to provide reference data to evaluate LWR integral blowdown, refill, and reflood behavior during a 200% cold leg break with emergency core coolant (ECC) injected into the intact loop cold leg. The calculated test results using RELAP5 illustrate many of the nonequilibrium and nonhomogeneous aspects of the ECC injection which are not directly observable in the test data. These results also demonstrate the capability of the RELAP5 code and compare well with the test data fo break flow, pressure, temperature, and density throughout the Semiscale Mod-3 system. The periodic depletion and replenishment of ECC water in the downcomer shown in the test data is also shown in the calculation

  17. ASME code safety valve rules - a review and discussion

    International Nuclear Information System (INIS)

    Safety valve rules, i.e., rules for overpressure protection by the use of various pressure relieving devices, vary somewhat among the five book sections of the ASME Boiler and Pressure Vessel Code which require such protection. This paper reviews those rules by discussing the following topics: Pressure relief device terminology and function. The problem of overpressure protection. Code rules for overpressure protection: rules for determining required relieving capacity; for allowable overpressure, for set pressure and set pressure tolerance; for blowdown. The various pressure relief devices permitted by the Code. Design of pressure relief valves. How relieving capacities are established and certified. The qualification of pressure relief device manufacturers. Installation guidelines. Concluding remarks. (orig.)

  18. 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.

  19. Chemical mode in secondary circuit of the Dukovany NPP units after TG condensers replacement

    International Nuclear Information System (INIS)

    The increase of the pH of SG feedwater on the 1. unit of Dukovany NPP led to enhancement of chemical mode of secondary circuit, what was identified in particular by the following: Reduction of concentration of iron in SG feedwater, Reduction of concentration of Sodium and Sulfates in SG blowdown water. This reduction is caused by shutdown of CPS thus by elimination of release of Na ions and SO4 from wrong operated ion-exchangers and their subsequent regeneration (part of cation exchanger in Na form and part of anion exchanger in SO4 form). Reduction of the WANO SG chemical index to the minimum theoretical value. It will be necessary to change criteria characterizing this index or to introduce our own modified index. In relation to CPS shutdown the costs for operating chemicals and for demineralized flushing water were reduced. (authors)

  20. Depressurisation studies. Phase 2: results of Tests 115 and 130

    International Nuclear Information System (INIS)

    Investigations into phenomena which might influence the course of a loss-of-coolant accident in a water reactor system have included depressurization studies of geometrically simple systems from typical reactor operating conditions. Tests on a simple pipe system containing water at pressures between 3.45 and 17.24 MPa and temperatures in the range of 2000 to 2500C have been concluded. Measurements were made of transient density, pressure and temperature variations in a two-phase fluid system during discharge. Phase 1 tests, previously reported, 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 dia pipe. This report is the first of three giving results of Phase 2 tests. It also includes a description of the test assembly, the instrumentation and the experimental procedure. (author)

  1. 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.

  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. Pulse Detonation Rocket Engine Research at NASA Marshall

    Science.gov (United States)

    Morris, Christopher I.

    2003-01-01

    This viewgraph representation provides an overview of research being conducted on Pulse Detonation Rocket Engines (PDRE) by the Propulsion Research Center (PRC) at the Marshall Space Flight Center. PDREs have a theoretical thermodynamic advantage over Steady-State Rocket Engines (SSREs) although unsteady blowdown processes complicate effective use of this advantage in practice; PRE is engaged in a fundamental study of PDRE gas dynamics to improve understanding of performance issues. Topics covered include: simplified PDRE cycle, comparison of PDRE and SSRE performance, numerical modeling of quasi 1-D rocket flows, time-accurate thrust calculations, finite-rate chemistry effects in nozzles, effect of F-R chemistry on specific impulse, effect of F-R chemistry on exit species mole fractions and PDRE performance optimization studies.

  4. 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.

  5. Development of nuclear standard filter elements for PWR plant

    International Nuclear Information System (INIS)

    Model FRX-5 and FRX-10 nuclear standard filter elements are used for the fluid clarification of the chemical and volume control system (CVCS), boron recycle system (BRS), spent fuel pit cooling system (SFPCS) and steam generator blowdown system (SGBS) in Qinshan Nuclear Power Plant. The radioactive contaminant, fragment of resin and impurity are collected by these filter elements, The core of filter elements consists of polypropylene frames and paper filter medium bonded by resin. A variety of filter papers are tested for optimization. The flow rate and comprehensive performance have been measured in the simulation condition. The results showed that the performance and lifetime have met the designing requirements. The advantages of the filter elements are simple in manufacturing, less expense and facilities for waste-disposal. At present, some of filter elements have been produced and put in operation

  6. 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

  7. Lower plenum voiding

    International Nuclear Information System (INIS)

    One of the phenomena involved in a loss-of-coolant accident in a pressurized water reactor may be lower plenum voiding. This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. Steam generated in the reactor core may flow out of the bottom of the reactor core, turn in the lower plenum of the vessel, in a direction countercurrent to the emergency core coolant flow, and escape via the break. If its velocity is high enough, this steam may sweep water from the bottom (lower plenum) of the reactor vessel. Emergency coolant added to the vessel may also be carried out by the escaping steam and thus reflooding of the core would be delayed. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with data obtained by other researchers

  8. Aerosol transport analysis of LWR high-consequence accidents using the HAA-4A code

    International Nuclear Information System (INIS)

    Use of the HAA-4A code to calculate removal of aerosol in containment due to inherent behavior mechanisms is described. Results for a PWR TMLB' scenario showed a source reduction of about a factor of 50 in CsI available for release to the environment through a catastrophic containment failure. Respirable CsI entering containment from the primary coolant system and melt-through blowdown was a factor of 25 less than the source. The principal removal mechanisms were particle growth due to Brownian and differential settling agglomeration and subsequent fallout. Sensitivities to important and uncertain parameters are discussed. Increased removal due to turbulent agglomeration and a larger expected source particle size are indicated. A seven control volume analysis took less than 1 minute of CPU time on an IBM 3033

  9. 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.

  10. 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.

  11. 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

  12. 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.)

  13. Fast reactor primary cover gas system proposals for CDFR

    International Nuclear Information System (INIS)

    A primary sodium gas cover has been designed for CDFR, it comprises plant to maintain and control; cover gas pressure for all reactor operating at fault conditions, cover gas purity by both blowdown and by a special clean-up facility and the clean argon supply for the failed fuel detection system and the primary pump seal purge. The design philosophy is to devise a cover gas system that can be specified for any LMFBR where only features like vessel and pipework size need to be altered to suit different design and operating conditions. The choice of full power and shutdown operating pressures is derived and the method chosen to control these values is described. A part active/part passive system is proposed for this duty, a surge volume of 250 m3 gives passive control between full power and hot shutdown. Pressure control operation criteria is presented for various reactor operating conditions. A design for a sodium aerosol filter, based on that used on PFR is presented, it is specifically designed so that it can be fitted with an etched disc type particulate filter and maintenance is minimised. Two methods that maintain cover gas purity are described. The first, used during normal reactor operation with a small impurities ingress, utilises the continuous blowdown associated with the inevitable clean argon purge through the various reactor component seals. The second method physically removes the impurities xenon and krypton from the cover gas by their adsorption, at cryogenic temperature, onto a bed of activated carbon. The equipment required for these two duties and their mode of operation is described with the aid of a system flow diagram. The primary pump seals requires a gas purge to suppress aerosol migration. A system where the argon used for this task is recirculated and partially purified is described. (author)

  14. Development of a CATHENA Fuel Channel Analysis Model for a Fuel Channel with Axial Variation of Radial Pressure Tube Creep in a Stratified Two-Phase Flow Condition

