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Sample records for blowdown

  1. Blowdown experiments and interpretation

    International Nuclear Information System (INIS)

    Rousseau, J.C.

    1975-01-01

    The CANON experiments which are being carried out in Grenoble, are intended for providing data for the development of a new theoretical analysis programmed in a computer code named BERTHA, which will predict the hydrodynamic phenomena of a blowdown accident, in a light water reactor. CANON experiments, carried out under adiabatic conditions, are a means of checking methods of pressure and temperature measurements. Presently, they allow the development of a new technique of measuring the mean void fraction in a section of the channel from epithermal neutron absorption, such measurements being made every one or two milliseconds. the BERTHA code is a one-dimensional model with the hypothesis of equal velocity of each phase, but taking into account a thermodynamic nonequilibrium. The energy flux at the phase interface is evaluated with a conduction model in the liquid layer at this interface. The numerical method used is a characteristic one. It is very slow as soon as the flow is in liquid phase, but it leads to an acceptable time-step in two-phase flow. Consequently, the method is well adapted to the problem of blowdown in which the fluid remains in liquid phase during a few milliseconds [fr

  2. Blowdown heat transfer experiment, (1)

    International Nuclear Information System (INIS)

    Soda, Kunihisa; Yamamoto, Nobuo; Osaki, Hideki; Shiba, Masayoshi

    1976-09-01

    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.)

  3. Forest blowdown and lake acidification

    International Nuclear Information System (INIS)

    Dobson, J.E.; Rush, R.M.; Peplies, R.W.

    1990-01-01

    The authors examine the role of forest blowdown in lake acidification. The approach combines geographic information systems (GIS) and digital remote sensing with traditional field methods. The methods of analysis consist of direct observation, interpretation of satellite imagery and aerial photographs, and statistical comparison of two geographical distributions-one representing forest blow-down and another representing lake chemistry. Spatial and temporal associations between surface water pH and landscape disturbance are strong and consistent in the Adirondack Mountains of New York. In 43 Adirondack Mountain watersheds, lake pH is associated with the percentage of the watershed area blown down and with hydrogen ion deposition (Spearman rank correlation coefficients of -0.67 and -0.73, respectively). Evidence of a temporal association is found at Big Moose Lake and Jerseyfield Lake in New York and the Lygners Vider Plateau of Sweden. They conclude that forest blowdown facilities the acidification of some lakes by altering hydrologic pathways so that waters (previously acidified by acid deposition and/or other sources) do not experience the neutralization normally available through contact with subsurface soils and bedrock. Increased pipeflow is suggested as a mechanism that may link the biogeochemical impacts of forest blowdown to lake chemistry

  4. Vapor generating unit blowdown arrangement

    International Nuclear Information System (INIS)

    McDonald, B.N.

    1978-01-01

    A vapor generating unit having a U-shaped tube bundle is provided with an orificed downcomer shroud and a fluid flow distribution plate between the lower hot and cold leg regions to promote fluid entrained sediment deposition in proximity to an apertured blowdown pipe

  5. LMR steam generator blowdown with RETRAN

    International Nuclear Information System (INIS)

    Wei, T.Y.C.

    1985-01-01

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

  6. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

    2000-01-01

    There is about 80m 3 /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

  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. Pos-calculation of blowdown experiments

    International Nuclear Information System (INIS)

    Gebrim, Anibal N.

    1997-01-01

    Three best estimate codes were utilized to analyse four blowdown separated tests. The experiments were created to study the pressure behavior, mass flow, mixture level movement, etc, related to the break area and the position. The theoretical results have a good agreement with the experimental results in three of the four tests. (author). 5 refs., 21 figs., 4 tabs

  9. Construction of the blowdown and condensation loop

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  10. LOFT system structural response during subcooled blowdown

    International Nuclear Information System (INIS)

    Martinell, J.S.

    1978-01-01

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

  11. PIV measurement at the blowdown pipe outlet

    International Nuclear Information System (INIS)

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

    2013-04-01

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

  12. The NRU blowdown test facility commissioning program

    Energy Technology Data Exchange (ETDEWEB)

    Walsworth, J A; Zanatta, R J; Yamazaki, A R; Semeniuk, D D; Wong, W; Dickson, L W; Ferris, C E; Burton, D H [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.

    1990-12-31

    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.

  13. A correlation for safety valve blowdown and ring settings

    International Nuclear Information System (INIS)

    Singh, A.; Shak, D.

    1982-01-01

    The blowdown of a spring loaded safety valve is defined as the difference between the pressure at which the valve opens and the pressure at which the valve fully closes under certain fluid flow conditions. Generally, the blowdown is expressed in terms of percentage of the opening pressure. An extensive series of tests carried out in the EPRI/PWR Utilities Valve Test Program has shown that the blowdown of safety valves can in general be strongly dependent upon the valve geometry and other parameters such as ring adjustments, spring stiffness, backpressure etc. In the present study, correlations have been developed using the EPRI safety valve test data to predict the expected blowdown as a function of adjustment ring settings for geometrically similar valves under steam discharge conditions. The correlation is validated against two different size Dresser valves

  14. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

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

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

    International Nuclear Information System (INIS)

    Hutcherson, M.N.

    1975-12-01

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

  16. PPOOLEX experiments with a modified blowdown pipe outlet

    International Nuclear Information System (INIS)

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

    2009-08-01

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

  17. Steam blowdown experiments with the condensation pool test rig

    International Nuclear Information System (INIS)

    Purhonen, H.; Puustinen, M.; Laine, J.; Raesaenen, A.; Kyrki-Rajamaeki, R.; Vihavainen, J.

    2005-01-01

    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)

  18. LOFT blowdown loop piping thermal analysis Class I review

    International Nuclear Information System (INIS)

    Kinnaman, T.L.

    1978-01-01

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

  19. 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

  20. Blowdown heat transfer surface in RELAP4/MOD6

    International Nuclear Information System (INIS)

    Nelson, R.A.; Sullivan, L.H.

    1978-01-01

    New heat transfer correlations for both PWR and BWR blowdowns have been implemented in the RELAP4/MOD6 program. The concept of a multidimensional surface is introduced with the heat flux from a given heat transfer correlation or correlations depicted as a mathematical surface that is dependent upon quality, wall superheat, mass flow and pressure. The heat transfer logic has been modularized to facilitate replacing boiling curves for future correlation data comparisons and investigations. To determine the validity of the blowdown surface, comparison has been performed using data from the Semiscale experimental facility. (author)

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

    International Nuclear Information System (INIS)

    Leung, J.C.M.

    1976-04-01

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

  2. Application of RELAP5 to a pipe blowdown experiment

    International Nuclear Information System (INIS)

    Carlson, K.E.; Ransom, V.H.; Wagner, R.J.

    1980-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1993-03-01

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

  4. Multiple blowdown pipe experiments with the PPOOLEX facility

    International Nuclear Information System (INIS)

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

    2011-03-01

    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

  5. 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

  6. Computer calculations of air and steam blowdown suppression

    International Nuclear Information System (INIS)

    Norris, D.M. Jr.; McMaster, W.H.; Landram, C.S.; Quinones, D.F.; Gong, E.Y.; Macken, N.A.

    1980-01-01

    We describe a computer code that combines an Eulerian incompressible-fluid algorithm (SOLA) with a Lagrangian finite-element shell algorithm. The former models the fluid and the latter models the containing structure in an analysis of pressure suppression in boiling-water reactors. The code (PELE-IC) calculates loads and structural response from air blow-down and from the oscillatory condensation of steam bubbles in a water pool. The fluid, structure, and coupling algorithms are tested by recalculating problems that have known analytical solutions, including tank drainage, spherical bubble growth, coupling for circular plates, and submerged cylinder vibration. Code calculations are also compared with the results of small-scale blowdown experiments. (orig.)

  7. Membrane distillation of industrial cooling tower blowdown water

    Directory of Open Access Journals (Sweden)

    N.E. Koeman-Stein

    2016-06-01

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

  8. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

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

  9. Early results of gate valve flow interruption blowdown tests

    International Nuclear Information System (INIS)

    DeWall, K.G.

    1988-01-01

    The preliminary results of the USNRC/INEL high-energy BWR line break flow interruption testing are presented. Two representative nuclear valve assemblies were cycled under design basis Reactor Water Cleanup pipe break conditions to provide input for the technical basis for resolving the Nuclear Regulatory Commission's Generic Issue 87. The effects of the blowdown hydraulic loadings on valve operability, especially valve closure stem forces, were studied. The blowdown tests showed that, given enough thrust, typical gate valves will close against the high flow resulting from a line break. The tests also showed that proper operator sizing depends on the correct identification of values for the sizing equation. Evidence exists that values used in the past may not be conservative for all valve applications. The tests showed that improper operator lock ring installation following test or maintenance can invalidate in-situ test results and prevent the valve from performing its design function. 2 refs., 12 figs., 2 tabs

  10. Condensation pool experiments with steam using DN200 blowdown pipe

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    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)

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

    International Nuclear Information System (INIS)

    Doval, A.

    1990-01-01

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

  12. High temperature pressure water's blowdown into water. Experimental results

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Kasahara, Yoshiyuki; Iida, Hiromasa

    1994-01-01

    The purpose of the present experimental study is to clarify the phenomena in blowdown of high temperature and pressure water in pressure vessel into the containment water for evaluation of design of an advanced marine reactor(MRX). The water blown into the containment water flushed and formed steam jet plume. The steam jet condensed in the water, but some stream penetrated to gas phase of containment and contributed to increase of containment pressure. (author)

  13. Fluid-structure interactions in PWR vessels during blowdown

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Leung, J.C.M.

    1977-01-01

    A small-scale experiment using Freon-11 at 130 0 F (54.4 0 C) 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 130 0 F (54.4 0 C) 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

  15. Prediction of LOFT core fluid conditions during blowdown and refill

    International Nuclear Information System (INIS)

    Grush, W.H.; White, J.R.

    1978-01-01

    One of the primary objectives of the LOFT (Loss-of-Fluid Test) Program is to provide data required to evaluate and improve the analytical methods currently used to predict the LOCA (Loss-of-Coolant Accident) response of large pressurized water reactors. The purpose of the paper is to describe the computer modeling methods used in predicting the fluid conditions in the LOFT core during the blowdown and refill phases of a nuclear LOCE (Loss-of-Coolant Experiment). Prediction results for a LOFT nonnuclear isothermal LOCE are compared to the experimental data to illustrate the validity of the modeling choices

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

    International Nuclear Information System (INIS)

    Fritz, G.

    1981-01-01

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

  17. Innovation of blow-down system in steam generators of a VVER 440 unit

    International Nuclear Information System (INIS)

    Matal, O.; Simo, T.; Mancev, M.D.

    1997-01-01

    The impurities getting into the steam generator with the feedwater are continually removed by the blowdown and unit sludge system. The mostly non-symmetrical type of pipe branches under steam generators at WWER-440 units causes nonuniform blowdown flow rates at the halves of the steam generator; this often leads to a blocking of the pipe with the lower flow rate. The most simple way of hydraulically equalizing the blowdown pipes is to implement symmetric blowdown pipes and to install efficient throttling elements in the pipe. The proposed innovation will make it possible to re-distribute the blowdown flow rates and to reduce more effectively the concentrations of impurities in steam generators. (M.D.)

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

    Directory of Open Access Journals (Sweden)

    Vuković Đorđe S.

    2016-01-01

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

  19. The test section of the COSIMA blowdown test facility

    International Nuclear Information System (INIS)

    Bruederle, F.; Hain, K.

    1980-08-01

    The test section of the COSIMA blowdown test facility has been designed as a geometric analogy of the core of a pressurized water reactor for a shortened single fuel rod simulator. Its design and instrumentation together with the whole loop allow to simulate out of pile and trace by measurements the energy and hydraulic conditions arising in a blowdown. Special attention is being given in this report to one particular design problem: the number of load cycles up to incipient cracking of the test section as a pressure vessel containing hot water at high pressures and subjected to extreme rates of temperature variation in excess of 300 K/min. The methods of calculating cyclic loads as specified in the German Technical Rules for Boilers (TRD) have been supplemented in such a way that the number of load cycles up to incipient cracking may now be determined not only by the mean wall temperature, which is difficult to measure, but equally also well by the outer wall temperature, which is easy to measure precisely. (orig.) [de

  20. 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

  1. Development of the CATHENA fuel channel model for an integrated blowdown and post-blowdown analysis for a 37-element CANDU fuel channel

    International Nuclear Information System (INIS)

    Rhee, B.W.; Shin, T.Y.; Yoo, K.M.; Kim, H.T.; Min, B.-J.; Park, J.H.

    2006-01-01

    The objective of this study is to develop a new fuel channel safety analysis system for covering both the blowdown analysis including the power pulse and the post-blowdown analysis with the same safety analysis code, CATHENA in a consistent manner. This new safety analysis methodology for a fuel channel analysis is expected to be better than the previous one used for the Wolsong 2,3,4 licensing which used CATHENA for the blowdown analysis and CHAN-II for the post-blowdown analysis, in several areas; consistency in the computer codes used and the modeling methods, the degree of uncertainty in the modeling and calculation. For this aim the existing CATHENA subchannel fuel channel model for a post blowdown analysis has been modified, and thus improved, and a processing program that conveys all the final state of the fuel channel at the end of blowdown analysis to the post-blowdown analysis as the initial conditions has been developed, and tested for its proper implementation for the intended purposes. A comparison of the results of this new analysis method with those of the Wolsong 2/3/4 Safety Analysis confirmed that the total heat transfer rate matches well up to 1000 sec, and then that of the new method begins to under-predict it consistently. On the other hand, the fuel temperatures of the center pin, inner ring fuel and the middle ring fuel are predicted by this new method to be lower than the old method by about 200 - 250 o C at the peak time. Considering the differences in these two analyses methodologies, especially the modeling of the fuel ring, a subchannel flow passage with an intermixing, and the radiation among the solid structures by considering every fuel individually, this trend of the results seems to be physically reasonable. However considerable future validation works are necessary to justify this new methodology for a licensing. (author)

  2. Overview PWR-Blowdown Heat Transfer Separate-Effects Program

    International Nuclear Information System (INIS)

    White, J.D.

    1978-01-01

    The ORNL Pressurized Water Reactor Blowdown Heat Transfer Program (PWR-BDHT) is a separate-effects experimental study of thermal-hydraulic phenomena occurring during the first 20 sec of a hypothetical LOCA. Specific objectives include the determination, for a wide range of parameters, of time to CHF and the following variables for both pre- and post-CHF: heat fluxes, ΔT (temperature difference between pin surface and fluid), heat transfer coefficients, and local fluid properties. A summary of the most interesting results from the program obtained during the past year is presented. These results are in the area of: (1) RELAP verification, (2) electric pin calibration, (3) time to critical heat flux (CHF), (4) heat transfer coefficient comparisons, and (5) nuclear fuel pin simulation

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

    International Nuclear Information System (INIS)

    Fayers, F.J.

    1975-10-01

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

  4. Analytical studies of blowdown thrust force and dynamic response of pipe at pipe rupture accident

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki

    1985-01-01

    The motion of a pipe due to blowdown thrust when the pipe broke is called pipe whip. In LWR power plants, by installing restraints, the motion of a pipe when it broke is suppressed, so that the damage does not spread to neighboring equipment by pipe whip. When the pipe whip of a piping system in a LWR power plant is analyzed, blowdown thrust and the dynamic response of a pipe-restraint system are calculated with a computer. The blowdown thrust can be calculated by using such physical quantities as the pressure, flow velocity, density and so on in the system at the time of blowdown, obtained by the thermal-fluid analysis code at LOCA. The dynamic response of a piping-restraint system can be determined by the stress analysis code using finite element method taking the blowdown thrust as an external force acting on the piping. In this study, the validity of the analysis techniques was verified by comparing with the experimental results of the measurement of blowdown thrust and the pipe whip of a piping-restraint system, carried out in the Japan Atomic Energy Research Institute. Also the simplified analysis method to give the maximum strain on a pipe surface is presented. (Kako, I.)

  5. The present status of the blowdown code BRUCH

    International Nuclear Information System (INIS)

    Karwat, H.

    1975-01-01

    The present status and the important features of the blowdown code for a PWR, BRUCH-D version 04 which is presently in use, are described. The code is to investigate the depressurization process, fluid dynamic situation in the core, the fuel temperature and the core mass flow as influenced by important primary system components such as steam generators, pumps etc. The code is a multinode point model with a fixed node arrangement. It makes use of the basic fluid dynamic equations describing the mass, energy and volume conservation as well as the momentum equation and the equation of state with appropriate assumptions. The core heat generation and heat transfer to the fluid is simulated by a given number of average fuel rods with up to 20 axial segmentation independent of the axial subdivision of the core fluid region. In parallel up to 5 types of the fuel rods can be studied. The pump behaviour is specified by input. For the break flow, the code provides three models; Bernoulli, homogenous and moody. The implicit-explicit method IMEX is used for the integration of the differential equations. An example for application of BRUCH-S to an experiment which is for a BWR but has some basis of BRUCH-D is also shown in the paper

  6. 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)

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

    International Nuclear Information System (INIS)

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

    2009-12-01

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

  8. Assessment of integrity for the pressure vessel internals of PWRs under blowdown loadings

    International Nuclear Information System (INIS)

    Geiss, M.; Benner, J.; Ludwig, A.

    1984-01-01

    In safety analysis of pressurized water reactors the loss-of-coolant accident plays a central role. Thereby a sudden break of a cold primary coolant pipe close to the reactor pressure vessel is postulated. The sudden pressure release of the primary system (blowdown) causes high dynamic loading on the pressure vessel internals. The resulting deformations must not impair shut down of the reactor and decay heat removal in an inadmissible way. For this assessment a blowdown analysis for a 1300 MW pressurized water reactor is carried out. These investigations are completed with a detailed stress analysis for the highly loaded core barrel clamping. The results show that the reactor pressure vessel internals are able to withstand blowdown loading. Even in case of a sudden and complete break of the primary coolant pipe the loading has to be twice as high to endanger the structural integrity. (orig.) [de

  9. Blowdown heat transfer surface in RELAP4/MOD6 and data comparisons

    International Nuclear Information System (INIS)

    Nelson, R.A.; Sullivan, L.H.

    1978-01-01

    RELAP4 is a thermal hydraulic analysis tool written to analyze transients in light water reactors (LWR). To date, most of the applications for RELAP4 have been to analyze postulated LOCA transients in LWR and the response of experimental systems to loss-of-coolant experiments. An important part of these analyses is the prediction of the fuel rod or heater surface temperature which involves the calculation of surface heat transfer coefficients. The paper describes the outcome of a significant blowdown heat transfer development effort which is incorporated in RELAP4/MOD6 (the current version of the code available to the United States public from the Argonne Code Center). The primary emphasis in the MOD6 development was on a PWR reflood capability. The best-estimate blowdown heat transfer correlation and logic were added to provide improved blowdown predictive capability

  10. Avoiding pressure shocks in HP blowdown lines; Vermeidung von Druckstossen in einer HD-Abschlammleitung

    Energy Technology Data Exchange (ETDEWEB)

    Stemme, R. [GESTRA AG, Bremen (Germany); Klackl, J. [EICHLER GmbH, Wien (Austria)

    2007-06-15

    Intermittent blowdown valves are installed in steam boilers as close as possible to the drum in order to avoid hydraulic pressure shocks. In the here presented case in the district heating plant Wels in Austria (gas-heated steam boiler 25 t/h 69 bar/290 C) this was not possible, and as a consequence the intermittent blowdown valves were damaged. By selecting valves suitable for this particular operating condition we have found a way to prevent pressure shocks. This example shows clearly that not only the operating data but also the right selection of the most suitable valve are of prime importance. (orig.)