    International Nuclear Information System (INIS)

    A two-phase heat transfer phenomena in the fuel bundle strings located in a horizontal pressure tube with an axial variation of the radial creep, especially under a low stratified two-phase flow condition such as encountered in the CANDU reactor under the later stage of the blowdown phase of a LBLOCA, involves a complex heat transfer nature. This includes the conduction in the fuel rods, pressure tube, convection in the vapor and liquid regions, and radiation between the fuel rods exposed in the steam and the pressure tube, pressure tube and calandria tube. As these three modes of heat transfer has to be treated in a combined way, modeling the heat transfer phenomena inside the fuel bundle under the stratified flow during the later stage of LBLOCA blowdown has been one of the most challenging tasks in the CANDU safety analyses. The main reason for this hot attention is that it closely related to the integrity of the pressure tube. In this study a heat transfer model for handling this situation is developed, implemented and under preliminary testing of the analysis results. The analysis result up to now is encouraging and the validation of the model developed is ongoing. The major motivation of this study is to evaluate the conservatism of the current CANDU safety analysis methodology for a fuel channel with an axial variation of the radial creep of the pressure tube as easily experienced in the aged CANDU plant as it assumes the centerline of the fuel bundle string is the same as that of the pressure tube

  15. 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)

  16. The Savannah River Plant Consolidated Incineration Facility

    International Nuclear Information System (INIS)

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

  17. Preliminary study of coupling CFD code FLUENT and system code RELAP5

    International Nuclear Information System (INIS)

    Highlights: • System code RELAP5/MOD3.1 is coupled with CFD code FLUENT through DLL and UDF. • Transient water flow in a simple straight tube is tested using the coupled tool. • Simulation of Edwards’ pipe blowdown experiment using the coupled tool is conducted. • Coupled analysis of a more comprehensive thermal–hydraulic system is performed. - Abstract: The present paper discusses a coupling strategy of the 3D (three-dimensional) computational fluid dynamics (CFD) code ANSYS-FLUENT with the best estimate 1D (one-dimensional) thermal–hydraulic system code RELAP5/MOD3.1. Preliminarily, by using DLL (Dynamic Link Library) technology and FLUENT UDF (User Defined Functions), an explicit coupling method expected to be able to support the analysis of multi-purpose thermal–hydraulic phenomena in nuclear reactor systems has been developed. Calculations for two test cases using the coupled FLUENT/RELAP5 code have been carried out to test and demonstrate the coupling capability: (i) the first one consisting of single-phase water transient flow in a square straight tube with well controlled mass flow rates; (ii) the second one illustrating the process of single-phase water flow in a system including two closed loops and one vessel, on which loss of loop water flow due to pump trip and increase of loop water temperature are studied. Both reasonable 1D systematic behaviors and 3D distribution information are naturally obtained for the test cases. Besides, a study of a highly transient experiment problem, i.e. Edwards–O’Brien pipe blowdown problem, has been performed by using the coupled FLUENT/RELAP5 code. The results are compared with standalone RELAP5 calculation and available experimental data, which shows the coupled FLUENT/RELAP5 code’s acceptable potential for the capability of analyzing either simple single-phase or complex two-phase flow problem

  18. Experiment data report for Semiscale Mod-1 Test S-29-3; integral test from reduced initial pressure. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Crapo, H.S.; Collins, B.L.; Sackett, K.E.

    1976-09-01

    Recorded test data are presented for Test S-29-3 of the Semiscale Mod-1 special heat transfer 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-29-3 was conducted from an initial cold leg fluid temperature of 544/sup 0/F and an initial pressure of 1,760 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient starting from a lower initial pressure than that usually associated with pressurized water reactor operation. System flow was set to achieve a full core fluid temperature differential of 66/sup 0/F at 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 peaked 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 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 testtermination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-29-3 for future data analysis and test results reporting activites. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent.

  19. LOCA analysis of the IRIS reactor

    International Nuclear Information System (INIS)

    The IRIS reactor (International Reactor Innovative and Secure) is an integral, light water cooled, medium power reactor. IRIS has been selected as an International Near Term Deployable (INTD) reactor, within the Generation IV International Forum activities. The IRIS concept addresses the key-requirements defined by the US DOE for next generation reactors, i.e. enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). An innovative safety approach has been developed to mitigate the IRIS response to small-to-medium Loss of Coolant Accident (LOCA). This strategy is based on the interaction of IRIS compact containment with the reactor vessel to limit initial blowdown, and on depressurization through the use of a passive Emergency Heat Removal System (EHRS). A small Automatic Depressurization System (ADS) provides supplementary depressurization capability. A pressure suppression system is provided to limit the pressure peak following the initial blowdown to well below the containment design limit. The ultimate result is that during a small-to-medium LOCA, the core remains covered for an extended period of time, without credit for emergency water injection or external core makeup. The IRIS LOCA response is based on 'maintaining water inventory' rather than on the principle of safety injection. This novel safety approach poses significant issues for computational and analysis methods since the IRIS vessel and containment are strongly coupled, and the system response is based on the interaction between the two. The small break LOCA was calculated using RELAP5/mod3.3 and GOTHIC codes. Break of the largest line connected to the IRIS Reactor Pressure Vessel (RPV) was analyzed. The results of the calculations confirmed good performance of the IRIS system during LOCA. (author)

  20. Assessment of cooling tower (ultimate heat sink) performance in the Byron individual plant examination

    International Nuclear Information System (INIS)

    A time-dependent model of the Byron Nuclear Generation Station safety-related cooling towers has been developed for use with the Byron PRA (IPE). The model can either be run in a stand-alone program with externally supplied heat loads, or can be directly coupled into MAAP (Modular Accident Analysis Program). The primary feature of the model is a careful tracking of the basin temperature through the progression of different severe accidents. Heat removal rates from containment, both from containment fan-coolers and the residual heat removal system, are determined by the feed-back of this time-varying return temperature. Also, the inventory of the basin is tracked in time, and this is controlled by make-up, evaporative losses due to the heat load supplied to the towers, and the possibility of unsecured blowdown. The model has been used to determine the overall capabilities and vulnerabilities of the Byron Ultimate Heat Sink (UHS). It was determined that the UHS is very reliable with respect to maintaining acceptably low basin temperatures, requiring only at most two of eight operating cooling tower fans. Further, when the two units have their Essential Service Water (ESW) systems cross-tied, one of four ESW operating pumps is sufficient to handle the loads from the accident unit with the other unit proceeding to an orderly shutdown. The major vulnerability of the Byron UHS is shown to be the ability to maintain inventory, although the time-scales for basin dry-out are relatively long, being eight to twenty-one hours, depending upon when blowdown is secured. (author)

  1. 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.

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  6. 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

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

    International Nuclear Information System (INIS)

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

  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. Modification and application of the system analysis code ATHLET to trans-critical simulations

    International Nuclear Information System (INIS)

    Highlights: ► The pseudo two-phase method is proposed and utilized to modify ATHLET code. ► A smooth transition of void fraction under trans-critical transient can be realized by this method. ► The newly developed ATHLET-SC code can be adopted to simulate the blowdown process of a simplified model. - Abstract: During the loss of coolant accident (LOCA) of supercritical water cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from supercritical to subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. Using the current version of system code (e.g. ATHLET, REALP), calculation will be terminated due to the abrupt change of void fraction across the critical point (22.064 MPa). To solve this problem, a pseudo two-phase method is proposed by introducing a fictitious region of latent heat (enthalpy of vaporization hfg∗) at pseudo-critical temperatures. A smooth transition of void fraction can be realized by using liquid-field conservation equations at temperatures lower than the pseudo-critical temperature, and vapor-field conservation equations at temperatures higher than the pseudo-critical temperature. Adopting this method, the system code ATHLET is modified to ATHLET-SC mod 2 on the basic of the previous version ATHLET-SC mod 1 modified by Shanghai Jiao Tong University. When the fictitious region of latent heat is kept as a small region, the code can achieve an acceptable accuracy. Moreover, the ATHLET-SC mod 2 code is applied to simulate the blowdown process of a simplified model. The results achieved so far indicate a good applicability of the new modified code for the trans-critical transient.