  11. Analysis of the fluid-structure dynamic interaction of reactor pressure vessel internals during blowdown

    International Nuclear Information System (INIS)

    Schlechtendahl, E.G.; Krieg, R.; Schumann, U.

    1977-01-01

    The loadings on reactor internal structures (in particular the core barrel) induced during a PWR-blowdown must not result in excessive stresses and strains. The deformations are strongly influenced by the coupling of fluid and structure dynamics and it is necessary, therefore, to develop and apply new coupled analysis tools. In this paper a survey is given over work currently in progress in the Nuclear Research Center Karlsruhe and the Los Alamos Scientific Laboratory which aim towards 'best estimate codes'. The new methods will be verified by means of the HDR-blowdown tests and other experiments. The results of several scoping calculations are presented and illustrated by movie films. (orig.) [de

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

    International Nuclear Information System (INIS)

    Marshall, J.; Holland, P.G.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Singh, J.N.

    1978-01-01

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

  14. Full-scale HDR blowdown experiments as a tool for investigating dynamic fluid-structural coupling

    International Nuclear Information System (INIS)

    Krieg, R.; Schlechtendahl, E.G.; Scholl, K.-H.; Schumann, U.

    1977-01-01

    As an answer to rigorous safety requirements in reactor technology an experimental-theoretical program has been established to investigate safety-relevant mechanical aspects of LWR-blowdown accidents. Part of the program are several full-scale blowdown experiments which will be performed in the former HDR-reactor. As the conceptional study confirms, the primary goal is to find out, how big the safety margins of present LWR's in the case of a blowdown actually are, rather than simply to show that essential parts of the reactor will withstand such an accident. However, to determine the safety margins, the physical phenomena involved in the blowdown process must be understood and appropriate wave of description must be found. Therefore the experimental program is accompanied by the development of theoretical models and computer codes. A survey is given over existing methods for coupled fluid structural dynamics. The following approaches are used: - Specific finite difference-code for integrated treatment of both fluid and structure in 3D-geometry using the fast cyclic reduction scheme for solving Poisson's equation. - Modification of mass and stiffness matrices of FEM-models for shell dynamics by reducing the 3D incompressible fluid problem to 2D with the boundary integral equation method. This presently developed method has the capacity to deal with general problems in fluid-structural coupling. (Auth.)

  15. Blowdown and rewetting characteristics for AHWR under postulated LOCA - an analytical study

    International Nuclear Information System (INIS)

    Mukhopadhyay, D.; Chatterjee, B.

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) is a thorium fuelled, natural circulation driven and heavy water moderated reactor. The cooling of the nuclear fuel is achieved through natural circulation mode for the tube type reactor where hot and cold leg of the reactor has been designed to be long and high enough to avail the gravity head desired to overcome the hydraulic resistances in the flow path. The natural circulation cooling mode makes AHWR very different as compared to other tube type reactors with forced circulation e.g RBMK. This cooling feature which calls for longer pipes length and elevation head is having an influence on the blowdown characteristic and the initial fuel heatup characteristic of the reactor. Analyses of Loss of Coolant Accident carried out for different break sizes in the inlet header of the reactor identifies two competing transient forces namely 'blowdown force' and 'natural circulation' which act against each other due to virtue of the break location. The flow in the reactor channel is being decided by these two forces and eventually the flow condition decides the fuel heatup. It has been observed through analyses that variation of break sizes from moving smaller break sizes to bigger one (30% to 200%), causes an enhancement in blowdown forces and weakening of driving force for natural circulation as quality appears in cold leg section. A balance of these two forces is observed for 200% break case, causing a sustained flow stagnation condition leading to maximum fuel heat up among all the break cases. The blowdown characterization study is being carried out with RELAP5/mod3.4 code and the influences of transient forces on the fuel heatup are presented. It is concluded that the fuel heat up during blowdown phase is significantly dependent on the two competing forces namely blowdown and natural circulation which eventually depend on break sizes. The mist flow regime remains for a longer period during rewetting phase and the

  16. Design optimization on structure of blowdown in CPR1000 steam generator

    International Nuclear Information System (INIS)

    Wang Guoxian; Ren Hongbing; Zuo Chaoping; Zhu Yong; Mo Shaojia

    2014-01-01

    The structure of blowdown in CPR1000 steam generator has been optimized by eliminating the blowdown pipe and tube lane blocking, drilling holes in the peripheral tube lane, which can improve the accessibility of the central tube lane and facilitate inspecting and lancing. This paper detailed compares and analyzes the thermal hydraulic characteristic before and after optimization using GENEPI code which a special software for SG thermal hydraulic analysis. The results showed that the thermal hydraulic characteristic of steam generator meets the design requirements compared with the original design. Structure optimization can improve lancing effects, although the change of flow field distribution above the tubesheet leads to increase the number of tube subjected to sludge deposit. The analysis results proved the feasibility of the optimization. (authors)

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

    International Nuclear Information System (INIS)

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

    1994-12-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  19. Calculations of Edwards' pipe blowdown tests using the code TRAC P1

    International Nuclear Information System (INIS)

    O'Mahoney, R.

    1979-05-01

    The paper describes the results obtained using the non-thermal equilibrium LOCA code TRAC-P1 for two of a series of Pipe Blowdown Tests. Comparisons are made with the experimental values and RELAP-UK Mark IV predictions. Some discrepancies between prediction and experiment are observed, and certain aspects of the model are considered to warrant possible further attention. (U.K.)

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

    Science.gov (United States)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Krieg, R.; Schlechtendahl, E.G.

    1977-01-01

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

  2. Performance of R + D works to study the behavior during blowdown of LWRs

    International Nuclear Information System (INIS)

    Kaffanke, E.

    1980-03-01

    The aim of the RS 16B experiments is to study thermohydraulic behavior during blowdown of pressure vessels (with internals) of BWRS or PWR-steam generators. A small scale vessel will be used for simulation and measurement. Three series of experiments with varying paramters will be carried out in order to gain insight into flow behavior and loadings on internals caused by steam and feedwater line breaks. The experiments, besides supplying experimental results, will also serve to verify mathematical codes that have been developed to describe such highly complex phenomena. The contributions was as follows: 1) Planning and design of models to simulate BWR-internals. 2) Pre-test calculation of the effects of pressure differences on these models. 3) Post-test calculations upon completion of experiments. Since the beginning of the RS 16B-project in 1972, blowdown effects due to steam and feedwater line breaks have been calculated. A version of the short time computer program LAMB was used for the predictions. This program is an accepted code for the licensing procedure where blowdown analysis of BWR and PWR-steam generator is involved. (orig./HP) [de

  3. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon

    2016-01-01

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  4. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  5. Mitigation of caustic stress corrosion cracking of steam generator tube materials by blowdown -a case study

    International Nuclear Information System (INIS)

    Dutta, Anu; Patwegar, I.A.; Chaki, S.K.; Venkat Raj, V.

    2000-01-01

    The vertical U-tube steam generators are among the most important equipment in nuclear power plants as they form the vital link between the reactor and the turbogenerator. Over ∼ 35 years of operating experience of water cooled reactor has demonstrated that steam generator tubes are susceptible to various forms of degradation. This degradation leads to failure and outages of the power plant. A majority of these failures have been attributed to concentrated alkali attacks in the low flow areas such as crevices in the tube to tube sheet joints, baffle plate location and the areas of sludge deposits. Free hydroxides can be produced by improper maintenance of phosphate chemical control in the secondary side of the steam generators and also by the thermal decomposition of impurities present in the condenser cooling water which may leak into the feed water through the condenser tubes. The free hydroxides concentrate in the low flow areas. This buildup of free hydroxide in combination with residual stress leads to caustic stress corrosion cracking. In order to mitigate caustic stress corrosion cracking of Inconel 600 tubes, the trend is to avoid phosphate dosing. Instead All Volatile Treatment (AVT) for secondary water is used backed by full flow condensate polishing. Sodium hydroxide concentration is now being considered as the basis for steam generator blowdown. A methodology has been established for determining the blowdown requirement in order to mitigate caustic stress corrosion cracking in the secondary side of the vertical U-tube natural circulation steam generator. A case study has been carried out for zero solid treatment (AVT coupled with full flow condensate polishing plant) water chemistry. Only continuous blowdown schemes have been studied based on maximum caustic concentration permissible in the secondary side of the steam generator. The methodology established can also be used for deciding concentration of any other impurities

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

    International Nuclear Information System (INIS)

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

    2013-04-01

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

  7. 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.

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

    International Nuclear Information System (INIS)

    Gebrim, A.N.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Pappx, L.

    1994-01-01

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

  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)

    1996-12-31

    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. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    International Nuclear Information System (INIS)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments

  12. 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

  13. Thermal-hydraulic analysis of the semiscale Mod-1 blowdown heat transfer test series

    International Nuclear Information System (INIS)

    Cozzuol, J.M.

    1976-06-01

    Selected experimental thermal-hydraulic data from the recent Semiscale Mod-1 blowdown heat transfer test series are analyzed from an experimental viewpoint with emphasis on explaining those phenomena which influence core fluid behavior. Comparisons are made between the trends measured by the system instrumentation and the trends predicted by the RELAP4 computer code to aid in obtaining an understanding of the interactions between phenomena occurring in different parts of the system. The analyses presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a pressurized water reactor during a postulated loss-of-coolant accident

  14. RELAP5 progress summary: simulation of semiscale isothermal blowdown (Test S-01-4A)

    International Nuclear Information System (INIS)

    Kuo, H.H.; Wagner, R.J.; Carlson, K.E.; Kiser, D.M.; Trapp, J.A.; Ransom, V.H.

    1978-07-01

    The RELAP5/MOD''O'' LOCA analysis code has been applied to Simulation of the Semiscale Isothermal Blowdown Test (S-01-4A) from initiation to 60 seconds. Subcooled ECC injection was simulated from 23 seconds until accumulator emptying. The calculated results are in very good agreement with the experimental data. This is the first full system application of the RELAP5 code and only the pressurizer surge line resistance was modified to achieve the results reported. An analysis of the code execution time using a time-step statistical edit is included

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

    International Nuclear Information System (INIS)

    Papp, L.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-03-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  19. Experiment data report for semiscale Mod-1 test S-02-3 (blowdown heat transfer test)

    International Nuclear Information System (INIS)

    Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

    1975-09-01

    Recorded test data are presented for Test S-02-3 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-3 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,263 psig. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with a moderately heated core (75 percent design power level). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was also set at the 75 percent design level to achieve full core temperature differential. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.2 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling (DNB) occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core coolant injection or pressure suppression system coolant spray

  20. Exergy analysis and evolutionary optimization of boiler blowdown heat recovery in steam power plants

    International Nuclear Information System (INIS)

    Vandani, Amin Mohammadi Khoshkar; Bidi, Mokhtar; Ahmadi, Fatemeh

    2015-01-01

    Highlights: • Heat recovery of boiler blow downed water using a flash tank is modeled. • Exergy destruction of each component is calculated. • Exergy efficiency of the whole system is optimized using GA and PSO algorithms. • Utilizing the flash tank increases the net power and efficiency of the system. - Abstract: In this study, energy and exergy analyses of boiler blowdown heat recovery are performed. To evaluate the effect of heat recovery on the system performance, a steam power plant in Iran is selected and the results of implementation of heat recovery system on the power plant are investigated. Also two different optimization algorithms including GA and PSO are established to increase the plant efficiency. The decision variables are extraction pressure from steam turbine and temperature and pressure of boiler outlet stream. The results indicate that using blowdown recovery technique, the net generated power increases 0.72%. Also energy and exergy efficiency of the system increase by 0.23 and 0.22, respectively. The optimization results show that temperature and pressure of boiler outlet stream have a higher effect on the exergy efficiency of the system in respect to the other decision variables. Using optimization methods, exergy efficiency of the system reaches to 30.66% which shows a 1.86% augmentation with regard to the situation when a flash tank is implemented.

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

    Science.gov (United States)

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

    2012-01-01

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

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

    Science.gov (United States)

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

    2009-01-01

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

  3. The Toshiba Blow-Down MHD Test Facility, and Experiments on Non-Equilibrium Ionization

    International Nuclear Information System (INIS)

    Yamamoto, Y.; Ogiwara, H.; Shioda, S.; Miyata, M.; Goto, M.; Kasahara, E.

    1968-01-01

    The Toshiba blow-down MHD test facility, which was constructed in 1966 and has operated successfully in many experiments, is described. Operating conditions achieved are: the working gas is helium seeded with potassium, the maximum mass flow being 80 g/sec, the maximum seed fraction 0.1%; the gas static lies between temperature 1200 and 1700°K, the static pressure between 2.0 ∼ 1.2 atm, the velocity of gas in the generator channel between 1000 and 200 m/sec; the duration is up to 30 sec; the magnetic field is 2.7 T; the impurity of working gas is below 150 ppm. (author)

  4. Use of the modal superposition technique for piping system blowdown analyses

    International Nuclear Information System (INIS)

    Ware, A.G.; Macek, R.W.

    1983-01-01

    A standard method of solving for the seismic response of piping systems is the modal superposition technique. Only a limited number of structural modes are considered (typically those up to 33 Hz in the U.S.), since the effect on the calculated response due to higher modes is generally small, and the method can result in considerable computer cost savings over the direct integration method. The modal superposition technique has also been applied to piping response problems in which the forcing functions are due to fluid excitation. Application of the technique to this case is somewhat more difficult, because a well defined cutoff frequency for determining structural modes to be included has not been established. This paper outlines a method for higher mode corrections, and suggests methods to determine suitable cutoff frequencies for piping system blowdown analyses. A numerical example illustrates how uncorrected modal superposition results can produce erroneous stress results

  5. Core thermal response during Semiscale Mod-1 blowdown heat transfer tests

    International Nuclear Information System (INIS)

    Larson, T.K.

    1976-06-01

    Selected experimental data and results calculated from experimental data obtained from the Semiscale Mod-1 PWR blowdown heat transfer test series are analyzed. These tests were designed primarily to provide information on the core thermal response to a loss-of-coolant accident. The data are analyzed to determine the effect of core flow on the heater rod thermal response. The data are also analyzed to determine the effects of initial operating conditions on the rod cladding temperature behavior during the transient. The departure from nucleate boiling and rewetting characteristics of the rod surfaces are examined for radial and axial patterns in the response. Repeatability of core thermal response data is also investigated. The test data and the core thermal response calculated with the RELAP4 code are compared

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

    International Nuclear Information System (INIS)

    Umminger, K.; Mandl, R.; Nopper, H.; Siemens AG Unternehmensbereich KWU, Erlangen

    1987-01-01

    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 [de

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

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2004-03-01

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

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

    International Nuclear Information System (INIS)

    Chen, M.M.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    Howell, S.K.

    1978-01-01

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

  10. Effect of Uncertainty Parameters in Blowdown and Reflood Models for OPR1000 LBLOCA Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Byung Gil; Jin, Chang Yong; Seul, Kwangwon; Hwang, Taesuk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    KINS(Korea Institute of Nuclear Safety) has also performed the audit calculation with the KINS Realistic Evaluation Methodology(KINS-REM) to confirm the validity of licensee's calculation. In the BEPU method, it is very important to quantify the code and model uncertainty. It is referred in the following requirement: BE calculations in Regulatory Guide 1.157 - 'the code and models used are acceptable and applicable to the specific facility over the intended operating range and must quantify the uncertainty in the specific application'. In general, the uncertainty of model/code should be obtained through the data comparison with relevant integral- and separate-effect tests at different scales. However, it is not easy to determine these kinds of uncertainty because of the difficulty for evaluating accurately various experiments. Therefore, the expert judgment has been used in many cases even with the limitation that the uncertainty range of important parameters can be wide and inaccurate. In the KINS-REM, six heat transfer parameters in the blowdown phase have been used to consider the uncertainty of models. Recently, MARS-KS code was modified to consider the uncertainty of the five heat transfer parameters in the reflood phase. Accordingly, it is required that the uncertainty range for parameters of reflood models is determined and the effect of these ranges is evaluated. In this study, the large break LOCA (LBLOCA) analysis for OPR1000 was performed to identify the effect of uncertainty parameters in blowdown and reflood models.

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

    International Nuclear Information System (INIS)

    Benner, J.

    1984-03-01

    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.) [de

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

    International Nuclear Information System (INIS)

    1976-02-01

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

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

    International Nuclear Information System (INIS)

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

    1984-02-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. French studies on blow-down accident in light water reactors

    International Nuclear Information System (INIS)

    Pelce, J.

    1977-01-01

    The effects on fuel elements and containment buildings resulting from a rapid blow-down accident and the effectiveness of proposed emergency systems are currently being evaluated in France, using the so-called first generation computer codes. Some of these were developed by the constructing organization Framatome for the design of actual power plants; others were developed by the Nuclear Safety Division to back-up related safety studies. These codes are considered to be inadequate and for several years a large effort has been made jointly by EDF and safety authorities, and with the technical assistance of CEA, to make a significant improvement in the methods of assessment. Framatome also participates in this work to some extent. A more physical method is proposed. In particular, selected models are supported by a quite comprehensive experimental programme which is mainly analytical in nature, as follows: (1) Basic analysis, using experiments which are planned or in progress, such as CANON, MOBY-DICK, SUPER MOBY-DICK, REBECA (critical flow at the break and between sub-compartments of the containment building), ECOTRA (condensation on inner walls), TAPIOCA (phase separation at small cracks), EPIS (water and steam mixing during emergency injection), EDGAR (fuel cladding behavior). (2) More intricate or semi-integral analysis such as OMEGA and ERSEC (tests on in-core heat transfer during blow-down and rewetting), both of which are in progress, in-pile PHEBUS loop due to start operating in 1977 (fuel behavior during the accident), pump tests (EVA, POMPE). Future methods of assessing the reactor itself will include the physical models thus perfected: A first code, (CLYSTERE) has been written and can be used. Although it has not been validated experimentally, it can already evaluate the effect of some physical phenomena on the development of the accident. Work is being done on reconstructing the general flow chart of this code in order to improve the conditions of use, in

  16. Thermal-hydraulically controlled blowdown tests in the experimental facility COSIMA to study PWR fuel behavior: experimental and theoretical results

    International Nuclear Information System (INIS)

    Class, G.; Hain, K.; Meyder, R.

    1978-01-01

    The fuel behavior in the blow-down phase of a LOCA is of importance for fuel rods with high internal pressure and high rod power, because of the effects on clad failure of the small cladding deformations occurring. The operating results of the COSIMA facility show that, on the basis of the new developments for measuring technique and fuel rod simulators performed, reactor relevant blow-down performances can be conducted in a controlled and reproduceable manner. The mechanical and thermal-hydraulic states occurring in the test bed may be subject to computational checking. This permits on one hand to improve the computing models and on the other yields a confirmation of the high state of development of the available computer codes. Therefore it appears that, with the results from COSIMA and the associated theoretical work in the field of the blow-down process, difficult to treat experimentally, an essential contribution to verifying the models for accident calculations is given. The work scheduled for the next about 1 1/2 years will serve to further support the rather preliminary results and to extend the range of then application. (orig.) [de

  17. The Development of Computer Code for Safety Injection Tank (SIT) with Fluidic Device(FD) Blowdown Test

    International Nuclear Information System (INIS)

    Lee, Joo Hee; Kim, Tae Han; Choi, Hae Yun; Lee, Kwang Won; Chung, Chang Kyu

    2007-01-01

    Safety Injection Tanks (SITs) with the Fluidic Device (FD) of APR1400 provides a means of rapid reflooding of the core following a large break Loss Of Coolant Accident (LOCA), and keeping it covered until flow from the Safety Injection Pump (SIP) becomes available. A passive FD can provide two operation stages of a safety water injection into the RCS and allow more effective use of borated water in case of LOCA. Once a large break LOCA occurs, the system will deliver a high flow rate of cooling water for a certain period of time, and thereafter, the flow rate is reduced to a lower flow rate. The conventional computer code 'TURTLE' used to simulate the blowdown of OPR1000 SIT can not be directly applied to simulate a blowdown process of the SIT with FD. A new computer code is needed to be developed for the blowdown test evaluation of the APR1400 SIT with FD. Korea Power Engineering Company (KOPEC) has developed a new computer code to analyze the characteristics of the SIT with FD and validated the code through the comparison of the calculation results with the test results obtained by Ulchin 5 and 6 units pre-operational test and VAlve Performance Evaluation Rig (VAPER) tests performed by The Korea Atomic Energy Research Institute (KAERI)

  18. Experiment data report for semiscale Mod-1 Test S-01-5 (isothermal blowdown with core resistance simulator)

    International Nuclear Information System (INIS)

    Zender, S.N.; Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

    1975-04-01

    Recorded test data are presented for Test S-01-5 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-5 is one of several semiscale Mod-1 experiments which are counterparts of the LOFT nonnuclear experiments. System hardware is representative of LOFT with the design based on volumetric scaling methods and with initial conditions duplicating those identified for LOFT nonnuclear tests. Test S-01-5 was conducted with the secondary side of the steam generator pressurized with nitrogen gas in order to effectively eliminate heat transfer from the steam generator during blowdown and thereby to investigate the effect on overall system behavior of heat transfer from the steam generator. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The test was initiated at isothermal conditions of 2270 psig and 540 0 F by a simulated offset shear of the cold leg broken loop piping. During system depressurization, coolant was injected into the cold leg of the operating loop to simulate emergency core cooling (ECC). Following the blowdown portion of the test, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. The uninterpreted data from Test S-01-5 and the reference material needed for future data analysis and test results reporting activities are presented. 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. (U.S.)

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

    International Nuclear Information System (INIS)

    Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki

    1980-01-01

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

  20. Experiment data report for Semiscale Mod-1 test S-02-5 (blowdown heat transfer test)

    International Nuclear Information System (INIS)

    1975-12-01

    Recorded test data are presented for Test S-02-5 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-5 is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Test S-02-5 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,253 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was set to achieve the full design core temperature differential of 66 0 F. The flow resistance of the intact loop was based on core area scaling. 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 the LOFT nuclear fuel rods until such time that departure from nucleate boiling occurs

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

    International Nuclear Information System (INIS)

    Reeder, D.L.

    1979-01-01

    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/m 3 . These conditions are typical of those in commercial PWR systems at normal operating conditions

  2. Treatment of cooling tower blowdown water containing silica, calcium and magnesium by electrocoagulation.

    Science.gov (United States)

    Liao, Z; Gu, Z; Schulz, M C; Davis, J R; Baygents, J C; Farrell, J

    2009-01-01

    This research investigated the effectiveness of electrocoagulation using iron and aluminium electrodes for treating cooling tower blowdown (CTB) waters containing dissolved silica (Si(OH)(4)), Ca(2 + ) and Mg(2 + ). The removal of each target species was measured as a function of the coagulant dose in simulated CTB waters with initial pH values of 5, 7, and 9. Experiments were also performed to investigate the effect of antiscaling compounds and coagulation aids on hardness ion removal. Both iron and aluminum electrodes were effective at removing dissolved silica. For coagulant doses < or =3 mM, silica removal was a linear function of the coagulant dose, with 0.4 to 0.5 moles of silica removed per mole of iron or aluminium. Iron electrodes were only 30% as effective at removing Ca(2 + ) and Mg(2 + ) as compared to silica. There was no measurable removal of hardness ions by aluminium electrodes in the absence of organic additives. Phosphonate based antiscaling compounds were uniformly effective at increasing the removal of Ca(2 + ) and Mg(2 + ) by both iron and aluminium electrodes. Cationic and amphoteric polymers used as coagulation aids were also effective at increasing hardness ion removal.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. Evaluation of pressure drop across area changes during blowdown. Quarterly progress report for period ending June 30, 1976

    International Nuclear Information System (INIS)

    Weisman, J.

    1976-11-01

    Transient pressure drops across abrupt area changes are being determined in a series of blowdown experiments. These tests are being conducted with Freon 113 as the test fluid in a well instrumented apparatus. During this period, test runs were obtained with the first abrupt expansion test section. Test data from two typical runs are included in this report. Additional progress was made in developing the computer programs which were to be used in analyzing this data but funding of this analytical effort has been suspended

  6. Experiment data report for semiscale MOD-1 test S-01-3 (isothermal blowdown with core resistance simulator)

    International Nuclear Information System (INIS)

    Zender, S.N.

    1975-03-01

    Recorded test data are presented for Test S-01-3 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-3 is one of several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-3 employed an intact loop resistance that was low relative to that of the first test in the series (Test S-01-2) to establish the importance of intact loop resistance on system response during blowdown. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The test was initiated at isothermal conditions of 2245 psig and 538 0 F by a simulated offset shear of the cold-leg broken loop piping. During system depressurization, coolant was injected into the lower plenum of the pressure vessel to provide data on the effects of emergency core cooling on system response. Additionally, to aid in determination of the effects of accumulator gas on pressure suppression system response, the nitrogen used to charge the accumulator systems for Test S-01-3 was allowed to vent into the lower plenum following depletion of the coolant. (U.S.)

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

    International Nuclear Information System (INIS)

    Haynes, H.D.

    1988-01-01

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

  8. Experiment data report for semiscale Mod-1 test S-01-1B (isothermal blowdown with core resistance simulator)

    International Nuclear Information System (INIS)

    Crapo, H.S.; Jensen, M.F.; Sackett, K.E.; Zender, S.N.

    1975-05-01

    Recorded test data are presented for Test S-01-1B of the semiscale Mod-1 isothermal blowdown test series. System hardware is representative of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-1B is a repeat of Test S-01-1 with the exception that simulated ECC was injected into the cold leg of the intact loop rather than into the inlet annulus of the downcomer. The principal objective of Test S-01-1B was to determine whether a different ECC injection would significantly alter the system response during the period of ECC injection. Test S-01-1B was conducted from an initial temperature of 541 0 F and an initial pressure of 1630 psig. A simulated intermediate size double-ended hot leg break (0.00145 ft 2 break area on each end) was used to investigate the system response to a slow de-pressurization transient. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. Following the blowdown portion of Test S-01-1B, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. (U.S.)

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

    International Nuclear Information System (INIS)

    Follmer, Bernhard; Schnettler, Armin

    2002-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bertolini, E.; Toschi, R. [CNEN, Frascati (Italy); Lindley, B. C. [C.A. Parsons and Co. Ltd (United Kingdom); Brown, R.; McNab, I. R. [International Research and Development Co. Ltd., Newcastle Upon Tyne (United Kingdom)

    1966-11-15

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

  11. PWR blowdown heat transfer separate-effects program: Thermal-Hydraulic Test Facility experimental data report for test 100

    International Nuclear Information System (INIS)

    White, M.D.; Hedrick, R.A.

    1977-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 100, which is part of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 100 was conducted to investigate the response of heater rod bundle 1 and instrumented spool pieces with flow homogenizing screens to a double-ended rupture with equal break areas at the test section inlet and outlet. The primary purpose of this report is to make the reduced instrument responses during test 100 available. The responses are presented in graphical form in engineering units and have been analyzed only to the extent necessary to assure reasonableness and consistency

  12. Elastic and elastic-plastic behaviour of a piping system during blowdown - Comparison of measurement and calculation

    International Nuclear Information System (INIS)

    Petruschke, W.; Strunk, G.

    1987-01-01

    The investigations according to the system identification show that the piping model using beam theory and flexibility factors according to the Karman theory are adequate for evaluating natural frequencies, mode shapes, static displacements and stresses. The same accuracy can be seen by comparing the piping response due to blowdown within the elastic range. The simplified elastic-plastic analysis in general overestimates the maximum amplitudes while the frequency content is not simulated very well. For practical purposes, it can be an adequate tool in many cases. The elastic-plastic analysis is the most expensive procedure but gives also the best results. The use of beam elements with multilinear moment-curvature relationships results in a good approximation for the global behaviour (displacements). The strains according to this theory only include the beam deformation modes

  13. PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 166S

    International Nuclear Information System (INIS)

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

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 166S, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 166S was conducted to obtain thermal-hydraulic and CHF information in THTF bundle 1 with an intact hot leg. The primary purpose of this report is to make the reduced instrument responses during tests 166S available. These are presented in graphical form in engineering units and have been analyzed only to the extent necessary to ensure reasonableness and consistency

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

    International Nuclear Information System (INIS)

    Haynes, H.D.

    1990-01-01

    In support of the NRC-funded Nuclear Plant Aging Research (NPAR) program, Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of motor-operated valves (MOVs). As part of this work, ORNL participated in the gate valve flow interrruption blowdown (GVFIB) tests carried out in Huntsville, Alabama. The tests provided an excellent opportunity to evaluate signature analysis methods for determining the operability of MOVs under accident conditions. ORNL acquired motor current and torque switch shaft angular position signatures on two test MOVs during several GVFIB tests. The reduction in operating ''margin'' of both MOVs due to the presence of additional valve running loads imposed by high flow was clearly observed in motor current and torque switch angular position signatures. In addition, the effects of differential pressure, fluid temperature, and line voltage on MOV operations were observed and more clearly understood as a result of utilizing the signature analysis techniques. (orig.)

  15. Preliminary condensation pool experiments with steam using DN80 and DN100 blowdown pipes[VIDEO CAMERAS

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-03-01

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

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

    International Nuclear Information System (INIS)

    Calay, J.C.; Goffin, C.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  18. Development and verification of coupled fluid-structural dynamic codes for stress analysis of reactor vessel internals under blowdown loading

    International Nuclear Information System (INIS)

    Krieg, R.; Schlechtendahl, E.G.

    1977-01-01

    YAQUIR has been applied to large PWR blowdown problems and compared with LECK results. The structural model of CYLDY2 and the fluid model of YAQUIR have been coupled in the code STRUYA. First tests with the fluid dynamic systems code FLUST have been successful. The incompressible fluid version of the 3D coupled code FLUX for HDR-geometry was checked against some analytical test cases and was used for evaluation of the eigenfrequencies of the coupled system. Several test cases were run with the two phase flow code SOLA-DF with satisfactory results. Remarkable agreement was found between YAQUIR results and experimental data obtained from shallow water analogy experiments. A test for investigation of nonequilibrium twophase flow dynamics has been specified in some detail. The test is to be performed early 1978 in the water loop of the IRB. Good agreement was found between the natural frequency predictions for the core barrel obtained from CYLDY2 and STRUDL/DYNAL. Work started on improvement of the beam mode treatment in CYLDY2. The name of this modified version will be CYLDY3. The fluiddynamic code SING1, based on an advanced singularity method and applicable to a broad class of highly transient, incompressible 3D-problems with negligible viscosity has been developed and tested. It will be used in connection with the planned laboratory experiments in order to investigate the effect of the core structure on the blowdown process. Coupling of SING1 with structural dynamics is on the way. (orig./RW) [de

  19. Experiment data report for semiscale Mod-1 test S-01-1 (isothermal blowdown with core resistance simulator)

    International Nuclear Information System (INIS)

    Zender, S.N.; Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

    1975-04-01

    Recorded test data are presented for Test S-01-1 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-1 is one of several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is representative of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-1 was conducted from an initial temperature of 540 0 F and an initial pressure of 1596 psig. A simulated intermediate size double-ended hot leg break (0.00145 ft 2 break area on each end) was used to investigate the system response to a slow depressurization transient. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. During system depressurization, coolant was injected into the vessel downcomer inlet annulus to investigate the effectiveness of injection into the inlet annulus with respect to delivery of coolant to the lower plenum. Following the blowdown portion of Test S-01-1, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. The purpose of this report is to make available the uninterpreted data from Test S-01-1 for future data analysis and test results reporting activities. 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. (U.S.)

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

    International Nuclear Information System (INIS)

    Wang, Yan; Zheng, Yanhua; Li, Fu; Shi, Lei

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Moesinger, H.

    1979-08-01

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

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

    International Nuclear Information System (INIS)

    Waaranperae, Y.; Nilsson, L.; Gustafsson, P.Aa.; Jonsson, N.O.

    1979-06-01

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  4. Heat transfer correlation development and assessment: a summary and assessment of return to nucleate boiling phenomena during blowdown tests conducted at the Idaho National Engineering Laboratory (INEL)

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.

    1979-04-01

    The data are presented which were obtained in Loss-of-Coolant Experiments (LOCE) at Idaho National Engineering Laboratory (INEL) which demonstrate the presence of cladding rewetting after the critical heat flux has been exceeded as a viable cooling mechanism during the blowdown phase of a LOCE. A brief review of the mechanisms associated with the boiling crisis and rewetting is also provided. The relevance of INEL LOCE rewetting data to nuclear reactor licensing Evaluation Model Requirements is considered, and the conclusion is made that the elimination of rewetting and return to nucleate boiling (RNB) in Evaluation Models represents a definite conservatism

  5. Application of analytical capability to predict rapid cladding cooling and quench during the blowdown phase of a large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Aksan, S.N.; Tolman, E.L.; Nelson, R.A.

    1983-01-01

    Large-break Experiments L2-2 and L2-3 conducted in the Loss-of-Fluid Test (LOFT) facility experienced core-wide rapid quenches early in the blowdown transients. To further investigate rapid cladding quenches, separate effects experiments using Semiscale solid-type electric heater rods were conducted in the LOFT Test Support Facility (LTSF) over a wide range of inlet coolant conditions. The analytical capability to predict the cladding temperature response from selected LTSF experiments estimated to bound the hydraulic conditions causing the LOFT early blowdown quenches was investigated using the RELAP4 computer code and was shown to be acceptable over the film boiling cooldown phase. This analytical capability was then used to investigate the behavior of nuclear fuel rods under the same hydraulic conditions. The calculations show that, under rapid cooling conditions, the behaviors of nuclear and electrical heater rods are significantly different because the nuclear rods are conduction limited, while the electrical rods are convection limited

  6. Experiment data report for semiscale Mod-1 Tests S-01-4 and S-01-4A (isothermal blowdown with core resistance simulator)

    International Nuclear Information System (INIS)

    Zender, S.N.; Jensen, M.F.; Sackett, K.E.

    1975-03-01

    Recorded test data are presented for Tests S-01-4 and S-01-4A of the semiscale Mod-1 isothermal blowdown test series. These tests are among several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is representative of LOFT design based on volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Tests S-01-4 and S-01-4A employed an intact loop resistance that was similar to that of Test S-01-3 and low relative to that of Test S-01-2. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The tests were initiated at initial isothermal conditions of about 2250 psig and 540 0 F by a simulated offset shear of the cold-leg broken-loop piping. During system depressurization, coolant was injected into the cold leg of the intact loop to provide data on the effects of emergency core cooling on system response. Following the blowdown portion of Test S-01-4, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. The uninterpreted data are presented. 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. (U.S.)

  7. 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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  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)

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

    2008-01-01

    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 H 2 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 CO 2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using 'transparent' assumption for the CO 2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA

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

    Science.gov (United States)

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

    2014-12-01

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

  11. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  12. Mathematical aspects of reactor blowdown

    International Nuclear Information System (INIS)

    Esposito, V.J.

    1975-01-01

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

  13. Measurement of blowdown flow rates using load cells

    International Nuclear Information System (INIS)

    Dolas, P.K.; Venkat Raj, V.; Ghosh, A.K.; Murty, L.G.K.; Muralidhar Rao, S.

    1980-01-01

    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.)

  14. Fundamental Studies in Blow-Down and Cryogenic Cooling

    Science.gov (United States)

    1993-09-01

    Mudawar , I. and Anderson, T.M., -High Flux Electronic Cooling by Means of Pool Boiling - Part I: Parametric Investigation of the Effects of Coolant...Electronics, pp. 25-34, 1989. 30 Mudawar , I. and Anderson, T.M., "High Flux Electronic Cooling by Means of Pool Boiling - Part 1I: Optimization of

  15. Performance of the CNEN MHD Blow-Down Loop Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bertolini, E.; Brown, R.; Gasparotto, M.; Gay, P.; Toschi, R. [Laboratorio Conversione Diretta, CNEN, Frascati (Italy)

    1968-11-15

    The CNEN facility has been designed, manufactured and used for alkali-seeded noble gas MHD energy conversion research, as the major experimental effort during the first five-year CNEN Research Programme on MHD. The main specifications and the general arrangement with information on preliminary commissioning tests of some components were given at the Salzburg Symposium. Since then the facility has been successfully commissioned and from March 1967 has been working on MHD experiments. Efforts were made to reduce any adverse effects on the experimental MHD results that were due to inherent limitations of an experimental apparatus (particularly under open-circuit conditions). Great emphasis was placed on problems of caesium vaporization and the mixing with helium, the purity level of the mixture, measurements and the control system. The insulation of the plasma from ground was carefully treated, increasing the ratio between insulator resistance and typical plasma resistance as much as possible. Fluidynamic tests at room and high temperatures have shown that stability in the gas parameters (temperature, pressure and mass flow) can be maintained within few per cent for tens of seconds after a transient, giving a behaviour similar to a continuously running system. The high- temperature, alumina pebble-bed heater has successfully operated, bringing the helium-caesium mixtures up to 2000 Degree-Sign K and up to 4 atm abs pressure, and undergoing seven thermal cycles, for a total of more than 2000 hours operation at top temperature. Preheated generator ducts using alumina as insulator and tantalum for electrodes performed satisfactorily, very much attention having been given in the design to reduction of thermal shocks and to obviating possible paths for caesium leakage and short-circuiting of electrode leads. The pulsed liquid nitrogen precooled magnet has been run for about 50 pulses at high field ( Asymptotically-Equal-To 4.5 tesla) with an operating time of about 10 seconds per pulse. The performance of each item and the whole performance of the facility as an experimental apparatus for closed-cycle MHD research is discussed and the technical solutions adopted for plasma insulation from ground are analysed in detail. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  17. Analytical modeling of bwr safety relief valve blowdown phenomenon

    International Nuclear Information System (INIS)

    Hwang, J.G.; Singh, A.

    1984-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. Two-dimensional, two-phase jet loading on containment structures during blowdown

    International Nuclear Information System (INIS)

    Mohammadian, S.; Slegers, L.

    1983-01-01

    Pressure profiles of impinging jets are calculated using the computer code BEACON/MOD3. The code is used in post - as well as precalculations of experiments to demonstrate its applicability in 2-phase jet load calculation. Comparisons between measurements and predictions show that the code predicts pressure profiles within 15% accuracy. (orig./RW)

  20. Analysis of select Mod-1 semiscale blowdown heat transfer tests. Final report

    International Nuclear Information System (INIS)

    Irani, A.A.; Fujita, N.; Mecham, D.C.; Ching, J.T.; Gose, G.C.; Hentzen, R.D.; Sawtelle, G.R.; Moore, K.V.

    1976-10-01

    The report contains the RELAP4 analysis and sensitivity studies of Semiscale Tests S-02-2 and S-02-7. The Semiscale System is an electrically heated experiment designed to produce data on system performance typical of PWR thermal-hydraulic behavior. The RELAP4 program used for these analyses is a digital computer program developed to predict the thermal-hydraulic behavior of experimental systems and water-cooled nuclear reactors subjected to postulated transients. The results of the analysis for Test S-02-2 were in very good agreement with the data. Two parameters which required improvement were identified. These were the lower plenum density and the mass flow on the vessel side of the break. Subsequently, before analyzing Test S-02-7, the lower plenum was renodalized and the critical flow model at the vessel side break was modified. The results of the analysis of Test S-02-7 compared more favorably with the data than those of S-02-2. Additional sensitivity studies included time step studies, steam generator and downcomer modeling, and core nodalization

  1. Practical affairs of energy management: Operation management of cooling tower during winter and blow-down

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H.K. [Cheonsu Industry Co, Seoul (Korea, Republic of)

    1998-02-01

    In case the cooling tower is used throughout the year, operation managers should be careful to make sure that freezing at the surface of the loop in outside air inlet and the freezing of the reservoir due to the drop of ambient temperature during winter operation, or, freezing of condensed water drops on the interior surface of the fan blower cylinder, does not cause any plight that makes the original function not work to its capacity. To minimize the hindrance from freezing during winter, operation should be fully reviewed at the planning stage of the cooling tower. Those cooling towers used in the north of Central region and Kangwon Province should be especially taken consideration for heavy snowfall and severe cold. 6 figs., 1 tab.