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

    International Nuclear Information System (INIS)

    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

  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. 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.

  14. 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)

  15. Axisymmetric Numerical Modeling of Pulse Detonation Rocket Engines

    Science.gov (United States)

    Morris, Christopher I.

    2005-01-01

    Pulse detonation rocket engines (PDREs) have generated research interest in recent years as a chemical propulsion system potentially offering improved performance and reduced complexity compared to conventional rocket engines. The detonative mode of combustion employed by these devices offers a thermodynamic advantage over the constant-pressure deflagrative combustion mode used in conventional rocket engines and gas turbines. However, while this theoretical advantage has spurred considerable interest in building PDRE devices, the unsteady blowdown process intrinsic to the PDRE has made realistic estimates of the actual propulsive performance problematic. The recent review article by Kailasanath highlights some of the progress that has been made in comparing the available experimental measurements with analytical and numerical models. In recent work by the author, a quasi-one-dimensional, finite rate chemistry CFD model was utilized to study the gasdynamics and performance characteristics of PDREs over a range of blowdown pressure ratios from 1-1000. Models of this type are computationally inexpensive, and enable first-order parametric studies of the effect of several nozzle and extension geometries on PDRE performance over a wide range of conditions. However, the quasi-one-dimensional approach is limited in that it cannot properly capture the multidimensional blast wave and flow expansion downstream of the PDRE, nor can it resolve nozzle flow separation if present. Moreover, the previous work was limited to single-pulse calculations. In this paper, an axisymmetric finite rate chemistry model is described and utilized to study these issues in greater detail. Example Mach number contour plots showing the multidimensional blast wave and nozzle exhaust plume are shown. The performance results are compared with the quasi-one-dimensional results from the previous paper. Both Euler and Navier-Stokes solutions are calculated in order to determine the effect of viscous

  16. 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)

  17. 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.

  18. Numerical analyses of flow distributions in nuclear fuel assemblies affected by grid deformations

    International Nuclear Information System (INIS)

    Highlights: • Deformed spacer grid of a fuel assembly restricts coolant flow. • CFD analyses are conducted to assess flow redistribution and recovery. • Flow field is analyzed for normal operation, blowdown and reflood phases. • Forty-five times hydraulic diameter is required to recover 95% of flow rate. - Abstract: In the event of a safety shutdown earthquake (SSE) in a nuclear power plant, the spacer grid of the fuel assembly will be deformed as a result of the vibrations. If the flow area in a subchannel is reduced due to the grid deformation, the coolant flow will be restricted and consequently a loss of flow occurs in the affected fuel assembly during the accident. In this study, computational fluid dynamics (CFD) analyses are conducted in order to assess the flow redistribution and flow recovery in fuel assemblies. The real geometries of an outer grid and mixing vane are used in the simulation, and the region including the inner grid is modeled as a porous media zone. The resistance coefficients of the porous media model are determined using CFD analyses. The Reynolds-averaged Navier–Stokes equation with a non-linear turbulence model was used to solve the three-dimensional anisotropic turbulence flow in the rod bundles during normal operation, blowdown, and reflood phases following a loss-of-coolant accident (LOCA). In these analyses, it is assumed that forty percent of the flow area is blocked by grid deformations. The results demonstrate that a downstream distance of 45 times the hydraulic diameter is required for the coolant flow to recover to 95% of the original flow rate in the affected fuel assembly

  19. Experiment data report for Semiscale Mod-1 Test S-29-3; integral test from reduced initial pressure

    International Nuclear Information System (INIS)

    Recorded test data are presented for Test S-29-3 of the Semiscale Mod-1 special heat transfer 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-29-3 was conducted from an initial cold leg fluid temperature of 5440F and an initial pressure of 1,760 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient starting from a lower initial pressure than that usually associated with pressurized water reactor operation. System flow was set to achieve a full core fluid temperature differential of 660F at 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 peaked 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 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 testtermination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-29-3 for future data analysis and test results reporting activites. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent

  20. 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

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. 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

  4. Interactions Between Hillslope Processes and Channel Form in a Logged Coastal Redwood Forest

    Science.gov (United States)

    Lisle, T. E.; Reid, L. M.; Hilton, S.

    2009-12-01

    Fork mainstem. Blowdown in the 10 years following logging was most frequent along reaches adjacent to clearcuts, and cross sections have aggraded where wood accumulated. Downstream channel changes at Caspar Creek thus reflect a complex pathway of influence from upstream logging: 1) Reduced vegetation cover increases runoff and erosivity of flows; 2) Altered erosivity increases sediment input by expanding the channel network and increasing headcut and bank erosion; 3) A portion of the increased load is deposited in the mainstem channel where logging-related blowdown reduces sediment transport capacity and where the proportional increase in flow is lower because a lower percentage of the upstream catchment has been logged.

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yen-Shu, E-mail: yschen@iner.org.t [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

    2011-05-15

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

  7. 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

  8. Characterizing experiments of the PPOOLEX test facility

    International Nuclear Information System (INIS)

    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

  9. 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

  10. HTGR-process steam/cogeneration and HTGR-steam cycle program. Semiannual report, October 1, 1979-March 31, 1980

    International Nuclear Information System (INIS)

    Progress in the design of an 1170-MW(t) High-Temperature Gas-Cooled Reactor (HTGR) Nuclear Steam Supply (NSS) is described. This NSS can integrate favorably into present petrochemical and primary metal process industries, heavy oil recovery operations, and future shale oil recovery and synfuel processes. The economics appear especially attractive in comparison with alternative coal-fired steam generation. Cost estimates for central station power-generating 2240- and 3360-MW(t) HTGR-Steam Cycle (HTGR-SC) plants are updated. The 2240-MW(t) HTGR-SC is treated to a probabilistic risk evaluation. Compared with the earlier 3000-MW(t) design, the results predict a slightly increased risk of core heatup, owing to the result of eliminating the capability of the boiler feed pump to operate at atmospheric backpressure. The differences in risk, however, are within the calculational uncertainties. Preliminary results of the ranking of safety enhancement features for the 1170-MW(t) HTGR indicate that the following modifications offer the most promise: (1) capability for main loop rundown, (2) natural circulation core auxiliary cooling, and (3) PCRV blowdown capability through the helium purification system to minimize activity release during some core heatups

  11. 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

  12. 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.

  13. Dynamics of learning in multilayer perceptrons near singularities.

    Science.gov (United States)

    Cousseau, Florent; Ozeki, Tomoko; Amari, Shun-Ichi

    2008-08-01

    The dynamical behavior of learning is known to be very slow for the multilayer perceptron, being often trapped in the "plateau." It has been recently understood that this is due to the singularity in the parameter space of perceptrons, in which trajectories of learning are drawn. The space is Riemannian from the point of view of information geometry and contains singular regions where the Riemannian metric or the Fisher information matrix degenerates. This paper analyzes the dynamics of learning in a neighborhood of the singular regions when the true teacher machine lies at the singularity. We give explicit asymptotic analytical solutions (trajectories) both for the standard gradient (SGD) and natural gradient (NGD) methods. It is clearly shown, in the case of the SGD method, that the plateau phenomenon appears in a neighborhood of the critical regions, where the dynamical behavior is extremely slow. The analysis of the NGD method is much more difficult, because the inverse of the Fisher information matrix diverges. We conquer the difficulty by introducing the "blow-down" technique used in algebraic geometry. The NGD method works efficiently, and the state converges directly to the true parameters very quickly while it staggers in the case of the SGD method. The analytical results are compared with computer simulations, showing good agreement. The effects of singularities on learning are thus qualitatively clarified for both standard and NGD methods. PMID:18701364

  14. 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

  15. 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.)

  16. 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.

  17. Post-test analysis of the BTF-104 severe-fuel-damage experiment using the CATHENA thermal-hydraulics code

    International Nuclear Information System (INIS)

    In the BTF-104 (Blowdown Test Facility) experiment a single fully-instrumented, CANDU reg-sign-type fuel element was subjected to conditions representative of a Loss-Of-Coolant Accident (LOCA) with additional Loss-Of-Emergency-Coolant Injection (LOECI). Depressurization was followed by a period of degraded steam cooling achieving fuel temperatures of the order of 1800-1900 C. Material and coolant temperatures, and fission-product releases were monitored. The principal objective of the BTF-104 experiment was to determine the timing and amount of short-lived fission product release from the fuel element at the target fuel temperature. Information generated from BTF experiments is used by the CANDU Owners Group (a consortium of AECL and Canadian nuclear utilities) for validating computer codes used in licensing CANDU reactors. Simulations of the BTF-104 experiment were conducted using the CATHENA code (Canadian Algorithm for THErmalhydraulic Network Analysis) to provide a better understanding of the thermal-hydraulic phenomena observed in the experiment. This paper summarizes the major events and conditions of the BTF-104 experiment, describes the CATHENA idealization of the BTF-104 fuel stringer and BTF loop, and discusses important results obtained by these simulations

  18. Analysis code for medium and small rupture accidents in ATR. LOTRAC/HEATUP

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    In the evaluation of thermo-hydraulic and fuel temperature transient changes in the events which are classified in medium and small rupture accidents of reactor coolant loss that is the safety evaluation event of the ATR, the analysis code for synthetic thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC and the detailed analysis code for fuel temperature HEATUP are used, respectively. By using the LOTAC, the thermo-hydraulic behavior of reactor cooling facility and the temperature behavior of fuel at the time of blow-down are analyzed, and also the characteristics of changing reactor thermal output is analyzed, considering the functioning characteristics of emergency core cooling system. Based on the data of thermo-hydraulic behavior obtained by the LOTRAC, the time of beginning the turn-around of fuel cladding tube temperature obtained by the data of ECCS pouring characteristics, the heat transfer rate after the turn-around and so on, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The LOTRAC code, the HEATUP code, various analysis models, and rupture simulation experiment are reported. (K.I.)

  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. Simulation of COMEDIE Fission Product Plateout Experiment Using GAMMA-FP

    International Nuclear Information System (INIS)

    FThis phenomenon is particularly important under a VHTR design with vented low pressure confinement (VLPC), because the vent allows the prompt release of fission products accumulated within the primary circuit to environment during an initial blow-down phase after pipe break accidents. In order to analyze the fission product plateout, an numerical model was developed by Yoo et al. and incorporated into the GAMMA-FP code in the past. The GAMMA-FP model was validated against two experiment data, i.e., VAMPYR-1 and OGL, during the development phase. One of the well-known experiments for fission product plateout is the COMEDIE experiment. In this work, the COMEDIE experiment has been simulated using the GAMMA-FP code to investigate the reliability and applicability of the plateout model of GAMMA-FP. The COMEDIE experiment for fission product plateout was simulated using the GAMMA-FP code in this work. A good agreement was achieved between the measured and predicted plateout activities. The existing solution scheme was modified to allow larger time step size for fission product analysis in order to speed-up the computational time. Nevertheless, the modification of the existing numerical model of GAMMA-FP is necessary when a simulation capability of a long duration of plateout period (e.g., 60 years) is targeted

  1. Updated TRAC analysis of an 80% double-ended cold-leg break for the AP600 design

    International Nuclear Information System (INIS)

    An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) 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 80% break size was calculated by Westinghouse to be the most severe large-break size for the AP600 design. The LBLOCA transient was calculated to 144 s. Peak cladding temperatures (PCTS) were well below the Appendix K limit of 1,478 K (2,200 F), but very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCT 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 phase. The reasons for these differences are still being investigated. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse

  2. 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.

  3. Transitional behavior of a supersonic flow in a two-dimensional diffuser

    International Nuclear Information System (INIS)

    Two-dimensional blow-down type supersonic wind tunnel was designed and built to investigate the transient behavior of the startup of a supersonic flow from rest. The contour of the divergent part of the nozzle was determined by the MOC calculation. The converging part of the nozzle, upstream of the throat was contoured to make the flow profile uniform at the throat. The flow characteristics of the steady supersonic condition were visualized using the high-speed schlieren photography. The Mach number was evaluated from the oblique shock wave angle on a sharp wedge with half angle of 5 degree. The measured Mach number was 2.4 and was slightly less than the value predicted by the design calculation. The initial transient behavior of the nozzle was recorded by a high-speed digital video camera with schlieren technique. The measured transition time from standstill to a steady supersonic flow was estimated by analyzing the serial images. Typical transition time was approximately 0.1sec

  4. 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.

  5. Effectiveness of evacuating combustible gases by two parallel expellers closely coupled at one end of a gas pipeline

    International Nuclear Information System (INIS)

    Expeller performance has been formulated in terms of its capability to create suction pressure at the throat. This formulation has been used to assess the effectiveness of evacuating combustible gases from a pipeline section from one end using dual expellers mounted in parallel on two adjacent blow-down stacks. A general formulation was derived to address any situation of asymmetry in the stack resistance, asymmetry in the expellers' power as well overall pipeline resistance to suction flow. Solutions of the closed-form equations were obtained and presented on performance graphs showing the ratio of the suction flow using dual expellers to that using either one in a single mode. It was found that there are conditions at which expelling with dual expellers exceed that of either expeller operating alone. It was also shown that when asymmetric expellers are used, where one expeller is more powerful than the other, the benefits of using two expellers is realized up to a limiting degree of asymmetry, beyond which the weaker expeller could be stalled and then reverse flow

  6. 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

  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 any MOV measurable parameters was evaluated by ORNL using MOVs that were mounted on test stands. Those tests led to the conclusion that the single most informative MOV measurable parameter was also the one which was most easily acquired, namely the motor current. Motor current signature analysis (MCSA) was found to provide detailed information related to the condition of the motor, motor operator, and valve across a wide range of levels. As part of the MOV aging assessment, several tests were carried out by ORNL on MOVs having implanted defects and degradations. Tests were also performed on many MOVs located within a nuclear power plant. In addition, ORNL participated in the Gate Valve Flow Interruption Blowdown Test program carried out at Wyle Laboratories in Huntsville, Alabama. Results from all of these tests are summarized in this paper and several selected examples are given. Other areas covered in this paper include descriptions of relevant regulatory issues and activities, other related diagnostics research at ORNL, and interactions ORNL has had with outside organizations for the purpose of disseminating research results

  8. 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.