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

    International Nuclear Information System (INIS)

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

    1983-08-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  4. The effects of transient conditions on the onset of intermittent dryout during blowdown

    Energy Technology Data Exchange (ETDEWEB)

    Statham, B.A., E-mail: stathaba@mcmaster.ca; Novog, D.R., E-mail: novog@mcmaster.ca

    2017-06-15

    Highlights: • This papers presents the results of an experimental investigation of transient critical heat flux in high quality and intermediate pressure water. • In existing literature conclusions vary from those showing no effect of transient conditions to results which show 30–40% improvement in CHF. • Along with new CHF data points in the liquid film dominated flow regime, the authors provide a methodology for producing bias free estimates of CHF based on existing correlations. • With these bias free CHF estimates, comparisons are made between transient and steady-state CHF at comparable local conditions. • The work concludes that based on consistently collected and analyzed data that quasi-steady CHF experiments adequately predict transient CHF using the same local thermalhydraulic conditions. - Abstract: For a given set of conditions in a boiling system the point of liquid film dryout or departure from nucleate boiling corresponds to the change from convective or nucleate boiling to transition or film boiling. This change is associated with a rapid deterioration of the heat transfer coefficient and the heat flux at this transition is denoted the critical heat flux (CHF). Computer models used to predict station transients and CHF rely heavily on empirical correlations to predict the CHF. Liquid film CHF data are usually obtained using a quasi-steady method wherein the heat flux is incremented in small steps with each step being allowed to reach a new equilibrium until an abnormal temperature increase is detected on the experimental surfaces. In applying a correlation derived from steady-state experiments to transient analyses these codes implicitly assume that dryout will occur for the same local conditions during transients as during steady state conditions. There is some disagreement in literature as to the validity of this hypothesis. This paper provides new steady-state and transient experimental data for CHF in water at intermediate pressures and demonstrates that for high-quality CHF that indeed CHF occurs at comparable local conditions during transients and steady-sate.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bertolini, E.; Gasparotto, M.; Gay, P.; Toschi, R. [Laboratorio Conversione Diretta, CNEN, Frascati (Italy)

    1968-11-15

    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 Degree-Sign 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 cm{sup 2}, 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 uB{beta}L (50% at {beta} = 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)

  6. Prediction of the semiscale blowdown heat transfer test S-02-8 (NRC Standard Problem Five)

    International Nuclear Information System (INIS)

    Fujita, N.; Irani, A.A.; Mecham, D.C.; Sawtelle, G.R.; Moore, K.V.

    1976-10-01

    Standard Problem Five was the prediction of test S-02-8 in the Semiscale Mod-1 experimental program. The Semiscale System is an electrically heated experiment designed to produce data on system performance typical of PWR thermal-hydraulic behavior. The RELAP4 program used for these analyses is a digital computer program developed to predict the thermal-hydraulic behavior of experimental systems and water-cooled nuclear reactors subjected to postulated transients. The RELAP4 predictions of Standard Problem 5 were in good overall agreement with the measured hydraulic data. Fortunately, sufficient experience has been gained with the semiscale break configuration and the critical flow models in RELAP4 to accurately predict the break flow and, hence the overall system depressurization. Generally, the hydraulic predictions are quite good in regions where homogeneity existed. Where separation effects occurred, predictions are not as good, and the data oscillations and error bands are larger. A large discrepancy existed among the measured heater rod temperature data as well as between these data and predicted values. Several potential causes for these differences were considered, and several post test analyses were performed in order to evaluate the discrepancies

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

    International Nuclear Information System (INIS)

    Reocreux, M.

    1980-05-01

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

    Science.gov (United States)

    Oldfield, C.; Peterson, C. J.

    2017-12-01

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

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

    Science.gov (United States)

    Mark D. Nelson; W. Keith Moser

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Zentis, A.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  12. Efficient computations of three-dimensional fluid-structure interactions during blowdown of a pressurized water reactor - FLUX

    International Nuclear Information System (INIS)

    Schumann, U.

    1979-01-01

    A numerical method (computer program FLUX) for investigation of the loads on pressure vessel internal structures during a loss-of-coolant accident of a PWR is described. In particular, the deformation of the core barrel are determined. Under operating conditions the core barrel controls the flow path in the vessel and consists mainly out of a relatively thin cylindrical shell. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Moses Lemuel Raj, G.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    1998-12-01

    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)

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

    International Nuclear Information System (INIS)

    Wachs, D. M.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Namatame, Ken; Kobayashi, Kensuke

    1975-12-01

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

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

    International Nuclear Information System (INIS)

    Bryce, W.M.

    1977-10-01

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

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

    International Nuclear Information System (INIS)

    Lee, J.H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    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

  18. Analysis of LOCA/LOECC with a non-stop CATHENA simulation

    International Nuclear Information System (INIS)

    Sabourin, G.; Huynh, H.M.

    1997-01-01

    This paper documents a new approach which simulates without interruption the blowdown and the post-blowdown portions of a LOCA/LOECC. The blowdown portion is simulated first with the pressures, enthalpies, and void fractions of the headers as boundary conditions. The transient inlet header flowrates are written to a file. The blowdown portion is then simulated again with the inlet header flowrates as boundary conditions. At the end of the blowdown, the flowrates are gradually changed to obtain the desired constant gas flowrate of the post-blowdown portion. This new approach was applied with CATHENA MOD3.5a Rev. 0 for a 20% reactor inlet header break coincident with a total loss of emergency core cooling injection. In summary, this paper shows a successful new approach where the blowdown and the post-blowdown portions of a large LOCA coincident with a total loss of emergency core cooling are simulated continuously. (author)

  19. 75 FR 58315 - Hazardous Waste Management System; Identification and Listing of Hazardous Waste; Direct Final...

    Science.gov (United States)

    2010-09-24

    ... landfill. The scrubber water blowdown will be managed in the waste water treatment plant (WWTP). The sludge... waste streams included in the petition were: the RKI fly ash, RKI bottom ash and RKI scrubber water... water blowdown waste resulting from the operations of the rotary kiln incinerator at its facility. B...

  20. Process fluid cooling system

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  1. 40 CFR 471.102 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ... treatment wet air pollution control scrubber blowdown. Subpart J—BAT Pollutant or pollutant property Maximum...) Mixing wet air pollution control scrubber blowdown. Subpart J—BAT Pollutant or pollutant property Maximum... degree of effluent reduction attainable by the application of the best available technology economically...

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

    International Nuclear Information System (INIS)

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

    1977-08-01

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

  3. 40 CFR 98.250 - Definition of source category.

    Science.gov (United States)

    2010-07-01

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

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

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Jo, Y. K.; Loo, J. S.; Lim, N. Y.

    2001-01-01

    There is about 80 m 3 /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

  5. Experimental and theoretical studies of transient boiling and two-phase flow during the depressurisation of a simple glass vessel

    International Nuclear Information System (INIS)

    Ardron, K.H.; Furness, R.A.; Hall, P.C.

    1976-11-01

    Blowdown experiments using a glass pressure vessel containing saturated water at 4 bars have been performed to assist interpretation of the results of large scale experiments and aid understanding of the physical processes involved. Results have shown the strong dependence of depressurisation time, phase distribution and mass flow rate on the length to diameter ratio of the exit pipe. Preliminary observations of the flow regime in the discharge pipe are consistent with predictions of the flow regime map of Mandhane, Gregory and Aziz 1974. Different flow regimes have been observed at different axial positions along the pipe. Bubble growth rates during the non-equilibrium phase of blowdown are shown to be in reasonable agreement with a simple convective heat flux analysis previously used in blowdown calculations. The transient pressure and liquid distribution in the vessel have been compared with calculations using the blowdown code RELAP-UK. (U.K.)

  6. Screening and evaluation of Second Half 1980 licensee event reports

    International Nuclear Information System (INIS)

    Waage, J.M.

    1981-09-01

    During the second part of 1980, two individual plant events occurred: Indian Point-2 (containment flooding) and Pilgrim-1 (uncontrolled blowdown). Significant event reports and update on generic problem areas and major equipment problem areas are included in this report

  7. 40 CFR 471.35 - Pretreatment standards for new sources (PSNS).

    Science.gov (United States)

    2010-07-01

    ... metal powder atomized Chromium 0.970 0.393 Nickel 1.44 0.970 Fluoride 156 69.2 (q) Annealing and... of process wastewater pollutant. (r) Wet Air Pollution Control Scrubber Blowdown. Subpart C—PSNS...

  8. 40 CFR 471.32 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ... powder atomized Chromium 0.970 0.393 Nickel 1.44 .970 Fluoride 156 69.2 (q) Annealing and solution heat... pollutants. (r) Wet air pollution control scrubber blowdown. Subpart C—BAT Pollutant or pollutant property...

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

    Science.gov (United States)

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

    2017-11-01

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

  10. 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)

    Hughes, G.; Mueller, R.

    1980-03-01

    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) [de

  11. 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,...

  12. Secondary circuit water chemistry and related problems with SG

    Energy Technology Data Exchange (ETDEWEB)

    Ignatov, V; Ivanov, V [Balakovo Nuclear Power Plant (Russian Federation)

    2001-07-01

    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.)

  13. Secondary circuit water chemistry and related problems with SG

    International Nuclear Information System (INIS)

    Ignatov, V.; Ivanov, V.

    2001-01-01

    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.)

  14. Experimental and theoretical investigation on the depressurization of a vessel with internals

    International Nuclear Information System (INIS)

    Vigni, P.; Oriolo, F.; Rosa, U.

    1978-01-01

    This paper is about some blow-down experiments performed at the Scalbatraio Center of the University of Pisa. The blow-down tests have been made to investigate the depressurization of a vessel with internal structures, reproducing the geometry of a BWR. The experimental data have been compared with calculations performed by the RELAP program, in order to evaluate the scaling effects related to their application to large scale units. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Finicle, D.P.

    1978-06-06

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

  16. Problems of two-phase flows in water cooled and moderated reactors

    International Nuclear Information System (INIS)

    Syu, Yu.

    1984-01-01

    Heat exchange in two-phase flows of coolant in loss of coolant accidents in PWR and BWR reactors has been investigated. Three main stages of accident history are considered: blowdown, reflooding using emergency core cooling system and rewetting. Factors, determining the rate of coolant leakage and the rate of temperature increase in fuel cladding during blowdown, processes of vapour during reflooding and liquid priming by vapour during rewetting, are discussed

  17. Secondary-water chemistry at Millstone 2

    International Nuclear Information System (INIS)

    Putkey, T.A.; Pearl, W.L.; Sawochka, S.G.

    1983-04-01

    Secondary system chemistry and steam generator corrosion observations at the Millstone 2 pressurized water reactor are summarized. Condenser retubing and retrofit of full-flow condensate polishers led to significant improvements in steam generator blowdown chemistry following observations of denting after one year of operation at elevated blowdown chloride levels. Notwithstanding the chemistry improvements, denting has continued but at a much reduced rate. In addition, extensive pitting of the Alloy 600 tubing between the tubesheet and first support plate has been reported recently

  18. Secondary coolant purification system

    International Nuclear Information System (INIS)

    Stiteler, F.Z.; Donohue, J.P.

    1978-01-01

    The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

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

    International Nuclear Information System (INIS)

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

    1996-07-01

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

  20. Radioactive waste treatment apparatus

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Okazaki, Motoaki; Namatame, Ken; Shiba, Masayoshi

    1979-08-01

    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)

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

    International Nuclear Information System (INIS)

    Maqbool, M. U.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Allen, M.D.; Pitch, M.; Nichols, R.T.

    1990-08-01

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  6. Development of analysis methodology for hot leg break mass and energy release

    International Nuclear Information System (INIS)

    Song, Jin Ho; Kim, Cheol Woo; Kwon, Young Min; Kim, Sook Kwan

    1995-04-01

    A study for the development of an analysis methodology for hot leg break mass and energy release is performed. For the blowdown period a modified CEFLASH-4A methodology is suggested. For the post blowdown period a modified CONTRAST boil-off model is suggested. By using these computer code improved mass and energy release data are generated. Also, a RELAP5/MOD3 analysis for finally the FLOOD-3 computer code has been modified for use in the analysis of hot leg break. The results of analysis using modified FLOOD-3 are reasonable as we expected and their trends are good. 66 figs., 8 tabs. (Author) .new

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

    International Nuclear Information System (INIS)

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

    1968-01-01

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

  8. Study of the HTGR fission product migration at the Osiris experimental reactor

    International Nuclear Information System (INIS)

    Homme, A. l'; Lucot, M.

    1977-01-01

    A program of study on accidents in HTR reactor operation is presented: blowdown of primary coolant circuit, water inlet into the primary circuit, fuel element overheating by pipe logging or loss of cooling. These studies will be made in Aida irradiation loop in the pool of the Osiris reactor [fr

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

    Science.gov (United States)

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

    2012-01-01

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

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

    Science.gov (United States)

    Dennis Neitzke

    2007-01-01

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

  11. Research and Development for Health and Environmental Hazard Assessment. Task Order 1. Development of Data Base Requirements for Human Health Based Water Quality Criteria for Military Recycle/Reuse Applications.

    Science.gov (United States)

    1980-06-01

    laundries,4,5,6, coin operated laundromats " aircraft and vehicle wash racks, plating shops, cooling tower blowdowns, 3 etc. Table 2 gives an example of...and M. Chilson, "Treatment of Laundromat Wastes," Final report to EPA, Office of Research and Monitoring, PB227- 369, September, 1971. 8. See G.G., K.K

  12. Behavior of four PWR rods subjected to a simulated loss-of-coolant accient in the power burst facility

    International Nuclear Information System (INIS)

    Cook, T.F.; Hagrman, D.L.; Sepold, L.K.

    1978-01-01

    Cladding deformation characteristics resulting from the first nuclear blowdown tests (LOC-11) conducted in the Power Burst Facility (PBF) are emphasized in this paper. The objective of the LOC-11 tests was to obtain data on the thermal, mechanical, and materials behavior of pressurized and unpressurized fuel rods when exposed to a blowdown similiar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The test hardware consisted of four separately shrouded fresh fuel rods of PWR 15 x 15 design. Initial plenum pressures ranged from atmospheric to 4.8 MPa (representative of end-of-life). During LOC-11C, the four fuel rods were subjected to 6.5 hours of nuclear operation at approximately 67 kW/m average rod power to cause decay heat build-up. Just before the start of blowdown, cladding surface temperatures were about 620 K and fuel centerline temperatures were in the 2500 to 2600 K range. During the 30-second blowdown transient, CHF occurred 2 seconds after initiation. Fuel centerline temperature dropped continuously, while cladding surface temperatures increased. Maximum cladding temperatures of 1030 to 1050 K occurred 15 seconds into the transient. Posttest destructive examination revealed cladding microstructures and oxide thicknesses consistent with the measured cladding temperatures. The cladding surface thermocouples did not appreciably affect cladding temperature distributuion (fin cooling effect) in the vicinity of the thermocouples

  13. Error analysis for 1-1/2-loop semiscale system isothermal test data

    International Nuclear Information System (INIS)

    Feldman, E.M.; Naff, S.A.

    1975-05-01

    An error analysis was performed on the measurements made during the isothermal portion of the Semiscale Blowdown and Emergency Core Cooling (ECC) Project. A brief description of the measurement techniques employed, identification of potential sources of errors, and quantification of the errors associated with data is presented. (U.S.)

  14. 40 CFR 98.253 - Calculating GHG emissions.

    Science.gov (United States)

    2010-07-01

    ... (metric tons/year). NCD = Number of atmospheric crude oil distillation columns at the facility. NPU1... = Methane emission rate from blowdown systems (mt CH4/year). QRef = Quantity of crude oil plus the quantity...) For storage tanks other than those processing unstabilized crude oil, you must either calculate CH4...

  15. 40 CFR 63.1082 - What definitions do I need to know?

    Science.gov (United States)

    2010-07-01

    ... stream that results from condensation of dilution steam (in the cracking furnace quench system), blowdown...) National Emission Standards for Ethylene Manufacturing Process Units: Heat Exchange Systems and Waste... system. The continuous butadiene waste stream does not include butadiene streams generated from sampling...

  16. 7 CFR 1767.20 - Plant accounts.

    Science.gov (United States)

    2010-01-01

    ... line wholly identified with items included herein. 11. Retaining walls. 12. Water conductors and... settings, water walls, arches, grates, insulation, blowdown system, drying out of new boilers, also... disassembly machinery. 12. Reactor fuel element failure detection system. 13. Reactor emergency poison...

  17. 40 CFR 471.31 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

    ... 51.1 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Annealing and solution heat... wastewater pollutants. (r) Wet air pollution control scrubber blowdown. Subpart C—BPT Pollutant or pollutant...

  18. 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.

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

    International Nuclear Information System (INIS)

    Krutzik, Norbert

    2002-01-01

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

  20. 40 CFR 266.103 - Interim status standards for burners.

    Science.gov (United States)

    2010-07-01

    ...) Minimum scrubber blowdown from the system or maximum suspended solids content of scrubber water; and (C) Minimum pH level of the scrubber water; (x) For systems using venturi scrubbers, the minimum differential...) For systems using wet scrubbers, including wet ionizing scrubbers (unless complying with the Tier I or...

  1. 40 CFR 63.1209 - What are the monitoring requirements?

    Science.gov (United States)

    2010-07-01

    ... ionizing wet scrubbers, high energy wet scrubbers such as venturi, hydrosonic, collision, or free jet wet... is equipped with a wet scrubber, you must comply with the following unless you document in the... efficiency during the test: (1) Scrubber blowdown must be minimized during a pretest conditioning period and...

  2. 40 CFR 471.11 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

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

  3. 40 CFR 471.12 - Effluent limitations representing the degree of effluent reduction attainable by the application...

    Science.gov (United States)

    2010-07-01

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

  4. Geomorphological impacts of a tornado disturbance in a subtropical forest

    Science.gov (United States)

    Jonathan Phillips; Daniel A. Marion; Chad Yocum; Stephanie H. Mehlhope; Jeff W. Olson

    2015-01-01

    We studied tree uprooting associated with an EF2 tornado that touched down in portions of the Ouachita Mountains in western Arkansas in 2009. In the severe blowdown areas all trees in the mixed shortleaf pine–hardwood forest were uprooted or broken, with no relationship between tree species or size and whether uprooting or breakage occurred. There was also no...

  5. Posteruption arthropod succession on the Mount St. Helens volcano: the ground-dwelling beetle fauna (Coleoptera).

    Science.gov (United States)

    R.R. Parmenter; C.M. Crisafulli; N. Korbe; G. Parsons; M. Edgar; J.A. MacMahon

    2005-01-01

    The 1980 eruptions of Mount St. Helens created a complex mosaic of disturbance types over a 600 km2 area. From 1980 through 2000 we monitored beetle species relative abundance and faunal composition of assemblages at undisturbed reference sites and in areas subjected to tephra-fall, blowdown, and pyroclastic flow volcanic disturbance. We...

  6. 77 FR 41814 - Entergy Operations, Inc.; Grand Gulf Nuclear Station, Unit 1

    Science.gov (United States)

    2012-07-16

    ... result of the EPU, which will also results in an increase in water loss through evaporation, blowdown... supply piping to the plant service water header, discharge piping into the river, and electrical... only three public water supply systems in the State of Mississippi that use surface water as a source...

  7. Effects of timber harvest on aquatic vertebrates and habitat in the North Fork Caspar Creek

    Science.gov (United States)

    Rodney J. Nakamoto

    1998-01-01

    I examined the relationships between timber harvest, creek habitat, and vertebrate populations in the North and South forks of Caspar Creek. Habitat inventories suggested pool availability increased after the onset of timber harvest activities. Increased large woody debris in the channel was associated with an increase in the frequency of blowdown in the riparian...

  8. Heat-flux gage measurements on a flat plate at a Mach number of 4.6 in the VSD high speed wind tunnel, a feasibility test (LA28). [wind tunnel tests of measuring instruments for boundary layer flow

    Science.gov (United States)

    1975-01-01

    The feasibility of employing thin-film heat-flux gages was studied as a method of defining boundary layer characteristics at supersonic speeds in a high speed blowdown wind tunnel. Flow visualization techniques (using oil) were employed. Tabulated data (computer printouts), a test facility description, and photographs of test equipment are given.

  9. Performance evaluation of a drag-disc turbine transducer and three-beam gamma densitometer under transient two-phase flow conditions

    International Nuclear Information System (INIS)

    Nalezny, C.L.; Chen, L.L.; Solbrig, C.W.

    1979-01-01

    One of the primary variables measured in the Loss-of-Fluid Test (LOFT) Program is mass flow rate. LOFT uses drag-disc turbine tranducers (DTT) and a three-beam gamma densitometer to measure parameters from which mass flow may be computed. The transducer combination was performance tested under transient conditions in the blowdown loop at the LOFT Test Support Facility (LTSF). The performance tests consisted of three partial blowdowns of different durations starting from the same initial conditions. The reference mean mass flow rate was determined by measuring the amount of water required to reestablish initial conditions after each partial blowdown. The average mass flow rates computed from the output of the drag disc, turbine, and gamma densitometer were compared to the reference mean mass flow rates over three blowdown intervals. The tests indicated that the equal phase velocity mass measurement model provided excellent results through the use of the turbine and densitometer, and drag disc and densitometer when the phase velocities were nearly equal

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

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    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

  11. 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)

    Abouhadra, D.S.; Byrne, J.E.

    2003-01-01

    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

  12. Fluid structure coupling algorithm

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.; Landram, C.S.; Quinones, D.F.

    1980-01-01

    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 have been extended to three dimensions and implemented in the computer code PELE-3D

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

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1975-01-01

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

  14. FLUST-2D - A computer code for the calculation of the two-dimensional flow of a compressible medium in coupled retangular areas

    International Nuclear Information System (INIS)

    Enderle, G.

    1979-01-01

    The computer-code FLUST-2D is able to calculate the two-dimensional flow of a compressible fluid in arbitrary coupled rectangular areas. In a finite-difference scheme the program computes pressure, density, internal energy and velocity. Starting with a basic set of equations, the difference equations in a rectangular grid are developed. The computational cycle for coupled fluid areas is described. Results of test calculations are compared to analytical solutions and the influence of time step and mesh size are investigated. The program was used to precalculate the blowdown experiments of the HDR experimental program. Downcomer, plena, internal vessel region, blowdown pipe and a containment area have been modelled two-dimensionally. The major results of the precalculations are presented. This report also contains a description of the code structure and user information. (orig.) [de

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

    International Nuclear Information System (INIS)

    Miner, R.M.; Warrick, J.W.

    1975-01-01

    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.)

  16. Fluid-structure-coupling algorithm

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.; Landram, C.S.; Quinones, D.F.

    1980-01-01

    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. A coupled RELAPS-3D/CFD methodology with a proof-of-principle calculation; TOPICAL

    International Nuclear Information System (INIS)

    Aumiller, D.L.; Tomlinson, E.T.; Bauer, R.C.

    2000-01-01

    The RELAP5-3D computer code was modified to make the explicit coupling capability in the code fully functional. As a test of the modified code, a coupled RELAP5/RELAP5 analysis of the Edwards-O'Brien blowdown problem was performed which showed no significant deviations from the standard RELAP5-3D predictions. In addition, a multiphase Computational Fluid Dynamics (CFD) code was modified to permit explicit coupling to RELAP5-3D. Several calculations were performed with this code. The first analysis used the experimental pressure history from a point just upstream of the break as a boundary condition. This analysis showed that a multiphase CFD code could calculate the thermodynamic and hydrodynamic conditions during a rapid blowdown transient. Finally, a coupled RELAP5/CFD analysis was performed. The results are presented in this paper

  18. Progress report on LOFT rake designs

    International Nuclear Information System (INIS)

    Bearden, R.G.

    1977-01-01

    Evaluation of data from Loss-of-Fluid Test (LOFT) nonnuclear tests has shown a need for profile measurements at several locations in the LOFT piping. A prototype rake consisting of three Drag-Disc Turbine Transducers (DTT) has been designed and fabricated for installation at one location (FE-BL-1) in the blowdown loop. After successful operation during a LOFT nonnuclear test (L1-4) scheduled for May, 1977, additional rakes will be installed in the primary and blowdown loops. A research program to develop a pitot tube rake for measurement of steady state and transient two-phase flows is in progress at McMaster University, Hamilton, Ontario. A rake of thermocouples and pitot tubes will be developed for installation near the emergency core coolant (ECC) injection points

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  20. Loss-of-coolant accident test series TC-1 experiment operating specifications

    International Nuclear Information System (INIS)

    Yackle, T.R.

    1979-09-01

    The purpose of this document is to specify the experiment operating procedure for the test series TC-1. The effects of externally mounted cladding thermocouples on the fuel rod thermal behavior during LOCA blowdown and reflood cycles will be investigated in the test. Potential thermocouple effects include: (a) delayed DNB, (b) momentary cladding rewets following DNB, (c) premature cladding rewet during a blowdown two-phase slug period, and (d) early cladding rewet during reflood. The two-phase slug period will be controlled by momentarily opening the hot leg valve. The slug will consist of lower plenum liquid that is sent through the flow shrouds and will be designed to quench the fuel rods at a rate that is similar to the slug experienced early in the LOFT L2-2 and L2-3 tests

  1. Vitrification of hazardous and mixed wastes

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. RELAP5/MOD2 code assessment using a LOFT L2-3 loss of coolant experiment

    International Nuclear Information System (INIS)

    Bang, Young Seok; Chung, Bub Dong; Kim, Hho Jung

    1990-01-01

    The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of the PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core in a reasonable range and that the code had deficiencies in the critical flow model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. A Sensitivity calculation with an updated version from RELAP5/MOD2 Cycle 36.04 improved the prediction of the rewet phenomena

  3. Prediction of thermal-Hydraulic phenomena in the LBLOCA experiment L2-3 using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Bang, Young Seok; Chung, Bub Dong; Kim, Hho Jung

    1991-01-01

    The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical flow model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5/MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-01-01

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

  5. Eigenvibration measurement of the condensation of the GKSS pressure suppression test rig

    International Nuclear Information System (INIS)

    Boettcher, G.; Kolb, M.

    1981-01-01

    A modal analysis - which characterizes a structure's vibration modes by resonant frequency, damping and shape vector - was undertaken for the wetwell of the GKSS PSS facility in order to better explain the periodicities measured for the pool pressure and for the wetwell wall movements in blowdown experiments. The wetwell was hit at one point by a sledge hammer instrumented with a force transducer accelerometers were moved to all points to be included in the shape vector which was obtained by a computer-aided modal analysis system. Six global modes of the wetwell were identified. The frequencies of the three lowest modes (40, 52, 78 Hz) correspond plausibly to frequencies observed during blowdown experiments. (orig.) [de

  6. Transient analysis of DTT rakes

    International Nuclear Information System (INIS)

    Kamath, P.S.; Lahey, R.T. Jr.

    1981-01-01

    This paper presents an analytical model for the determination of the cross-sectionally averaged transient mass flux of a two-phase fluid flowing in a conduit instrumented by a Drag-Disk Turbine Transducer (DTT) Rake and a multibeam gamma densitometer. Parametric studies indicate that for a typical blowdown transient, dynamic effects such as rotor inertia can be important for the turbine-meter. In contrast, for the drag-disk, a frequency response analysis showed that the quasisteady solution is valid below a forcing frequency of about 10 Hz, which is faster than the time scale normally encountered during blowdowns. The model showed reasonably good agreement with full scale transient rake data, where the flow regimes were mostly homogeneous or stratified, thus indicating that the model is suitable for the analysis of a DTT rake. (orig.)

  7. Improved water chemistry controls for minimizing degradation of materials

    International Nuclear Information System (INIS)

    Sawochka, S.G.

    1986-01-01

    The Electric Power Research Institute and the Steam Generator Owners Group have sponsored several efforts to develop secondary water chemistry guidelines to minimize pressurized water reactor (PWR) steam generator tubing degradation. To develop these guidelines, chemical species known to accelerate corrosion of Alloy 600 were identified, and values for normal and abnormal chemistry situations were established. For example, sodium hydroxide was known to accelerate Alloy 600 intergranular attack stress corrosion cracking; thus, guidelines were developed for blowdown sodium concentrations in recirculating steam generator systems. Similarly, formation of acidic solutions, particularly as a result of chloride ingress at seawater sites, was known to accelerate denting; thus, chloride guidelines were established. A blowdown cation conductivity limit was established to minimize concentrations of other anionic species. Guidelines also were developed for condensate and feedwater chemistry to minimize general corrosion of system materials, thereby minimizing sludge and deposit buildup in the steam generators

  8. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

  9. Computer codes for the study of the loss of coolant accident of PWR reactors

    International Nuclear Information System (INIS)

    Gomolinski, M.; Menessier, D.; Tellier, N.

    1975-01-01

    The CEA has undertaken a large programme to study the consequence on the core of the LOCA of a PWR. In the programme, simultaneously carried out experiments and the development of the calculations means are described. Several experiments such as OMEGA, ERSEC and PHEBUS tests, which provide data to check the computer codes are outlined briefly in the paper. For analysis of the LOCA of a PWR, a series of computer codes, which are at present in use or under development, are linked with each other. The codes are DANAIDES for blowdown, CERES for refill and reflood, THETA-1B and FLIRA for heat up calculation during the blow-down and the reflooding period respectively. FLIRA-PASTEL, a combination of FLIRA and PASTEL which calculate the stress and deformations of material using the finite element method, will be used in place of FLIRA. The basic models and flowcharts of the above codes are described in the paper

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  11. Characterization of Transient Plasma Ignition Flame Kernel Growth for Varying Inlet Conditions

    Science.gov (United States)

    2009-12-01

    from Intercity Manufacturing. Without their expertise in precision machining this thesis would not have been possible. Their countless hours spent...somewhere within the combustor due to the time required to produce the required conditions, and will be travelling at near Mach 5 speeds for most...atmospheric pressure. This sudden drop in pressure creates a rarefaction wave that travels forward in the combustor. The blowdown time for a 1 meter long

  12. Early response of pressurized hot water in a pipe to a sudden break. Final report

    International Nuclear Information System (INIS)

    Alamgir, M.; Kan, C.Y.; Lienhard, J.H.

    1981-06-01

    Experimental and analytic studies that explain the details of early pressure variations during rapid depressurization in water-cooled reactors are presented as a means of assessing sudden break consequences in a coolant pipe. The report includes (1) a description of the experiment, (2) an analysis of the new bubble growth law for thermally controlled growth of vapor bubbles in an exponentially-varying pressure field, and (3) a review of previous studies and additional observations of blowdown behavior

  13. Development of Envelope Curves for Predicting Void Dimensions from Overturned Trees

    Science.gov (United States)

    2014-07-01

    Engineers Washington , DC 20314-1000 ERDC/GSL TR-14-27 ii Abstract To evaluate the potential influence of a tree on embankment stability, it is...4.5 Consistent influence of tree diameter and species on damage in nine eastern North America tornado blowdowns (Peterson 2007...likely to be blown down than deciduous trees . Phillips et al. (2008) independently came to the same conclusion that conifers have a higher

  14. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  15. Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory, August 1977

    International Nuclear Information System (INIS)

    Fee, G.G.

    1977-01-01

    Technical highlights are presented for the following safety-related studies: heavy section steel technology, fission product beta and gamma energy release, fission product release from LWR fuel, fission product transport tests, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, design criteria for piping and nozzles, and noise diagnostics for safety assessment

  16. Sargent and Lundy containment tests revisited

    International Nuclear Information System (INIS)

    Henry, Robert E.; Hammersley, Robert J.

    2005-01-01

    The pressurization experiments performed in the intermediate scale Sargent and Lundy containment test facility provide numerous insights into the dominant heat and mass transfer processes under design basis accident conditions similar to a large break Loss of Coolant Accident (LOCA). These experiments were the first integral tests to examine the containment response to a dynamic blowdown from the Reactor Coolant System (RCS). Measurements included the blowdown rate of the simulated Reactor Pressure Vessel (RPV), the pressure in containment as well as the containment temperatures in the top and bottom of the containment vessel. Furthermore, various experiments were performed with the blowdown location changed from the vessel bottom to the lower third of the vessel, the upper third of the vessel and near the top of the RPV to examine the influence of different types of break elevations, i.e. different characterizations of the exhausting steam-water mixture. Perhaps the most insightful set of measurements from these experiments were those that varied the cold water mass initially resident in the bottom of the simulated containment vessel. The role of this water as a function of its initial mass and the break location showed substantial influence of this water if the blowdown location provided sufficient energy to disperse this cold water into the containment building atmosphere. This is demonstrated in Figure 1 taken from Kolflat, 1960. All of these are relevant to an understanding of the dominant physical processes for this type of postulated accident condition. As such, it is important that all of these insights are retained and used in models for the containment building thermal-hydraulic response under accident conditions. Reference: Kolflat, A., 1960, 'Resulting of 1959 Nuclear Power Plant Containment Test', Sargent and Lundy Report SL-1800; Kolflat, A. and Chittenden, W. A., 1957, 'A New Approach to the Design of Containment Shells for Atomic Power Plants

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

    International Nuclear Information System (INIS)

    Adams, J.P.

    1982-01-01

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

  18. Transient two-phase flow

    International Nuclear Information System (INIS)

    Hsu, Y.Y.

    1974-01-01

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

  19. 40 CFR 471.14 - Pretreatment standards for existing sources (PSES).

    Science.gov (United States)

    2010-07-01

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

  20. Lignocellulosic Biomass to Ethanol Process Design and Economics Utilizing Co-Current Dilute Acid Prehydrolysis and Enzymatic Hydrolysis for Corn Stover

    Science.gov (United States)

    2002-06-01

    is pumped from the blowdown tank into the hydrolyzate mixing tank (T-205) and mixed with recycled filtrate liquor (from tank T-213) to reduce the...slurrying tank (T-232) where it is mixed with conditioned hydrolyzate liquor and additional recycle water. Product liquor to be conditioned is pumped from...lignin-derived compounds. These reactions detoxify the hydrolyzate and the detoxification is more efficient at higher pH. Investigations into the

  1. Simulation model of dynamical behaviour of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Planchard, J.

    1994-01-01

    This report briefly describes the homogenized dynamical equations of a tube bundle placed in a perfect irrotational fluid, on case of small displacements. This approach can be used to study the mechanical behaviour of fuel assemblies of PWR reactor submitted to earthquake or depressurization blow-down. The numerical calculations require to define the added mass matrix of the fuel assemblies, for which the principle of computation is presented. (author). 14 refs., 4 figs

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

    International Nuclear Information System (INIS)

    Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

    1980-01-01

    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

  3. Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Fee, G.G.

    1976-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Belda, W.

    1977-01-01

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

  7. Vermont Yankee Nuclear Power Station. Annual operating report: January--December 1976

    International Nuclear Information System (INIS)

    1977-01-01

    Net electrical energy generated was 3,260,016 MWH with the facility on line for 6,776 hrs. Information is presented concerning operation, procedure changes, tests, experiments, plant changes, corrective maintenance, license event reports, forced power reductions, shutdowns, personnel radiation exposures, use of chemicals, plant discharges, cooling tower blowdown, traveling screens, fish impingement, reactor start-up, scram reports, and primary coolant chemistry

  8. ORNL: PWR-BDHT analysis procedure, a preliminary overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The computer programs currently used in the analysis of the ORNL-PWR Blowdown Heat Transfer Separate-Effects Program are overviewed. The current linkages and relationships among the programs are given along with general comments about the future directions of some of these programs. The overview is strictly from the computer science point of view with only minimal information concerning the engineering aspects of the analysis procedure

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Fee, G.G.

    1975-04-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  12. Study of low cost eco-friendly compounds as corrosion inhibitors for cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Farooqi, I H; Hussain, A; Saini, P A [AMU, Aligarh (India). Dept. of Civil Engineering; Quraishi, M A [AMU, Aligarh (India). Dept. of Applied Chemistry

    1999-07-01

    Attempts are made to utilize the aqueous extracts of natural compounds, namely cordia latifolia and curcumin, as corrosion inhibitors for mild steel in cooling systems, and their inhibition efficiencies are compared with that of Hydroxyethylidene 1-1 diphosphonic acid (HEDP). HEDP is also blended with aqueous extracts of natural compounds so as to improve their inhibition efficiency. The blowdown of the cooling system is also analysed for environmental factors. (author)

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

    International Nuclear Information System (INIS)

    Fee, G.G.

    1976-10-01

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

  14. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

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

  15. 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

  16. Development of electromagnetic filtration in the feed water circuits

    International Nuclear Information System (INIS)

    Dolle, L.

    1980-01-01

    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 [fr

  17. Documentation of CATHENA input files for the APOLLO computer

    International Nuclear Information System (INIS)

    1988-06-01

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

  18. Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Fee, G.G.

    1976-02-01

    Brief highlights are presented for the following activities: heavy section steel technology program, fission product β and γ energy release, LOCA release from LWR fuel, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, and design criteria for piping and nozzles

  19. Development of numerical methods for thermohydraulic problems in reactor safety

    International Nuclear Information System (INIS)

    Chabrillac, M.; Kavenoky, A.; Le Coq, G.; L'Heriteau, J.P.; Stewart, B.; Rousseau, J.C.

    1976-01-01

    Numerical methods are being developed for the LOCA calculation; the first part is devoted to the BERTHA model and the associated characteristic treatment for the first seconds of the blowdown, the second part presents the problems encountered for accounting for velocity difference between phases. The FLIRA treatment of the reflooding is presented in the last part: this treatment allows the calculation of the quenching front velocity

  20. HDR-investigations of check valve closure and resultant water hammer effects

    International Nuclear Information System (INIS)

    Scholl, K.D.

    1983-01-01

    The presented investigations are based on the Loss of Coolant Accident (LOCA). They concentrate on the first blowdown phase after pipe break of a feedwater line. The effect of such a break is moderated by quick closing check valves, by which the loss of coolant water is reduced and optimal post accident conditions are obtained. Unfortunately the closure of the valve can cause high pressure peaks (water hammer effects) in the feedwater system which potentially could produce safety relevant secondary damage. The system loading by these effects has been analysed. The HDR-Investigation-results led to an improvement of the feedwater system safety by verifying damping measures of quick closing check valves. Pressure peaks obtained with undamped valves in the range of 300 bars, are reduced to zero or a few bars above the normal operation pressure in feedwater systems. For the analytical simulation of valve closure the following dominant acting forces are identified: the blowdown flow resistance of the valve cone and the damping pistong force. The analytical description and quantification of the forces depends on blowdown flow and valve friction parameters. These have been evaluated and are presented for practical use. (orig.)