  9. 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)

  10. Post-test analysis of hot leg 2x25% break at PSB VVER facility using TRACE V5.0

    International Nuclear Information System (INIS)

    The presentation is structured as follows: TRACE code; PSB-VVER facility description; TRACE input decks description; HL-2x25-02 test description; TRACE calculation results; and Conclusions. TRACE = TRAC/RELAP is an advanced computational engine whose features include: Component-oriented reactor systems analysis code; Finite-volume two-fluid compressible flow code with 1, 2 and 3D geometry; Designed to analyse light water reactor transients up to the point of significant fuel damage. TRACE is (officially) the successor to RELAP5. The HL-2x25-02 test, Hot Leg Large Break 2x25%, was performed at the PSB-VVER test facility at EREC (Elektrogorsk Research and Engineering Centre, Elektrogorsk, Russia). The thermal-hydraulic processes related to this accident were examined with focus on the long reflooding phase. It is concluded that relatively good results were achieved as regards comparison of the main predicted and observed parameters of the primary circuit during the blowdown phase whereas the agreement was not satisfactory for the reflood phase, which was due to the different distribution of liquid phase over the primary circuit. Improved results should emerge from the planned modifications in a new version of the TRACE code (improved spacer grids, better counter-current flow model, ..). (P.A.)

  11. 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)

  12. Corrosion product behavior in VVER secondary systems

    International Nuclear Information System (INIS)

    Accumulation of corrosion products lead to some problems during long-term operation of VVER plants, such as secondary system component degradation including crud-induced local corrosion and corrosion cracking. Corrosion sludge and deposit removal from steam generators and other equipment is costly and time-consuming and leads to additional waste production. This problem is vital in the case of plant life extension. Appropriate solutions of the problem could be developed based on both Russian and international experience of the VVER fleet. Recommendations on how to mitigate corrosion product accumulation in VVER secondary systems were developed based on comparative analysis of available long-term data on corrosion product behavior in all the operating VVER plants, such as the following: Sludge and deposit accumulation in inner surfaces of secondary piping and components; Corrosion rate measurements using in-situ specimen testing at operated VVER plants; Efficiency of corrosion product removal from secondary system water by means of condensate polishers and steam generator blowdown cleanup systems; Sludge and deposit removal from steam generators during chemical cleaning; Secondary piping and components conservation efficiency during long outages. Comparative data analysis of corrosion product behavior has shown different corrosion product accumulation rates in Novovoronezh, Kola, Kalinin, Balakovo and Rostov NPPs. The said difference is due to different design and operation peculiarities. (author)

  13. 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

  14. Analytical investigation of pipe whip restraints against postulated high energy pipe raptures

    International Nuclear Information System (INIS)

    This paper addresses results obtained during the project related with increasing of safety of VVER 440/230 type reactors of Kozloduy NPP. The scope of investigation is ensuring the safety related equipment located in Steam Generator's Confinement (SGC) against postulated pipe ruptures of second circuit main steam piping systems. After performing of preliminary PSA (Probabilistic Safety Assessment) for postulated pipe ruptures of main steam piping systems in SGC it was assumed that only the mechanical damage of equipment should be considered. The temperature and dynamic pressure effects of blowdown steam jet over surrounding equipment were assumed as not critical for the safety. With the performed PSA the postulated pipe ruptures were determined critical for the safety and on this basis significantly was reduced the total number of the design analyses. The restraining of pipes whipping was done with two types of pipe whip restraining devices. Investigation of mounting locations of restraining devices was performed after free whip motion study, walkdowns for checking of collision with other equipment and pipelines analyses with the new devices in their mounting positions. The free whip analyses were performed as nonlinear time history analyses. In the design of pipes whip restrains nonlinear static and energy balance type of analyses were used. (authors)

  15. 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)

  16. TBL analysis by best estimate codes

    International Nuclear Information System (INIS)

    TRAC-BD1 (Version 12) is a three-dimensional thermal-hydraulic code for analyzing boiling water reactor (BWR) loss of coolant accidents (LOCAs). The code was developed by EG and G Idaho, Inc. and General Electric Co. (GE) under the sponsorship of the US Nuclear Regulatory Commission. SAFER, which was developed under the cooperative efforts of GE, Hitachi and Toshiba as a licensing code, is a one-dimensional thermal-hydraulic code for analyzing long term coolant inventory of the reactor system in BWR-LOCAs. Analyses for large and small break tests conducted with the Two-Bundle Loop (TBL) have been performed to assess the capabilities of these codes. The TBL was constructed to simulate thermal-hydraulic behaviors during LOCAs in a BWR/5 plant. It is the only integral test facility with two full size electrically heated bundles and consists of two full length jet pumps, vessel internals, two recirculation loops, two blowdown lines and emergency core cooling systems

  17. 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.

  18. Pseudo-component, thermal, reservoir simulation study of a proposed, low pressure, steam-assisted gravity drainage pilot project in Northeast Alberta

    Energy Technology Data Exchange (ETDEWEB)

    Uwiera-Gartner, M.M.E.; Carlson, M.R. [RPS Energy Canada Ltd (Canada)

    2011-07-01

    Bitumen production in the Athabasca oil sands in northeast Alberta typically uses the steam-assisted gravity drainage (SAGD) technique. In shallow bitumen resources developments, low pressure SAGD (LP-SAGD) and expanding solvent LP-SAGD (SLP-SAGD) techniques are applicable options as they maximize production and control steam-oil ratios (SORs). This paper presents the pseudo-component, thermal reservoir simulation study of a proposed LP-SAGD pilot project in northeast Alberta. The STARS simulation program was used for the study. Three potential development strategies were evaluated, including LP-SAGD, where 100% cold-water equivalent (CWE) steam is continuously injected into the reservoir for 10 years. Other strategies were SLP-SAGD, where the injection steam consists of 75% CWE and 25% solvent injected for 7 or 10 years, followed by well-pair blow-down and termination of solvent injection. The study found that injecting solvent improved sweep and oil recovery and reduced cumulative SORs (CSORs) by compared to using CWE steam only.

  19. 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)

  20. A two end-member model of wood dynamics in headwater neotropical rivers

    Science.gov (United States)

    Wohl, Ellen; Bolton, Susan; Cadol, Daniel; Comiti, Francesco; Goode, Jaime R.; Mao, Luca

    2012-09-01

    SummaryGeomorphic and ecological effects of instream wood have been documented primarily along rivers in the temperate zones. Instream wood loads in tropical rivers might be expected to differ from those in analogous temperate rivers because of the higher transport capacity and higher rates of wood decay in the tropics. We use data from four field sites in Costa Rica and Panama to demonstrate that wood loads are consistently lower in tropical rivers, despite substantial variations among tropical sites as a result of differences in mechanisms of wood recruitment. We develop a model of wood dynamics (recruitment, transport, and retention) based on differences in dominant wood recruitment mechanism. The steady-state end-member reflects sites where gradual recruitment of wood through individual tree fall creates a relatively consistent wood load through time and development of logjams is minimal. The episodic end-member reflects sites dominated by episodic mass recruitment via landslides or blowdowns. This facilitates formation of transient logjams, so that wood loads exhibit substantial spatial and temporal variation along the channel network. The model presented here should also apply to headwater streams in the temperate zone, although existing documentation suggests that jams are more persistent along streams in the temperate zone.

  1. 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.

  2. 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.

  3. 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.