  1. LOFT fuel rod surface temperature measurement testing

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

    1978-01-01

    Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

  2. Pressure suppression pool mixing in passive advanced BWR plants

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  3. Vibration phenomena in large scale pressure suppression tests

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  4. Simulation of Fission Product Liftoff Behavior During Depressurization Transients

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Pilch, M.M.; Allen, M.D.; Bergeron, K.D.; Tadios, E.L.; Stamps, D.W.; Spencer, B.W.; Quick, K.S.; Knudson, D.L.

    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

  6. Mitigation of corrosion product ingress into SG's

    International Nuclear Information System (INIS)

    Han, S.H.

    1988-01-01

    Design and operation experiences to mitigate corrosion product ingress into SGs in Korea nuclear power plants are briefly reviewed. Maintaining the feedwater pH above 9.6 with morpholine seems to contribute significantly to reduction of iron transport to SGs. Measured iron transport rates were 4.8 g/hr/100 MWe at pH 9.8 and 2.8 g/hr/100 MWe at 9.3, respectively. Removal of corrosion products through SG blowdown is very limited. Its removal efficiency at the higher pH plant was in the neighborhood of 10 %. In one of the Korea Nuclear Units, a large amount of sludge piles were found in the middle of tube bundles especially on the cold leg side. Damaged tubes were identified by the multi-frequency eddy current tests and plugged later during the refueling period. Intermittent blowdown-rate increase was tried to enhance ionic impurity removal through SG blowdown. Even though it was not effective against Na, removal other impurity was improved, resulting in prolonged condensate polisher operation periods by 1 - 2 days. Two-bed polisher design, a cation bed followed by a mixed bed, was chosen for future PWR plants to enhance corrosion product filtering capability of the polishers. Condensate pump discharge polishing and divided hot well polishing methods are currently in consideration. (Nogami, K.)

  7. COMEDIE BD1 experiment: Fission product behaviour during depressurization transients

    International Nuclear Information System (INIS)

    Gillet, R.; Brenet, D.; Hanson, D.L.; Kimball, O.F.

    1996-01-01

    An experimental program in the CEA COMEDIE loop has been carried out to obtain integral test data to validate the methods and transport models used to predict fission product release from the core and plate-out in the primary coolant circuit of the Modular High Temperature Gas Cooled Reactor (MHTGR) during normal operation and liftoff, and during rapid depressurization transients. The loop consists of an in-pile section with the fuel element, deposition section (heat exchanger), filters for collecting condensible Fission Productions (FP) during depressurization tests and an out-of-pile section devoted to chemical composition control of the gas and on-line analysis of gaseous FP. After steady state irradiation, the loop was subjected to a series of in-situ blowdowns at shear ratios (ratio of the wall shear stress during blowdown to that during steady state operation) ranging from 0.7 to 5.6. The results regarding the FP profiles on the plate-out section, before and after blowdowns are given. It appears that: the plate-out profiles depend on the FP chemistry; the depressurization phases have led to significant desorption of I 131, but on the contrary, they have almost no effect for the other FP such as Ag 110m, Cs 134, Cs 137 and Te 132. (author). 1 ref., 15 figs

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

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

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

  9. Threshold for sweepout from pedestal region of Mark III containment

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Spencer, B.W.

    1984-01-01

    The assessment of the consequences of highly unlikely severe accident sequences in boiling water reactors includes those sequences in which molten corium is postulated to meltthrough the reactor pressure vessel (RPV) lower head and enter the pedestal region beneath the vessel. If localized melt-through of the reactor vessel occurs at elevated primary system pressure, the ejection of molten corium from the vessel will be followed by a blowdown of steam and hydrogen. The gases flowing from the breached vessel constitute a source of driving forces capable of dispersing corium from the pedestal into other parts of the containment. The extent of the gas blowdown-driven sweepout process depends upon a number of factors including the primary system pressure at melt through, breach flow area, overall blowdown timescale, and the specific pedestal/containment geometry. A model is presented to predict whether or not the conditions of gas flow from the failed RPV are sufficient to cause sweepout of corium and/or water from the pedestal. The model is shown to predict the onset of sweepout in scale model, simulant material experiments and is applied to the investigation of sweepout in the full-size reactor system

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

    International Nuclear Information System (INIS)

    Davis, C.M.; Day, R.S.; Smith, D.L.

    1996-07-01

    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

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

    International Nuclear Information System (INIS)

    Keen, R.; Duncan, G.M.

    1996-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. PPOOLEX experiments on dynamic loading with pressure feedback

    International Nuclear Information System (INIS)

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

    2011-01-01

    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

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-30

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

  16. Results of the General Atomic deposition loop program

    International Nuclear Information System (INIS)

    Hanson, D.L.

    1976-01-01

    The transport behavior of fission products in flowing helium streams has been studied to determine their deposition and re-entrainment characteristics. Such information is required for the design and safety analysis of high-temperature gas-cooled reactors (HTGRs). A small high-pressure, high-temperature loop was constructed for deposition studies at near-HTGR conditions. Five loop experiments were performed to determine the plateout distribution of iodine, strontium, and cesium. In general, the plateout activity showed an exponential decrease with distance from the source with enhanced plateout at flow disturber locations (contractions, bends, etc.) and especially in a chill section where the surface was cooled. Blowdown tests were performed on selected loop specimens to determine the amount of re-entrainment caused by abnormally high wall shear stresses. The liftoff fraction (fractional amount removed) was shown to vary approximately linearly with the shear ratio (defined as the ratio of the steady state wall shear stress under blowdown conditions to that under normal operating conditions). Blowdown results are also reported for pipe sections taken from the GAIL-IV in-pile loop. Attempts were made to correlate these plateout data with the PAD code (Plateout Activity Distribution) which was developed for prediction of plateout distribution in an HTGR primary circuit. Because of inadequate modeling of the effects of the chill section, the agreement was generally poor. Consequently, to test further the PAD code, a review of the available plateout literature was made. Plateout distributions in the Peach Bottom and Dragon HTGRs and the Battelle Memorial Institute out-of-pile loop were successfully modeled

  17. The potential for the recovery and reuse of cooling water in Taiwan

    Energy Technology Data Exchange (ETDEWEB)

    You, Shu-Hai; Tseng, Dyi-Hwa; Guo, Gia-Luen; Yang, Jyh-Jian [Graduate Institute of Environmental Engineering, National Central University, Chungli (Taiwan, Province of China)

    1999-04-01

    The cooling water is the major part of industrial water use in Taiwan, either from the view of demand priority or supply volume. In order to save water, the loading of supply system can be reduced if the cooling water can be recovered and reused. For this reason, exploration of the recent operation status of the cooling water system has become essential in Taiwan. This study was initially focused on the current applications and reuse trends of cooling water in oil refineries, chemical industry, steel mills, food industry, electronics works, textile plants and power stations. According to the statistical analysis, the portable water and groundwater are the primary sources of makeup water for cooling systems. The multiple-chemicals method and makeup treatment are increasingly accepted for the reclamation of cooling water. On the other hand, sidestream treatment and blowdown reuse are not popular in Taiwan. The recovery rate of blowdown is only 26.8%. The fact of higher cost is the major reason to depress the willingness of recovery. Some representative plants had been selected for case study. However, most cooling water systems are only operated by operator`s experience according to field investigation. In each case, the water quality indexes were used to evaluate the operational condition of cooling water systems. There was no case plant found to be operated at appropriate cycles of concentration. This paper also presented the bottlenecks of conservation technologies of cooling water in Taiwan. These bottlenecks include increasing the cycles of concentration, the reuse of wastewater, and the blowdown treatment for reuse. This paper also demonstrates that the recovery and reuse of cooling water has great potential and is feasible for the available technologies in present Taiwan, but the industries are still unwilling to upgrade because of initial cost. Finally, some approaches associated with technology, economics, environment and policy are proposed to be a

  18. Strain measurements at the HDR-pipe-system under LOCA-load: Effects on elbows and displaced weldings

    International Nuclear Information System (INIS)

    Hunger, H.

    1985-01-01

    This paper characterizes some effects which have been detected during strain gauge measurements on a test piping with feed water check valve oscillating under blowdown-load. The ovalization of a pipe elbow subjected to in-plane-bending affects the connected straight pipe; this is shown by means of circumferential stresses. Very high LOCA-load produces plastic strain and changes the pipe dynamics. Artificial displaced welds increase the local strain but no defects have occurred. One example compares stresses from measurement and post-calculation. Moreover there are given some remarks on the optimization of the comparison of measurement and calculation. (orig.)

  19. Steam line rupture experiments with the PPOOLEX test facility

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2008-07-01

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

  20. Review of the GOTHIC code and trial application

    Energy Technology Data Exchange (ETDEWEB)

    Lacroix, M; Galanis, N; Millette, J [Marcel Lacroix Enr., Sherbrooke, PQ (Canada)

    1996-01-01

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

  1. Steam line rupture experiments with the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

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

  2. Technical manual for COMET

    International Nuclear Information System (INIS)

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

    1996-07-01

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

  3. The effects of high-Ca hardness water treatment for secondary cooling water in HANARO

    International Nuclear Information System (INIS)

    Kang, T. J.; Park, Y. C.; Hwang, S. R.; Lim, I. C.; Choi, H. Y.

    2003-01-01

    Water-quality control of the second cooling system in HANARO has been altered from low Ca-hardness treatment to high Ca-hardness treatment since March, 2001. High Ca-hardness water treatment in HANARO is to maintain the calcium hardness around 12 by minimizing the blowdown of secondary cooling water. This paper describes the effect of cost reduction after change of water-quility treatment method. The result shows that the cost of the water could be reduced by 25% using the pond water in KAERI. The amount and cost for the chemical agent could be reduced by 40% and 10% respectively

  4. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1976-01-01

    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

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

    International Nuclear Information System (INIS)

    Gill, C.R.; Coddington, P.

    1988-05-01

    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)

  6. Passive containment system

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1977-01-01

    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

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

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

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

  8. Two-phase flow dynamics in ECC

    International Nuclear Information System (INIS)

    Albraaten, P.J.

    1981-07-01

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

  9. Modeling of two-phase flow with thermal and mechanical non-equilibrium

    International Nuclear Information System (INIS)

    Houdayer, G.; Pinet, B.; Le Coq, G.; Reocreux, M.; Rousseau, J.C.

    1977-01-01

    To improve two-phase flow modeling by taking into account thermal and mechanical non-equilibrium a joint effort on analytical experiment and physical modeling has been undertaken. A model describing thermal non-equilibrium effects is first presented. A correlation of mass transfer has been developed using steam water critical flow tests. This model has been used to predict in a satisfactory manner blowdown tests. It has been incorporated in CLYSTERE system code. To take into account mechanical non-equilibrium, a six equations model is written. To get information on the momentum transfers special nitrogen-water tests have been undertaken. The first results of these studies are presented

  10. Numerical solution of one dimensional two-phase drift flux equations with a blend of partially and fully implicit methods

    International Nuclear Information System (INIS)

    Mahaffy, J.H.; Liles, D.R.

    1977-01-01

    A numerical method for treating two-phase flow in pipes is presented which incorporates the use of a partially implicit scheme in regions of relatively low flow velocity and a fully implicit treatment in regions of high velocity. This method takes advantage of the lower cost per iteration of the partially implicit scheme, without being limited by its conditional stability. Applications of this approach to water reactor blowdown calculations produce reductions in computer time by factors of 2 to 4 without a significant loss of accuracy

  11. A computational method to predict fluid-structure interaction of pressure relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. K.; Lee, D. H.; Park, S. K.; Hong, S. R. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    An effective CFD (Computational fluid dynamics) method to predict important performance parameters, such as blowdown and chattering, for pressure relief valves in NPPs is provided in the present study. To calculate the valve motion, 6DOF (six degree of freedom) model is used. A chimera overset grid method is utilized to this study for the elimination of grid remeshing problem, when the disk moves. Further, CFD-Fastran which is developed by CFD-RC for compressible flow analysis is applied to an 1' safety valve. The prediction results ensure the applicability of the presented method in this study.

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

    International Nuclear Information System (INIS)

    Mosby, W.R.

    1978-01-01

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

  13. Benchmark of the HDR E11.2 containment hydrogen mixing experiment using the MAAP4 code

    International Nuclear Information System (INIS)

    Lee, Sung, Jin; Paik, Chan Y.; Henry, R.E.

    1997-01-01

    The MAAP4 code was benchmarked against the hydrogen mixing experiment in a full-size nuclear reactor containment. This particular experiment, designated as E11.2, simulated a small loss-of-coolant-accident steam blowdown into the containment followed by the release of a hydrogen-helium gas mixture. It also incorporated external spray cooling of the steel dome near the end of the transient. Specifically, the objective of this bench-mark was to demonstrate that MAAP4, using subnodal physics, can predict an observed gas stratification in the containment

  14. Analysis and design of flow limiter used in steam generator

    International Nuclear Information System (INIS)

    Liu Shixun; Gao Yongjun

    1995-10-01

    Flow limiter is an important safety component of PWR steam generator. It can limit the blowdown rate of steam generator inventory in case of the main steam pipeline breaks, so that the rate of the primary coolant temperature reduction can be slowed down in order to prevent fuel element from burn-out. The venturi type flow limiter is analysed, its flow characteristics are delineated, physical and mathematical models defined; the detail mathematical derivation provided. The research lays down a theoretic basis for flow limiter design. The governing equations and formulas given can be directly applied to computer analysis of the flow limiter. (3 refs., 3 figs.)

  15. Multinode analysis of small breaks for B and W's 205-fuel-assembly nuclear plants with internals vent valves

    International Nuclear Information System (INIS)

    Jones, R.C.; Dunn, B.M.; Parks, C.E.

    1976-03-01

    Multinode analyses were conducted for several small breaks in the reactor coolant system of B and W's 205-fuel-assembly nuclear plants with internals vent valves. The multinode blowdown code CRAFT was used to evaluate the hydrodynamics and transient water inventories of the reactor coolant system. The FOAM code was used to compute a swell level history for the core, and the THETA1-B code was used to perform transient fuel pin thermal calculations. Curves showing the parameters of interest are presented. The results are well within the Final Acceptance Criteria

  16. Multinode analysis of small breaks for B and W's 177-fuel-assembly nuclear plants with raised loop arrangement and internals vent valves

    International Nuclear Information System (INIS)

    Cartin, L.R.; Hill, J.M.; Parks, C.E.

    1976-03-01

    Multinode analyses were conducted for several small breaks in the reactor coolant system of B and W's 177-fuel-assembly nuclear plants with a raised loop arrangement and internals vent valves. The multinode blowdown code CRAFT was used to evaluate the hydrodynamics and transient water inventories of the reactor coolant system. The FOAM code was used to compute a swell level history for the core, and THETAL-B was used to perform transient fuel pin thermal calculations. Curves showing parameters of interest are presented. The results of these analyses are acceptable within the guidelines set forth in the Final Acceptance Criteria

  17. Multinode analysis of small breaks for B and W's 145-fuel-assembly nuclear plants with internals vent valves

    International Nuclear Information System (INIS)

    Parks, C.E.; Allen, R.J.; Cartin, L.R.

    1976-03-01

    Multinode analyses were conducted for several small breaks in the reactor coolant system of B and W's 145 fuel-assembly nuclear plants with internals vent valves. The multinode blowdown code CRAFT was used to evaluate the hydrodynamics and transient water inventories of the reactor coolant system. The FOAM code was used to compute a swell level history for the core, and the THETA1-B code was used to perform transient fuel pin thermal calculations. Curves showing the parameters of interest are presented. These results are well within the Final Acceptance Criteria

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

    International Nuclear Information System (INIS)

    Blanchat, T.K.; Allen, M.D.; Pilch, M.M.

    1994-01-01

    The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories (SNL) are used to perform scaled experiments for the Nuclear Regulatory Commission (NRC) that simulate High Pressure Melt Ejection (HPME) 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 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 atmospheres, and hydrogen generation and combustion, can be studied

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

    International Nuclear Information System (INIS)

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

    1991-08-01

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

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

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1976-04-01

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

  1. Non-intuitive fluid dynamics from reactor and containment technology

    International Nuclear Information System (INIS)

    Moody, F.J.

    1986-01-01

    One exciting aspect of fluid dynamics is that the subject has many surprises. The surprises can be good, but if not anticipated, they sometimes can be costly and embarrassing. Several non-intuitive fluid responses have emerged from studies in nuclear reactor and containment design. These responses include bubble behavior, blowdown, and waterhammer phenomena. Apologies are extended to those who are not surprised by the results. However, many will find the examples interesting; some have been amazed; a few have declared a personal crisis in their engineering perception

  2. Design of a hydrogen test facility

    International Nuclear Information System (INIS)

    Morgan, M.J.; Beam, J.E.; Sehmbey, M.S.; Pais, M.R.; Chow, L.C.; Hahn, O.J.

    1992-01-01

    The Air Force has sponsored a program at the University of Kentucky which will lead to a better understanding of the thermal and fluid instabilities during blowdown of supercritical fluids at cryogenic temperatures. An integral part of that program is the design and construction of that hydrogen test facility. This facility will be capable of providing supercritical hydrogen at 30 bars and 35 K at a maximum flow rate of 0.1 kg/s for 90 seconds. Also presented here is an extension of this facility to accommodate the use of supercritical helium

  3. Assessment of the RELAP4/MOD6 thermal-hydraulic transient code for PWR experimental applications. Addendum. Analyses completed and reported in FY 1979. Interim report

    International Nuclear Information System (INIS)

    1979-09-01

    The results of three subtasks that complete the assessment of the RELAP4/MOD6 computer code are reported. These subtasks constitute the remainder of a broadly scoped assessment matrix defined and described in detail in a previously published document. The specific subtasks provide comparisons of code calculations with experimental results from core blowdown and critical-flow separate-effects experiments and from an integral systems-effects loss-of-coolant experiment. The basic emphasis of the comparisons is in the presentation of the study results in error form suitable for statistical analysis

  4. Function analysis of steam isolation valves

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  5. Function analysis of steam isolation valves

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1975-06-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    1972-01-01

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

  9. Description of the THYDE-P code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    1978-07-01

    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.)

  10. 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.

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

    International Nuclear Information System (INIS)

    Chang, S.C.

    1979-01-01

    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

  12. Measurement of supercritical CO2 critical flow: Effects of L/D and surface roughness

    International Nuclear Information System (INIS)

    Mignot, Guillaume P.; Anderson, Mark H.; Corradini, Michael L.

    2009-01-01

    The use of supercritical fluids (SCF) has been proposed for advanced power systems including advanced sodium reactors, since these fluids can provide higher thermal efficiency and reduced system component size. Data characterizing the behavior of SCF during a blowdown or rapid depressurization are essential to validate certain aspects of safety analyses. This paper describes the results of an experiment to measure the critical mass flux for numerous stagnation thermodynamic conditions, geometry and outlet tube roughness. It was found that a 1D homogeneous equilibrium model (HEM) was capable of relatively good (less than 10% error) prediction of the test data.

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

    International Nuclear Information System (INIS)

    Ludewig, H.; Powell, J.R.; Todosow, M.; Maise, G.; Barletta, R.; Schweitzer, D.G.

    1996-01-01

    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)

  14. Thermal-hydraulic considerations for particle bed reactors

    Science.gov (United States)

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

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Larsson, H [JRC, Ispra (Italy)

    1977-07-01

    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.

  16. Review of the GOTHIC code and trial application

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. 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.

  18. 1993 RCRA Part B permit renewal application, Savannah River Site: Volume 10, Consolidated Incineration Facility, Section C, Revision 1

    International Nuclear Information System (INIS)

    Molen, G.

    1993-08-01

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

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

    International Nuclear Information System (INIS)

    Kim, B. H.; Choi, J. H.; Suk, S. D.; Kim, J. M.; Choi, B. H.; Kim, B. H.; Hahn, D. H.

    2007-01-01

    The utilization of modular sodium-to-supercritical CO 2 heat exchangers may yield significant improvements for an overall plant energy utilization. The consequences of a failure of the sodium CO 2 heat exchanger boundary, however, would involve the blowdown and intermixing of high-pressure CO 2 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 CO 2 would likely produce sodium oxides, sodium carbonate, carbon and carbon monoxide. Information on the kinetics of the sodium-CO 2 reaction is virtually non-existent

  20. LOCA verification and data bank

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. 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.

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

    International Nuclear Information System (INIS)

    Barre, F.; Dor, I.; Sun, C.

    1995-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    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.

  4. 40 CFR 471.64 - Pretreatment standards for existing sources (PSES).

    Science.gov (United States)

    2010-07-01

    ... 1.78 Ammonia 389 171 Fluoride 174 77.1 (o) Wet air pollution control scrubber blowdown. Subpart F... contact cooling water Cyanide 0.142 0.059 Lead 0.205 0.098 Zinc 0.713 0.298 Ammonia 65.1 28.6 Fluoride 29... 0.021 0.009 Lead 0.030 0.015 Zinc 0.105 0.044 Ammonia 9.59 4.22 Fluoride 4.28 1.90 (f) Extrusion...

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

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Namatame, Ken; Takeshita, Isao; Shiba, Masayoshi

    1982-12-01

    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)

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

    International Nuclear Information System (INIS)

    Engel, R.

    1980-02-01

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

  7. R.B. pressure and temperature transient following main steam line break

    International Nuclear Information System (INIS)

    Das, M.; Bhawal, R.N.; Prakash, P.

    1989-01-01

    The R.B. containment plays an important role in mitigating the consequences of any accident core. The analysis of Main Steam Line Break (MSLB), though not of relevance from activity release considerations, is essentially from structural integrity point of view. In this paper the outline of the likely scenario is drawn and the approach for thermal hydraulic simulation of the system for carrying out transient blowdown analysis is discussed. The results of the containment pressure and temperature transient analysis are also presented. (author). 4 refs., 7 figs

  8. Loads on EPR containment after RPV failure at high pressure; Belastungen des EPR-Containments in Falle eines RDB-Versagens bei hohem Druck

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, G.

    1995-08-01

    As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

  9. Loads on EPR containment after RPV failure at high pressure

    International Nuclear Information System (INIS)

    Jacobs, G.

    1995-01-01

    As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

  10. Analysis code for pressure in reactor containment vessel of ATR. CONPOL

    International Nuclear Information System (INIS)

    1997-08-01

    For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

  11. 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.

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1976-06-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. 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 user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc

  15. Fuel cell heat utilization system; Nenryo denchi netsuriyo sochi

    Energy Technology Data Exchange (ETDEWEB)

    Urata, T. [Tokyo (Japan); Omura, T. [Tokyo (Japan)

    1995-07-04

    In the conventional fuel cell heat utilization system, the waste heat is recovered to be utilized by either the waste heat recovery heat exchanger or the waste heat recovery steam. In the employment of the waste heat recovery heat exchanger system, however, the utility value is decreased when the temperature of the waste heat is lowered. Contrarily, in the employment of the waste heat recovery steam system, the supplementary water requirement is increased corresponding to the amount of waste heat recovery steam, resulting in the cost increase for water treatment. This invention solves the problem. In the invented fuel cell heat utilization system, a pressurized water from the steam separator is introduced into the second circuit to utilize directly the heat in the heat utilization system without employing the heat exchanger. If a blowdown valve is installed between the second circuit heat utilization system and the steam separator, the heat loss due to the blowdown can be reduced, since the low temperature water is blown down after being utilized in the heat utilization system. 4 figs.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-09-01

    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 542 0 F 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 66 0 F 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

  18. Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

    1984-01-01

    The results of reactor material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address exvessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debrids characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity. (orig.)

  19. The assessment of structural dynamics problems in nuclear reactor safety

    International Nuclear Information System (INIS)

    Liebe, R.

    1978-10-01

    The paper discusses important physical features of structural dynamics problems in reactor safety. First a general characterization is given of the following problems: Containment deformation due to pool-dynamics during BWR-blowdown; behavior of the core internals due to PWR-blowdown loads; dynamic response of a nuclear power plant during an earthquake; fuel element deformation due to local pressure pulses in an LMFBR core. Several criterias are formulated to classify typical problems so that a better choise can be made both of appropriate mathematical/numerical as well as experimental techniques. The degree of physical coupling between structural dynamics and fluid dynamics is discussed in more detail since it requires particular attention when selecting problem-oriented methods of solution. Some examples are given to illustrate the application and to compare advantages and disadvantages of several numerical methods. Then description is given of experimental techniques in structural dynamics and typical problem areas are identified. Finally some results are presented concerning the fuel element deformation problem in LMFBRs and from the general considerations some important conclusions are summarized. (orig.) 891 RW 892 AP [de

  20. Evaluation of the RELAP4/MOD6 thermal-hydraulic code

    International Nuclear Information System (INIS)

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA

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

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

    2005-03-01

    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

  2. Duke Power Company - McGuire Nuclear Station: steam-generator hideout return and cleanup

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    McGuire Nuclear Station steam generator hideout return and cleanup are discussed. Hideout return data are presented for Unit 1 shutdowns that occurred on November 23, 1984, and April 19, 1985, and a Unit 2 shutdown on January 25, 1985. The data are presented as the concentrations of various species as a function of time after power reduction and primary water temperature. The steam generator blowdown as a function of time after power reduction is also presented. The concentrations of sodium, potassium, calcium, magnesium, aluminum, iron, and copper cations, and chloride, fluoride, sulfate, phosphate and nitrite anions were monitored during the each shutdown. Silica was also measured in the two 1985 shutdowns. The return of sulfate, phosphate, calcium and magnesium showed retrograde solubility. Silica concentrations showed an increase as the temperature decreased to about 450 to 500 0 F and then they decreased as the temperature decreased. McGuire has a holf point at 300 at 350 0 F to clean up the steam generator secondary water. The return of sulfates should occur within 4 to 6 hours. The blowdown is maximized to reduce the secondary water impurity concentrations. Cleanup continues until the sulfate concentration is reduced to below 100 ppb. At that point cooldown is continued

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

    International Nuclear Information System (INIS)

    French, R.T.

    1975-08-01

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

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

    International Nuclear Information System (INIS)

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

    1975-12-01

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

  5. Modeling of corium dispersion in DCH accidents

    International Nuclear Information System (INIS)

    Wu, Q.

    1996-01-01

    A model that governs the dispersion process in the direct containment heating (DCH) reactor accident scenario is developed by a stepwise approach. In this model, the whole transient is subdivided into four phases with an isothermal assumption. These are the liquid and gas discharge, the liquid film flow in the cavity before gas blowdown, the liquid and gas flow in the cavity with droplet entrainment, and the liquid transport and re-entrainment in the subcompartment. In each step, the dominant driving mechanisms are identified to construct the governing equations. By combining all the steps together, the corium dispersion information is obtained in detail. The key parameters are predicted quantitatively. These include the fraction of liquid that flows out of the cavity before gas blowdown, the dispersion fraction and the mean droplet diameter in the cavity, the cavity pressure rise due to the liquid friction force, and the dispersion fractions in the containment via different paths. Compared with the data of the 1:10 scale experiments carried out at Purdue University, fairly good agreement is obtained. A stand-alone prediction of the corium dispersion under prototypic Zion reactor conditions is carried out by assuming an isothermal process without chemical reactions. (orig.)

  6. Coupled fluid-structure method for pressure suppression analysis

    International Nuclear Information System (INIS)

    McMaster, W.H.; Norris, D.M. Jr.; Goudreau, G.L.

    1979-01-01

    We have coupled an incompressible Eulerian hydrodynamic algorithm to a Lagrangian finite-element shell algorithm for the analysis of pressure suppression in boiling water reactors. The computer program calculates loads and structural response from air and steam blowdown and the oscillating condensation of steam bubbles in a water pool. 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 foundation of the program is the semi-implicit, two-dimensional SOLA algorithm. The shell structure algorithm uses conventional thin-shell theory with transverse shear. The finite-element spatial discretization employs piecewise-linear interpolation functions and one-point quadrature applied to conical frustra. We use the Newmark implicit time-integration method implemented as a one-step module. The algorithms are strongly coupled in the iteration loop using the iterated pressure in the fluid to drive the structure. The coupling algorithm requires normal velocity compatibility at the fluid-structure interface and incompressibility of the computational Eulerian zone overlaid by the structure. This is accomplished by iterating on the pressure field which is applied to the structure during each iteration until both conditions are satisfied

  7. PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History

    International Nuclear Information System (INIS)

    1987-01-01

    1 - Description of test facility: PHEBUS test facility operated at CEA Research Center Cadarache consists of a pressurized circuit involving pumps, heat exchangers and a blowdown tank - 25 nuclear fuel rod bundle, coupled to a separate driver core; - active length 0.8 m, cosine axial power profile; - pressurized and un-pressurized fuel rods; - controlled cooling conditions at the bundle inlet (blowdown, refill and reflood period); - de-pressurized test rig volume 0.22 m 3 . The following 'as measured' boundary conditions (B.C.) were offered to participants as options with decreasing challenge to their analytical approach: Boundary conditions B.C.0: - full thermal-hydraulic analysis of PHEBUS test rig (was not recommended). Boundary conditions B.C.1: - thermal power level of fuel bundle; - fluid inlet conditions to bundle section. Boundary conditions B.C.2: - local cladding temperatures of rods; - heat transfer coefficients. Boundary conditions B.C.3: - cladding temperatures of rods; - internal pressure of rods. 2 - Description of test: Post-test investigation into the response of a nuclear fuel bundle to a large break loss of coolant accident with respect to - local fuel temperatures, - cladding strain at the time of burst, - time to burst and under given thermal-hydraulic boundary conditions of PHEBUS-test 218

  8. 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.

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

    International Nuclear Information System (INIS)

    Langerman, M.A.

    1977-03-01

    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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

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

  12. Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

    International Nuclear Information System (INIS)

    Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

    1983-01-01

    The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

  13. Experimental investigation of a two-phase nozzle flow

    International Nuclear Information System (INIS)

    Kedziur, F.; John, H.; Loeffel, R.; Reimann, J.

    1980-07-01

    Stationary two-phase flow experiments with a convergent nozzle are performed. The experimental results are appropriate to validate advanced computer codes, which are applied to the blowdown-phase of a loss-of-coolant accident (LOCA). The steam-water experiments present 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 critical as well as subcritical 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 experiment. The layout of the nozzle and the applied measurement technique allow for a separate testing of blowdown-relevant, physical models and the determination of empirical model parameters, respectively. The measured quantities are essentially the mass flow rate, quality, axial pressure and temperature profiles as well as axial and radial density/void profiles obtained by a γ-ray absorption device. Moreover, impedance probes and a pitot probe are used. Observed phenomena like a flow contraction, radial pressure and void profiles as well as the appearance of two chocking locations are described, because their examination is rather instructive about the refinement of a program. The experimental facilities as well as the data of 36 characteristic experiments are documented. (orig.) [de

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

    International Nuclear Information System (INIS)

    Kahook, S.D.

    1992-08-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  17. Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

    1984-01-01

    Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

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

    International Nuclear Information System (INIS)

    1995-09-01

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

  19. Assessment of RELAP/MOD2 using large break loss-of-coolant experimental data

    International Nuclear Information System (INIS)

    Kao, L.; Liao, L.Y.; Liang, K.S.; Wang, S.F.; Chen, Y.B.

    1989-01-01

    In this paper assessment of RELAP5/MOD2 using LOFT L2-5 and Semiscale S-06-3 tests are performed to provide information of the code capability and its limitation in analyzing large break LOCA of a nuclear power plant. Experiments L2-5 and S-06-3 are conducted to simulate a hypothetical LOCA which results from a 200% double-ended offset shear break in the cold-leg of a typical pressurized water reactor by utilizing scaling facilities of the LOFT and Semiscale Mod-1 systems, respectively. The RELAP5/MOD2 calculations for both tests begin with break initiation and subsequent blowdown, continue through lower plenum refill, core reflood, and terminate with corewide quench. Major phenomena of both large break loss-of-coolant tests are well predicted by RELAP5/MOD2. The results indicate that the break flow and system pressure are reasonably calculated. The cladding temperature response during blowdown period, which is the major importance to a large break LOCA, calculated by RELAP5/MOD2 shows good agreement with the test data

  20. Impact of extreme load requirements and quality assurance on nuclear power plant costs

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1993-01-01

    Definitive costs, applicable to nuclear power plant concrete structures, as a function of National Regulatory Requirements, standardization, the effect of extreme load design associated with both design basis accidents and extreme external events and quality assurance are difficult to develop since such effects are interrelated and not only differ widely from country to country, project to project but also vary in time. Table 1 shows an estimate of the of the overall plant cost effects of external event extreme load design on nuclear power plant design for the U.S -and selected foreign countries for which experience with LWRs exist- Germany is the most expensive primarily due to a military aircraft crash resistance. However, the German requirement for 4 safeguards trains rather than 2 and the containment design requirement to consider one Steam Generator blowdown concurrent with a RCS blowdown. This presentation will concentrate on the direct current impact extreme load design and quality assurance have on concrete structures, systems and components for nuclear plants. This presentation is considered timely due to the increased interest in the c potential backfit of Eastern European nuclear power stations of the WWER 440 and WWER 1000 types which typically did not consider the extreme loads identified in Table 1 and accident loads in Table 3 and quality assurance in Table 5 in their original design. Concrete structures in particular are highlighted because they typically form the last barrier to radioactive release from the containment and other Safety Related Structures

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

    International Nuclear Information System (INIS)

    Malang, S.

    1975-11-01

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

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

    International Nuclear Information System (INIS)

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

    1993-07-01

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

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

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Frattini, P.; Robbins, P.; Miller, A.; Varrin, R.; Kreider, M.

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

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

  6. Experiment data report for Semiscale Mod-1 Test S-29-1 (integral test with asymmetrical break)

    International Nuclear Information System (INIS)

    Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

    1976-07-01

    Recorded test data are presented for Test S-29-1 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 (LOCA) in a pressurized-water reactor system. Test S-29-1 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,260 psia. An asymmetrical offset shear cold leg break was used to investigate the system response to a depressurization transient with a flow distribution different from that associated with a symmetrical cold leg break. System flow was set to achieve a core fluid temperature differential of 66 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 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 to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling (DNB) 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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

    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

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

    International Nuclear Information System (INIS)

    Rossi, C.E.

    1992-01-01

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

  9. Dryout delay in loss-of-coolant incidents in nuclear power plants

    International Nuclear Information System (INIS)

    Belda, W.

    1975-01-01

    The maximum credible accident (MCA) as a result of a fault in the system is assumed to be the rupture of a pipe in the primary circuit. During the outflow process following the rupture - called blowdown - it is possible that the internals of a reactor pressure vessel are exposed to extreme mechanical and thermal stresses. The fuel rods in the core, the Zircaloy cladding tubes of which can be heated up by lack of coolant to inadmissibly high temperatures, are particularly at risk. In case of the cladding tubes being damaged, radioactive substances are released. If they escape from the outer containment, this would lead to pressures on the immediate and more distant vicinity of the nuclear pover plant. In order to eliminate the factors of uncertainty when calculating the overall blowdown process in advance, it is necessary to have a relationship valid for the instationary circumstances to work out the burnout delay which is of decisive importance for the post-incident cooling phase of the reactor. The aim of this investigation, therefore, is to develop, with the aid of a suitable model, a method of calculating the burnout delay. (orig./TK) [de

  10. Post CHF heat transfer and quenching

    International Nuclear Information System (INIS)

    Nelson, R.A.; Condie, K.G.

    1980-01-01

    This paper describes quantitatively new mechanisms in the post-CHF regime which provide understanding and predictive capability for several current two-phase forced convective heat transfer problems. These mechanisms are important in predicting rod temperature turnaround and quenching during the reflood phase of either a hypothetical loss-of-coolant accident (LOCA) or the FLECHT and Semiscale experiments. The mechanisms are also important to the blowdown phase of a LOCA or the recent Loss-of-Fluid Test (LOFT) experiments L2-2 and L2-3, which were 200% cold leg break transients. These LOFT experiments experienced total core quenching in the early part of the blowdown phase at high (1000 psia) pressures. The mechanisms are also important to certain pressurized water reactor (PWR) operational transients where the reactor may operate in the post-CHF regime for short periods of time. Accurate prediction of the post-CHF heat transfer including core quench during these transients is of prime importance to limit maximum cladding temperatures and prevent cladding deformation

  11. A new water treatment scheme for thermal development : the SIBE process

    Energy Technology Data Exchange (ETDEWEB)

    Pedenaud, P.; Dang, F. [Total, Paris (France)

    2008-10-15

    The production of extra heavy oil or bitumen through thermal methods such as steam assisted gravity drainage (SAGD) involves the generation and injection into the reservoir of large quantities of steam which is recirculated with the produced bitumen. It is expected that maximizing the recycling of the produced water into steam will be mandatory, because of the need to minimize fresh water consumption and the possibility of increasingly stringent environmental regulations. The SAGD water treatment scheme is complex. It depends on the water characteristics, the steam generator type selected, and the decision to completely eliminate waste water disposal or use other waste handling and disposal methods. Other challenges such as the high silica content in the produced water, are encountered with SAGD water treatment. This paper presented an overview of the current water treatment process options for SAGD, as well as a new patented process called silica inhibition and blowdown evaporation (SIBE). The paper also presented an estimate of the economic benefit of the new SIBE process relative to conventional process schemes. Treatment objectives and water characteristics and the steps involved in conventional water treatment were first outlined. It was concluded that the silica and hardness removal scheme combined with the boiler blowdown evaporator were less economical because of higher investment cost due to the evaporation unit. 1 ref., 3 tabs., 4 figs.

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

    International Nuclear Information System (INIS)

    Julie M Jarvis; Allen T Vieira; James M Gilmer

    2005-01-01

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

  13. CFD Analysis for the Steady State Test of CS28-1 Simulating High Temperature Chemical Reactions in CANDU Fuel Channel

    International Nuclear Information System (INIS)

    Park, Ju Hwan; Kang, Hyung Seok; Rhee, Bo Wook

    2006-05-01

    The establishment of safety analysis system and technology for CANDU reactors has been performed at KAERI. As for one of these researches, single CANDU fuel bundle has been simulated by CATHENA for the post-blowdown event to consider the complicated geometry and heat transfer in the fuel channel. In the previous LBLOCA analysis methodology adopted for Wolsong 2, 3, 4 licensing, the fuel channel blowdown phase was analyzed by a CANDU system analysis code CATHENA and the post-blowdown phase of fuel channel was analyzed by CHAN-IIA code. The use of one computer code in consecutive analyses appeared to be desirable for consistency and simplicity in the safety analysis process. However, validation of the high temperature post-blowdown fuel channel model in the CATHENA before being used in the accident analysis is necessary. Experimental data for the 37-element fuel bundle that fueled CANDU-6 has not been performed. The benchmark problems for the 37-element fuel bundle using CFD code will be compared with the test results of the 28-element fuel bundle in the CS28-1 experiment. A full grid model of FES to the calandria tube simulating the test section was generated. The number of the generated mesh in the grid model was 4,324,340 cells. The boundary and heat source conditions, and properties data in the CFD analysis were given according to the test results and reference data. Thermal hydraulic phenomena in the fuel channel were simulated by a compressible flow, a highly turbulent flow, and a convection/conduction/radiation heat transfer. The natural convection flow of CO 2 due to a large temperature difference in the gap between the pressure and the calandria tubes was treated by Boussinesq's buoyancy model. The CFD results showed good agreement with the test results as a whole. The inner/middle/outer FES temperature distributions of the CFD results showed a small overestimated value of about 30 .deg. C at the entrance region, but good agreement at the outlet region. The

  14. 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

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

    International Nuclear Information System (INIS)

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

    2011-06-01

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

  16. 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

  17. ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA

    International Nuclear Information System (INIS)

    Araya, F.; Akimoto, M.