  4. A study on optimization of environmental qualification envelope for Kori 3 and 4 NPP

    International Nuclear Information System (INIS)

    The purpose of this study is to present the reevaluation of the Main Steam Line Break (MSLB) applied Boron Injection Tank (BIT) removal and to optimize the environmental qualification (EQ) temperature envelope with thermal lag analysis and liquid entrainment method. BIT alleviates the reactor power excursion during Main Steam Line Break (MSLB) accident. Thermal lag analysis methods by NUREG-0588 is using four times condensing heat transfer coefficient on the passive heat sink surface, the forced convection heat transfer coefficient whenever the condensing is not occurring and during blowdown stage. And the entrainment model is that the all of the break regions within the secondary side are represented by non-homogeneous vapor volumes in which the liquid and steam are uniformly mixed throughout. These methods are focused on making higher the surface temperature of the safety equipment. For the analysis, amount of released mass and energy is calculated using the LOFTRAN code and containment temperature is predicted by CONTEMPT-LT 28 code. These two codes are used to for safety analysis. In accordance with the analysis result, a plant specific EQ test envelope was proposed for Kori 3 and 4 NPP

  5. 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.

  6. 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

  7. 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.

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

    International Nuclear Information System (INIS)

    During the preparation of a Safety Analysis Report at the Lawrence Livermore National Laboratory, we studied the release of chlorine from a compressed gas experimental apparatus. This paper presents the second part in a series of two papers on this topic. The first paper focuses on the frequency of an unmitigated release from the system; this paper discusses the consequences of the release. The release of chlorine from the experimental apparatus was modeled as an unmitigated blowdown through a 0.25 inch (0.0064 m) outside diameter tube. The physical properties of chlorine were considered as were the dynamics of the fluid flow problem. The calculated release rate was used as input for the consequence assessment. Downwind concentrations as a function of time were evaluated and then compared to suggested guidelines. For comparison purposes, a typical water treatment plant was briefly studied. The lower hazard presented by the LLNL operation becomes evident when its release is compared to the release of material from a water treatment plant, a hazard which is generally accepted by the public

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

    International Nuclear Information System (INIS)

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

  10. 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.

  11. 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)

  12. 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

  13. 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)

  14. 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

  15. 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

  16. Exergy analysis of a cogeneration power plant

    International Nuclear Information System (INIS)

    In the following study exergetic evaluation of a cogeneration power plant in operation with installed electrical capacity of 24 MW and process heat demand of 190 MW it is performed. The main objective of the research was to determine the influence of the increase in power generation capacity, raising the superheated steam parameters and the number of regenerative heaters on the second law efficiency and irreversibilities in the different components of the plant. To study the power plant was divided into subsystems: steam generator blowdown expander, main steam pipe, steam turbine regenerative heaters, reduction system, deaerator and pumps. The study results show that exergy losses and irreversibilities differ widely from one subsystem to another. In general, the total irreversibility accounted for 70.7% of primary fuel availability. The steam generator subsystem had the highest contribution to the irreversibility of the plant by 54%. It was determined that the increased steam parameters helps reduce the irreversibility and increase the exergetic efficiency of installation. The suppression of the reduction and incorporation of extraction-condensing turbine produce the same effect and helps to reduce power consumption from the national grid. Based on the results recommendations for improving plant efficiency are made. (full text)

  17. Sensitivity analysis to improve the gap conductance uncertainty for KINS-REM

    International Nuclear Information System (INIS)

    KINS has been using the Best Estimate Plus Uncertainty(BEPU) methodology to analyze the LBLOCA that is the design basis accident of emergency core cooling system(ECCS). KINS-REM(Realistic Evaluation Methodology) is the currently used for LBLOCA analysis methodology and has been improved continuously. One of the important issue of the improvements is the consideration about the uncertainty parameters related to fuel rod behaviors during LBLOCA.. Effect of Thermal Conductivity Degradation(TCD) of fuel rod has been studied to be considered in KINS-REM. For this purpose, the sensitivity analysis has been performed to improve the gap conductance uncertainty parameter in this study. The OPR1000 plant, Hanul unit 3 and 4, was selected as the reference plant. LBLOCA transient calculations have been performed by MARSKS. As the method of uncertainty quantification for gap conductance, the controls of the cladding roughness parameter(B) is changed to the controls of the global variable, effective gap conductance(hg) that is physically reasonable manner. The sensitivity analysis has been performed on the uncertainty multiplication coefficient of hg corresponding to the previous uncertainty range of cladding roughness parameter B in PCT calculations of LBLOCA. Through the comparison and analysis of the PCT values and behavior trends for reflood and blowdown, the range of uncertainty of the multiplication coefficient of the global variable hg, 2.34 - 0.66 and mean value 1.5 are reasonable to replace the local variable called the cladding roughness

  18. 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)

  19. Modeling of the hydraulic flows in a 4-loop PWR with a pipe-geometry

    International Nuclear Information System (INIS)

    The PWR safety studies involve an analysis of the consequences of a hypothetical rupture of a primary pipe. From the opening time, 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. Today Now, a single code dedicated to Fast Dynamic Analysis is sufficient to carry out hydrodynamic calculations involving acoustic transients and fluid mass transfers. EUR O-PLEXUS is a general fast dynamics computer code developed by the CEA-DMT and by JRC-Ispra. Its main applications are impacts, explosions, pipe transients and hydrodynamics. This paper describes the geometric and hydraulic models used to mesh the primary circuit. In order to validate the numerical model and to initialize the LOCA simulation with the reactor operating conditions, a hydrodynamic simulation of the flows in operating conditions is carried out in the primary circuit of 4-loop reactor. This paper also presents the initial conditions and the results of the calculation in operating conditions. (author)

  20. An analysis of CSNI Standard Problem No.6 by ALARM-B1 computer code

    International Nuclear Information System (INIS)

    Presented in this report are the results of a computer simulation of the CSNI International Standard Problem No.6 (ISP6) by the ALARM-B1 code. The aim of this standard problem is particularly to examine the capability of analytical models relating to the two-phase mixture level and the discharge mass flow rate using the individual participants' computer codes. The theoretical predictions were performed to explore the separate effects during the initial three seconds attended with a non-equilibrium physical phenomenon. The ISP6, which is based on the Battelle Frankfurt experiment duplicating a BWR steam line break accident, might not be appropriate as a bench mark problem of the ALARM-B1 program (version 2), because it contains an unreasonable demand treating a thermal non-equilibrium process with an equilibrium model. Apart from the pressure history arising from the non-equilibrium feature, the transient steam-water interface in the vessel was tracked sufficiently by using the bubble-rise-model incorporated in the ALARM-B1 code. In addition, the calculated mass flow rate at the exit plane during the steam phase blowdown was correct within the 15% experimental error bands. (author)

  1. System pressure effects on reflooding phenomena observed in the SCTF Core-I forced flooding tests

    International Nuclear Information System (INIS)

    The Slab Core Test Facility was constructed to investigate two-dimensional thermo-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 posturated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report described the analytical results on the effects of system pressure on reflooding phenomena observed in Tests Sl-SH2, Sl-01 and Sl-02 which are belonging to the SCTF Core-I forced-feed reflooding test series. Nominal system pressures in these tests are 0.4, 0.2 and 0.15 MPa, respectively. By comparison among the data of these three tests, the effects of system pressure on thermo-hydrodynamic behavior in the pressure vessel including the core and the primary coolant loops of the SCTF can be clarified under the forced flooding condition. Major items investigated in the present report are (1) overall temperature behaviors in the core, (2) change of heat transfer coefficient and heat flux at the rod surface before the quench, (3) two-dimensional thermo-hydrodynamic behaviors in the core and upper plenum and (4) hot leg carryover. (author)

  2. Experimental data of ROSA-III integral test run 706

    International Nuclear Information System (INIS)

    RUN 706 in the ROSA-III program, which obtains information of the thermo-hydraulic phenomena of the coolant in a loss-of-coolant accident (LOCA) using a simulated test facility for assessment of the system computer code, simulates a 200% double-ended break at the inlet of a recirculation pump. Purpose of the test was to observe thermo-hydraulic behavior in the core under LOCA conditions in core heating without ECCS actuation. The primary initial conditions are steam dome pressure 7.17MPa, steam dome temperature 561K (saturation), lower plenum subcooling 12K, initial power 3.405MW and core inlet flow 36.2 kg/s. Recirculation pump power supply, main steam discharge flow and feed water flow were suspended instantaneously upon the break. Following are the experimental results. (1) Jet pump suction uncovering and lower plenum flashing occurred 8.5s and 17s, respectively, after the break. (2) Lower plenum flashing improved core cooling in the lower half of the heater rods. (3) Electric power supply to the heater rods was cut off 156s after the break to protect the rods. (4) Separate mixture levels were observed in the core and the lower plenum during blowdown. (author)

  3. 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)

  4. 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

  5. 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

  6. 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

  7. Safety aspects of water chemistry in light water reactors

    International Nuclear Information System (INIS)

    The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

  8. 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.

  9. Effluent testing for the Oak Ridge Mixed Waste Incinerator: Emissions test for August 27, 1990

    International Nuclear Information System (INIS)

    On August 27, 1990, a special emissions test was performed at the K-1435 Toxic Substance Control Act Mixed Waste Incinerator. A sampling and analysis plan was implemented to characterize the incinerator waste streams during a 6 hour burn of actual mixed waste. The results of this characterization are summarized in the present report. Significant among the findings is the observation that less than 3% of the uranium fed to the incinerator kiln was discharged as stack emission. This value is consistent with the estimate of 4% or less derived from long-term mass balance of previous operating experience and with the value assumed in the original Environmental Impact Statement. Approximately 1.4% of the total uranium fed to the incinerator kiln appeared in the aqueous scrubber blowdown; about 85% of the total uranium in the aqueous waste was insoluble (i.e., removable by filtration). The majority of the uranium fed to the incinerator kiln appeared in the ash material, apparently associated with phosphorous as a sparingly-soluble species. Many other metals of potential regulatory concern also appeared to concentrate in the ash as sparingly-soluble species, with minimal partition to the aqueous waste. The aqueous waste was discharged to the Central Neutralization Facility where it was effectively treated by coprecipitation with iron. The treated, filtered aqueous effluent met Environmental Protection Agency interim primary drinking water standards for regulated metals. 4 refs., 2 figs., 10 tabs

  10. Effluent testing for the Oak Ridge mixed waste incinerator: Emissions test for August 27, 1990

    International Nuclear Information System (INIS)

    On August 27, 1990, a special emissions test was performed at the K-1435 Toxic Substance Control Act Mixed Waste Incinerator. A sampling and analysis plan was implemented to characterize the incinerator waste streams during a 6 hour burn of actual mixed waste. The results of this characterization are summarized in the present report. Significant among the findings is the observation that less than 3% of the uranium fed to the incinerator kiln was discharged as stack emission. This value is consistent with the estimate of 4% or less derived from long-term mass balance of previous operating experience and with the value assumed in the original Environmental Impact Statement. Approximately 1.4% of the total uranium fed to the incinerator kiln appeared in the aqueous scrubber blowdown; about 85% of the total uranium in the aqueous waste was insoluble (i.e., removable by filtration). The majority of the uranium fed to the incinerator kiln appeared in the ash material, apparently associated with phosphorous as a sparingly-soluble species. Many other metals of potential regulatory concern also appeared to concentrate in the ash as sparingly-soluble species, with minimal partition to the aqueous waste. The aqueous waste was discharged to the Central Neutralization Facility where it was effectively treated by coprecipitation with iron. The treated, filtered aqueous effluent met Environmental Protection Agency interim primary drinking water standards for regulated metals

  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. 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)

  13. 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

  14. 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.

  15. Microbiological treatment for removal of heavy metals and nutrients in FGD wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Shulder, Stephen J. [Structural Integrity Associates, Annapolis, MD (United States); Riffe, Michael R. [Siemens Water Technologies, General Industry Solutions, Warrendale, PA (United States); Walp, Richard J. [URS Corporation, Princeton, NJ (United States)

    2010-12-15

    In efforts to comply with the Clean Air Act many coal-fired fossil plants are installing wet flue gas desulfurization (WFGD) systems, also known as scrubbers, to remove sulfur dioxide (SO{sub 2}). Limestone slurry is injected into an absorber to promote the formation of calcium sulfate (CaSO{sub 4}) or gypsum. Chloride (chlorine in the fuel) becomes dissolved and increases in the absorber loop, which can lead to a more corrosive environment. Inert matter in the limestone also enters the absorber and must be reduced to meet the gypsum quality specification. To control the buildup of chloride and fines in the flue gas desulfurization (FGD) system a continuous blowdown or purge stream is utilized. Environmental regulations on the discharge of treated FGD wastewater are becoming increasingly more stringent to control impacts on the receiving body of water (stream, lake, river, or ocean). These new limitations often focus on heavy metals such as selenium and nutrients including nitrogen and phosphorus compounds. The FGD chloride purge stream is typically treated by chemical addition and clarification to remove excess calcium and heavy metals with pH adjustment prior to discharge. However this process is not efficient at selenium or nutrient removal. Information on a new approach using biological reactor systems or sequencing batch reactors (SBRs) to achieve reductions in selenium and nitrogen compounds (ammonia, nitrite, and nitrate) is discussed. A brief discussion on the physical/chemical pretreatment is also provided. (orig.)

  16. 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.

  17. 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.

  18. 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)

  19. 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

  20. Experiment data report for LOFT nonnuclear Test L1-4

    International Nuclear Information System (INIS)

    Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 2790C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response

  1. Surry Power Station secondary water chemistry improvement since steam generator replacement and the unit two feedwater pipe rupture

    International Nuclear Information System (INIS)

    Surry Power Station has two Westinghouse-designed three-loop PWRs of 811 MWe design rating. The start of commercial operation was in July, 1972 in No.1 plant, and May, 1973 in No.2 plant. Both plants began the operation using controlled phosphate chemistry for the steam generators. In 1975, both plants were converted to all volatile treatment on the secondary side due to the tube wall thinning corrosion in the steam generators, which was associated with the phosphate sludge that was building up on the tube sheets and created acidic condition. Thereafter, condenser and air leakage and steam generator denting occurred, and after the operation of 8 years 2 month of No.1 plant and 5 years 9 months of No.2 plant, the steam generators were replaced. A major plant improvement program was designed and implemented from 1979 to 1980. The improvement in new steam generators, the modification for preventing corrosion, the addition of a steam generator blowdown recovery system, the reconstruction of condensers, the installation of full flow, deep bed condensate polishers, the addition of Dionex 8,000 on-line ion chromatograph, a long term maintenance agreement with Westinghouse and so on are reported. (Kako, I.)

  2. 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)

  3. Reducing ion exchange resins rad-wastes, experience at EDF PWRs

    International Nuclear Information System (INIS)

    Life time of an ion exchange resin in a Nuclear Power Station (EDF PWR). At the end of its life, an ion exchange resin which has been used to treat radioactive streams becomes a radwaste itself. Its level of radioactivity depends on the point of use and consequently on the circuit where it was used. Roughly speaking, in a Nuclear Power Station PWR we can consider two types of radwaste families: High radioactive family Ion exchange resins which come from primary circuit: reactor control and storage pools. Ion exchange resins which have worked in a decontamination circuit: waste water treatment. Low radioactive family Ion exchange resins which come from secondary circuit: Steam Generator Blowdown By understanding and carefully applying some critical properties of ion exchange resins, such as total capacity, selectivity, and physical structure, it is possible for nuclear power stations to minimize radwaste volumes, while at the same time improving plant performance. This type of improvement can be facilitated by close cooperation and communication between the resin producer and the nuclear power user. (authors)

  4. 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)

  5. 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

  6. 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.

  7. The development of LOCA analysis codes for nuclear power plant

    International Nuclear Information System (INIS)

    This research aims at assessment of the best-estimate codes, so as to develop a reliable analysis method for their actual applications. There are two additional purposes in the study: The first is the development of methodology for sizing the safety systems for advanced reactor design using the best-estimate codes, and the other is the development of our own best-estimate methodology, referring to USNRC approval of the acceptance criteria for ECCS based on the best-estimate method. The use of the best-estimate codes as those assumed in FSAR by only input arrangement, has resulted in achievement of at least 250 K safety margin. The fact that the predicted PCT in LBLOCA analysis is well bounded within the acceptance criteria using the best-estimates codes, should be verified by the quantification of the code uncertainty in the future. In the case of computer code improvement, the reflood models have been improved and satisfactory results have been obtained. In the case of uncertainty evaluation, the calculational matrices based on the assessment of experiments with the improved RELAP5 code for the quantification of the code uncertainty have been formulated separately for blowdown and reflood phases. (Author)

  8. Application of the method for uncertainty and sensitivity evaluation to results of PWR LBLOCA analysis calculated with the code ATHLET. Pt. 2. Sensitivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Yaroslav; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Dresden (Germany)

    2014-04-15

    In the previously published part 1 of the paper the uncertainty analysis of the large break loss-of-coolant accident (LBLOCA) for German PWR Konvoi was performed using a statistical method, which is based on the Wilks' theory. The evaluated output parameter is the peak cladding temperature (PCT). The primary goal of this (second) part of the paper is a ranking of the input uncertainties, according to their contributions to the PCT uncertainty in the ATHLET simulation of PWR LBLOCA, by performing a sensitivity analysis. It was shown, that the first extended set of varied parameters used in part 1 can be considerably reduced without any statistically significant influence on the uncertainty analysis results. Thus, it can be shown that the input uncertainty vector based on the LBLOCA PIRT of AREVA GmbH was complete. To minimize the number of varied parameters the statistical t-test was used and, thus, a set of uncertainty parameters with significant impact on the uncertainty of the PCT was identified. The main contribution to the uncertainty of the first cladding temperature maximum during the blowdown phase of the accident is produced by the core parameters that affect the fuel's stored energy at the beginning of the accident. However, the major contributors to the uncertainty of the second PCT maximum are the uncertainties in the code models, and first of all the uncertainties in the heat transfer coefficients for dispersed and pure steam flows. (orig.)

  9. 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)

  10. 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

  11. 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.)

  12. Experimental data of ROSA-III integral test run 708

    International Nuclear Information System (INIS)

    Run 708 of the ROSA-III experimental program is an blowdown test without ECCS actuation in the BWR LOCA test series simulating a double-ended break on the inlet side of a recirculation pump. Electric power on the heater pins during transient simulates heat generation in the core by decay-heat power, delayed neutron fission power and stored heat release of fuel rods in BWR. Purpose of the ROSA-III test program is to provide comprehensive experimental data of thermal hydraulic behavior during BWR LOCA to assess the system computer code. The ROSA-III test facility is a volume scaled (1/424) system of large (--1000MWe) BWR to conduct an integral test on a BWR LOCA such as a 200% double-ended offset shear break on the inlet side of the pump in a recirculation loop. The primary initial conditions are steam dome pressure 7.12 MPa, steam dome temperature 548.7K, lower plenum subcooling 11K and core inlet flow 38.8 kg/s. During the system depressurization, ECCS was not actuated, so that temperatures of the heater pins rose after departure from nucleate boiling. Therefore, the 220 s after break, when fuel surface temperature at two points reached 973K (7000C), the electric power supplies to the heater pins were cut off to protect the pins. Run 708 was carried out successfully; the data are presented in graph. (author)

  13. Assessment study of RELAP5/MOD2 Cycle 36.04 based on pressurizer safety and relief valve tests

    International Nuclear Information System (INIS)

    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

  14. Consolidated Incineration Facility metals partitioning test

    International Nuclear Information System (INIS)

    Test burns were conducted at Energy and Environmental Research Corporation's rotary kiln simulator, the Solid Waste Incineration Test Facility, using surrogate CIF wastes spiked with hazardous metals and organics. The primary objective for this test program was measuring heavy metals partition between the kiln bottom ash, scrubber blowdown solution, and incinerator stack gas. Also, these secondary waste streams were characterized to determine waste treatment requirements prior to final disposal. These tests were designed to investigate the effect of several parameters on metals partitioning: incineration temperature; waste chloride concentration; waste form (solid or liquid); and chloride concentration in the scrubber water. Tests were conducted at three kiln operating temperatures. Three waste simulants were burned, two solid waste mixtures (paper, plastic, latex, and one with and one without PVC), and a liquid waste mixture (containing benzene and chlorobenzene). Toxic organic and metal compounds were spiked into the simulated wastes to evaluate their fate under various combustion conditions. Kiln offgases were sampled for volatile organic compounds (VOC), semi-volatile organic compounds (SVOC), polychlorinated dibenz[p]dioxins and polychlorinated dibenzofurans (PCDD/PCDF), metals, particulate loading and size distribution, HCl, and combustion products. Stack gas sampling was performed to determine additional treatment requirements prior to final waste disposal. Significant test results are summarized below

  15. 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)

  16. 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.

  17. Support of the eight-foot high temperature tunnel modification project

    Science.gov (United States)

    Ngo, Kim Chi; Mielke, Roland R.

    1987-01-01

    In order to meet the need for propulsion testing in the high supersonic range from Mach 4 to Mach 7, NASA has undertaken the modification of the Langley Eight Foot High Temperature Tunnel to add alternate Mach number capability and oxygen enrichment to allow the testing of operating engines at these Mach numbers and at true temperature tunnel. The transfer of liquid oxygen (LOX) from a storage vessel to a rocket engine generally requires the use of a pressurizing gas at high pressures. Although nitrogen is preferred, unfortunately, when gaseous nitrogen (GN2) is used as the pressurant to transfer liquid oxygen from a storage tank to the tunnel combustor, it contaminates the liquid oxygen and effects a loss of performance in the engine. The contamination of the LOX by the pressurizing GN2 is described, which may prove to be an important operational constraint. It is desirable to have reliable data concerning the penetration of GN2 into LOX during pressurization and the subsequent of self cleaning after blowdown.

  18. 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)

  19. 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.

  20. Planetary mission applications for space storable propulsion

    Science.gov (United States)

    Chase, R. L.; Cork, M. J.; Young, D. L.

    1974-01-01

    This paper presents the results of a study to compare space-storable with earth-storable spacecraft propulsion systems, space-storable with solid kick stages, and several space-storable development options on the basis of benefits received for cost expenditures required. The results show that, for a launch vehicle with performance less than that of Shuttle/Centaur, space-storable spacecraft propulsion offers an incremental benefit/cost ratio between 1.0 and 5.5 when compared to earth-storable systems for three of the four missions considered. In the case of VOIR 83, positive benefits were apparent only for a specific launch vehicle-spacecraft propulsion combination. A space-storable propulsion system operating at thrust of 600 lbf, 355 units of specific impulse, and with blowdown pressurization, represents the best choice for the JO 81 mission on a Titan/Centaur if only spacecraft propulsion modifications are considered. For still higher performance, a new solid-propellant kick stage with space-storable spacecraft propulsion is preferred over a system which uses space-storable propellants for both the kick stage and the spacecraft system.