    1985-01-01

    1 - Nature of physical problem solved: ALARM-B2 which is an improved version of ALARM-B1 is a computer program to analyze thermo-hydraulic phenomena of BWR during a blowdown period under a large-break loss-of-coolant accident condition with special emphasis on the heat transfer phenomena in the core region. 2 - Method of solution: A so called volume-junction method is used to present fluid conservations. The primary system is divided into a number of special elements called 'control-volumes'. The system of partial differential equations describing fluid conservations for a stream-tube are integrated over a number of control volumes. The resulting set of simultaneous differential equations that is based on the assumptions of one-dimensional, homogeneous and thermal- equilibrium flow is linearized and solved for a small time increment by a simple explicit numerical technique. The one-dimensional heat conduction equations describing temperature profiles within solid material are written in finite difference forms which are linearized and solved by the Crank-Nicholson implicit method. In order to simulate the blowdown heat transfer phenomena, the code has correlation packages for heat transfer coefficient and critical heat flux. The heat generation in the core is given by a point reactor kinetics model with six groups of delayed neutrons and decay of eleven groups of fission products and actinides. The solution technique of the reactor kinetics is based on the Runge-Kutta method. ALARM-B2 has the models to simulate various components incorporated in BWRs such as jet pumps, recirculation pumps, steam separators, valves, and so on. The discharge and injection systems are modeled by leak and fill systems, respectively. 3 - Restrictions on the complexity of the problem: As this has been developed to simulate a blowdown thermo-hydraulic transient during a large break LOCA, users must pay attention when applying the code to any medium or small break LOCAs or to later phases

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

    International Nuclear Information System (INIS)

    Andropenkov, S.

    2000-01-01

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

  19. Full scale reactor safety experiments performed in the Marviken Power Station Sweden

    International Nuclear Information System (INIS)

    Thoren, H.G.; Ericson, L.

    1977-01-01

    Since 1972 experiments oriented towards increasing the understanding of reactor safety processes have been performed at the Marviken Power Station. This was originally built as a direct cycle BHWR but was never taken into nuclear operation. In addition to Sweden, the countries represented in these experiments are Denmark, the Federal Republic of Germany, Finland, Norway, the United States, the Netherlands, France and Japan. The first series of sixteen experiments included studies of the response of the PS-containment to simulated ruptures in the pipe systems that are connected to the pressure vessel. These tests were completed in 1973 and also included experimental studies of iodine transport, containment leakage, the behaviour of auxiliary components under accident conditions and pressure fluctuations in the wetwell water pool. One of the more essential findings of the tests was that the containment performance was in accordance with the pre-test calculations. A second series of eight blowdown tests was begun in February 1976. The main purpose of these tests is to provide additional information as to the characteristics of the pressure oscillations inside the containment and primarily in the wetwell water pool under different conditions. These oscillations were observed in the first series of blowdowns but only low frequencies could then be detected due to limitations in the measurement system. The measurement system was therefore substantially extended for this second series of experiments. A summary of the results from these two sets of blowdown tests are given in the paper. In 1976 preparations for a new test program were initiated. The objective of these tests is to improve the understanding of critical flow in the low quality and subcooled flow regions through short length, large diameter pipes. Extensive modifications of the test facility will be necessary in order to allow a discharge flow through openings which are up to 500 mm in diameter. Advanced plans

  20. Experimental investigation of void distribution in suppression pool over the duration of a loss of coolant accident using steam–water two-phase mixture

    International Nuclear Information System (INIS)

    Rassame, Somboon; Griffiths, Matthew; Yang, Jun; Ju, Peng; Sharma, Subash; Hibiki, Takashi; Ishii, Mamoru

    2015-01-01

    Highlights: • Experiments were conducted to study void fraction distribution in SP during blowdown. • 3 Experimental phases, namely, an initial and a quasi-steady phase, chugging were observed. • The maximum void penetration depth was experienced during the initial phase. • The quasi-steady phase provided less void penetration depth with oscillations. • The chugging phase was experienced at the end of experimental phase. - Abstract: Studies are underway to determine if a large amount gas discharged through the downcomer pipes in the pressure suppression chamber during the blowdown of Loss of Coolant Accident (LOCA) can potentially be entrained into the Emergency Core Cooling System (ECCS) suction piping of BWR. This may result in degraded ECCS pumps performance which could affect the ability to maintain or recover the water inventory level in the Reactor Pressure Vessel (RPV) during a LOCA. Therefore, it is very important to understand the void behavior in the pressure suppression chamber during the blowdown period of a LOCA. To address this issue, a set of experiments is conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. The geometry of the test apparatus is determined based on the basic geometrical scaling analysis from a prototypical BWR containment (MARK I) with a consideration of downcomer size, downcomer water submergence depth and Suppression Pool (SP) water level. Several instruments are installed in the test facility to measure the required experimental data such as the steam mass flow rate, void fraction, pressure and temperature. In the experiments, sequential flows of air, steam–air mixture and pure steam-each with the various flow rate conditions are injected from the Drywell (DW) through a downcomer pipe in the SP. Eight tests with two different downcomer sizes, various initial gas volumetric fluxes at the downcomer, and two different initial non-condensable gas

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  2. Analyse de l'emballement thermique d'un système chimique hybride non tempéré

    OpenAIRE

    Véchot , Luc; Bigot , Jean-Pierre; Minko , Wilfried ,; Kazmierczak , Marc; Vicot , Patricia

    2009-01-01

    National audience; Ce travail s'intéresse au “blow-down” (emballement thermique en présence d'un évent de sécurité) d'un système chimique non tempéré (30% CHP) soumis à un incendie. Il utilise une maquette à l'échelle 0,1 l. L'analyse des données post décomposition a montré que la vapeur présente est principalement un produit de la réaction. Toutes les expériences de blow-down ont présenté deux pics de pression, quel que soit le rapport A/V, ce qui est typique des systèmes non tempérés. Nous ...

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

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kodama, T.; Shiraishi, T.

    1985-01-01

    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.)

  4. Effects of thermocouple installation and location on fuel rod temperature measurements

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1983-01-01

    This paper describes the results of analyses of nuclear fuel rod cladding temperature data obtained during in-reactor experiments under steady state and transient (simulated loss-of-coolant accident) operating conditions. The objective of the analyses was to determine the effect of thermocouple attachment method and location on measured thermal response. The use of external thermocouples increased the time to critical heat flux (CHF), reduced the blowdown peak temperature, and enhanced rod quench. A comparison of laser welded and resistance welded external thermocouple responses showed that the laser welding technique reduced the indicated cladding steady state temperatures and provided shorter time-to-CHF. A comparison of internal welded and embedded thermocouples indicated that the welded technique gave generally unsatisfactory cladding temperature measurements. The embedded thermocouple gave good, consistent results, but was possibly more fragile than the welded thermocouples. Detailed descriptions of the thermocouple designs, attachment methods and locations, and test conditions are provided

  5. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the loss-of-fluid-test (LOFT) Reactor

    International Nuclear Information System (INIS)

    Welty, R.K.

    1980-01-01

    A welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods has been developed. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests. 2 refs

  6. A theoretical approach for energy saving in industrial steam boilers

    International Nuclear Information System (INIS)

    Sabry, T.I.; Mohamed, N.H.; Elghonimy, A.M.

    1993-01-01

    Optimization of the performance characteristics of such a steam boiler has been analyzed theoretically. Suitable thermodynamic relations have been utilized here to construct a computer model that would carry out the boiler performance characteristics at different operating parameters (e.g.; amount of excess air, fuel type, rate of blowdown preheating of combustion air and flow gases temperature). The results demonstrate that this computer model is to be used successfully in selecting the different operating parameters of the steam boiler at variant loads considering the best economical operation. Besides, this model can be used to investigate the sensitivity of the performance characteristics to the deviation of the boiler operating parameters from their optimum values. It was found also that changing the operating parameters beside the type of fuel in a boiler affects its performance characteristics. 3 figs

  7. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

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

    International Nuclear Information System (INIS)

    Sobajima, Makoto

    1985-08-01

    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)

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

    International Nuclear Information System (INIS)

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

    1976-10-01

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

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

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

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

  11. Condensate-polisher resin-leakage quantification and resin-transport studies

    International Nuclear Information System (INIS)

    Stauffer, C.C.; Doss, P.L.

    1983-04-01

    The objectives of this program were to: (1) determine the extent of resin leakage from current generation condensate polisher systems, both deep bed and powdered resin design, during cut-in, steady-state and flow transient operation, (2) analyze moisture separator drains and other secondary system samples for resin fragments and (3) document the level of organics in the secondary system. Resin leakage samples were obtained from nine-power stations that have either recirculating steam generators or once through steam generators. Secondary system samples were obtained from steam generator feedwater, recirculating steam generator blowdown and moisture separator drains. Analysis included ultraviolet light examination, SEM/EDX, resin quantification and infrared analysis. Data obtained from the various plants were compared and factors affecting resin leakage were summarized

  12. Draft report of a consultants' meeting on confinement improvement options for NPPs with WWER 440/230 reactors

    International Nuclear Information System (INIS)

    1996-01-01

    In the defence-in-depth safety approach for nuclear power systems, the containment is the final barrier against release of radioactive materials in the event of an accident. To function effectively, the containment needs good leaktightness and should be designed to withstand pressure loadings due to a wide range of postulated accidents (used as the basis of the design of the containment) in order to preserve the leaktightness and therefore prevent radioactive releases. It is typical that the containment be designed to withstand pressure loading due to the break of the largest size pipe in the reactor system involving blowdown from each end of the break. Different containment designs use different approaches to limit the resulting maximum pressure, including measures such as large volume, steam condensation in water pools, steam condensation in ice-bearing chambers, use of vacuum building, etc.

  13. Analysis of severe core damage accident progression for the heavy water reactor

    International Nuclear Information System (INIS)

    Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

    2010-01-01

    In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  15. SCTF Core-I test results

    International Nuclear Information System (INIS)

    Adachi, Hiromichi; Sudo, Yukio; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Hirano, Kemmei

    1982-07-01

    The Slab Core Test Facility (SCTF) of Japan Atomic Energy Research Institute (JAERI) was constructed to investigate two-dimensional thermohydrodynamics in the core and the communication 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). In the present report, effects of system pressure on reflooding phenomena shall be discussed based on the data of Tests S1-SH2, S1-01 and S1-02 which are the parameteris tests for system pressure effects belonging to the SCTF Core-I forced flooding test series. Major items discussed in this report are (1) hydrodynamic behavior in the system, (2) core thermal behavior, (3) core heat transfer and (4) two-dimensional hydrodynamic behavior in the pressure vessel including the core. (author)

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

    International Nuclear Information System (INIS)

    Bassett, B.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-03-01

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

  18. An experimental study of the effects of bodyside compression on forward swept sidewall compression inlets ingesting a turbulent boundary layer

    Science.gov (United States)

    Rodi, Patrick E.

    1993-01-01

    Forward swept sidewall compression inlets have been tested in the Mach 4 Blowdown Facility at the NASA Langley Research Center to study the effects of bodyside compression surfaces on inlet performance in the presence of an incoming turbulent boundary layer. The measurements include mass flow capture and mean surface pressure distributions obtained during simulated combustion pressure increases downstream of the inlet. The kerosene-lampblack surface tracer technique has been used to obtain patterns of the local wall shear stress direction. Inlet performance is evaluated using starting and unstarting characteristics, mass capture, mean surface pressure distributions and permissible back pressure limits. The results indicate that inlet performance can be improved with selected bodyside compression surfaces placed between the inlet sidewalls.

  19. Development of nuclear standard filter elements for PWR plant

    International Nuclear Information System (INIS)

    Weng Minghui; Wu Jidong; Gu Xiuzhang; Zhang Jinghua

    1988-11-01

    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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  2. Thermal and stress analysis of a fuel rod research project 277

    International Nuclear Information System (INIS)

    1975-04-01

    The purpose of this investigation was to perform an analytical evaluation of a postulated loss of coolant incident in a large pressurized water reactor. A coupled thermal and stress finite element analysis of a fuel rod subjected to a hypothetical blow-down transient was carried out. The effect of two gap conditions and two initial stress states on the response of the fuel rod was studied. Both one-dimensional and three-dimensional models were investigated. To study the heat transfer in the gap region one assumes a conductive mode of heat transfer in the gap characterized by an equivalent thermal conductivity, which is dependent on the current gap width. Accordingly, coupled analysis procedure and computational scheme were established. A mesh generation computer program was developed for the three-dimensional model

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  4. Analysis of standard problem six (Semiscale test S-02-6) data

    International Nuclear Information System (INIS)

    Cartmill, C.E.

    1977-08-01

    Test S-02-6 of the Semiscale Mod-1 blowdown heat transfer test series was conducted to supply data for the U.S. Nuclear Regulatory Commission Standard Problem Six. To determine the credibility of the data and thus establish the validity of Standard Problem Six, an analysis of the results of Test S-02-6 was performed and is presented. This analysis consisted of investigations of system hydraulic and core thermal data. The credibility of the system hydraulic data was investigated through comparisons of the data with data and calculations from related sources (Test S-02-4) and, when necessary, through assessment of physical events. The credibility of the core thermal data was based on a thorough analysis of physical events. The results of these investigations substantiate the validity of Test S-02-6 data

  5. Comparison of BEACON and COMPARE reactor cavity subcompartment analyses

    International Nuclear Information System (INIS)

    Burkett, M.W.; Idar, E.S.; Gido, R.G.; Lime, J.F.; Koestel, A.

    1984-04-01

    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. Experimental studies on transient water-steam impinging jet

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  7. Exxon Nuclear WREM-based NJP-BWR ECCS evaluation model and example application to the Oyster Creek Plant

    International Nuclear Information System (INIS)

    Krysinski, T.L.; Bjornard, T.A.; Steves, L.H.

    1975-01-01

    A proposed integrated ECCS model for non-jet pump boiling water reactors is presented, using the RELAP4-EM/BLOWDOWN and RELAP4-EM/SMALL BREAK portions of the Exxon Nuclear WREM-based Generic PWR Evaluation Model coupled with the ENC NJP-BWR Fuel Heatup Model. The results of the application of the proposed model to Oyster Creek are summarized. The results of the break size sensitivity study using the proposed model for the Oyster Creek Plant are presented. The application of the above results yielded the MAPLHGR curves. Included are a description of the proposed non-jet pump boiling water reaction evaluation model, justification of its conformance with TOCFR50, Appendix K, the adopted Oyster Creek plant model, and results of the analysis and sensitivity studies. (auth)

  8. Analysis of small leaks

    International Nuclear Information System (INIS)

    Frisch, W.; Hofmann, K.

    1979-01-01

    Problems associated with 'small leaks' are described and requirements are derived for experimental facilities and computer codes. Based on these requirements, a valuation of the existing experimental facilities and codes is presented. Facilities for integral tests in relatively large scale (ex. LOFT) are suitable for small leak test in principle, however minor changes (instrumentation, secondary side) are necessary for the evaluation of certain phenomena. The 'advanced blowdown codes' are capable of describing most of the phenomena occurring during small leak events, however a substantial amount of code development and verification is still needed. In addition, the use of transient codes in small leak analysis is demonstrated. There are some areas (neutronics feedback, influence of control system) in which the use of transient codes is possible and advantageous. (orig.) 891 HP/orig. 892 BRE [de

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

    International Nuclear Information System (INIS)

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

    1975-10-01

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

  10. Investigation of HEPA filters subjected to tornado pressure pulses

    International Nuclear Information System (INIS)

    Gregory, W.S.; Horak, H.L.; Smith, P.R.; Ricketts, C.

    1977-03-01

    An experimental program is described that will determine the response of 0.6-x 0.6-m (24-x 24-in.) high-efficiency particulate air (HEPA) filters to tornado-induced pressure transients. A blow-down system will be used to impose pressure differentials across the filters. Progress in construction of this system is reported with a description of the component parts and their functions. The test facility is essentially complete with the exception of an air dryer system that has not yet been delivered. Initial structural testing will begin in March 1977. A description is given of the instrumentation needed to measure air pressure, velocity, turbulence, humidity and particulate concentration. This instrumentation includes pressure transducers, humidity equipment, laser Doppler velocimeters (LDV), signal processors and a data acquisition system. Operational theory of the LDV and its proposed use as a particle counting device are described

  11. Analytical modelling of hydrogen transport in reactor containments

    International Nuclear Information System (INIS)

    Manno, V.P.

    1983-09-01

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

  12. 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)

    Modro, S.M.; Carboneau, M.L.

    1990-01-01

    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

  13. Full-Scale Mark II CRT Program data report, 1

    International Nuclear Information System (INIS)

    Namatame, Ken; Kukita, Yutaka; Yamamoto, Nobuo; Shiba, Masayoshi

    1979-12-01

    The Full-Scale Mark II CRT (Containment Response Test) Program was initiated in April 1976 to provide a full-scale data basis for the evaluation of the pressure suppression pool hydrodynamic loads associated with a hypothetical LOCA in a BWR Mark II Containment. The test facility, completed in March 1979, is 1/18 in volume of a typical 1100 MWe Mark II, and has a wetwell which is a full-scale replica of one 20 0 -sector of that of the reference Mark II. The present report documents experimental data from TEST 0002, a medium size (100 mm) water blowdown test, performed by Hitachi Ltd. for JAERI as the second of the four shakedown tests. Test data is provided for the vessel depressurization, the pressure and temperature responses in the test containment, and especially for the chugging phenomena associated with low flux steam condensation in the pool. (author)

  14. Simple interphase drag model for numerical two-fluid modeling of two-phase flow systems

    International Nuclear Information System (INIS)

    Chow, H.; Ransom, V.H.

    1984-01-01

    The interphase drag model that has been developed for RELAP5/MOD2 is based on a simple formulation having flow regime maps for both horizontal and vertical flows. The model is based on a conventional semi-empirical formulation that includes the product of drag coefficient, interfacial area, and relative dynamic pressure. The interphase drag model is implemented in the RELAP5/MOD2 light water reactor transient analysis code and has been used to simulate a variety of separate effects experiments to assess the model accuracy. The results from three of these simulations, the General Electric Company small vessel blowdown experiment, Dukler and Smith's counter-current flow experiment, and a Westinghouse Electric Company FLECHT-SEASET forced reflood experiment, are presented and discussed

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

    International Nuclear Information System (INIS)

    Nakao, Yoshinobu; Koda, Eiichi; Takahashi, Toru

    2009-01-01

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

  16. 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)

    D'Auria, F.; Vigni, P.

    1980-05-01

    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

  17. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

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

  18. 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.

  19. Trial evaluations in comparison with the 1983 safety goals

    International Nuclear Information System (INIS)

    Riggs, R.; Sege, G.

    1985-06-01

    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

  20. The COSIMA-experiments, a data base for validation of two-phase flow computer codes

    International Nuclear Information System (INIS)

    Class, G.; Meyder, R.; Stratmanns, E.

    1985-12-01

    The report presents an overview on the large data base generated with COSIMA. The data base is to be used to validate and develop computer codes for two-phase flow. In terms of fuel rod behavior it was found that during blowdown under realistic conditions only small strains are reached. For clad rupture extremely high rod internal pressure is necessary. Additionally important results were found in the behavior of a fuel rod simulator and on the effect of thermocouples attached on the cladding outer surface. Post-test calculations, performed with the codes RELAP and DRUFAN show a good agreement with the experiments. This however can be improved if the phase separation models in the codes would be updated. (orig./HP) [de

  1. 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.

  2. System containing a safety disk

    International Nuclear Information System (INIS)

    Schupp, W.

    1975-01-01

    The safety element is not overdimensioned at pressures between 2 and 150 atmospheric excess pressure. Therefore the flat bursting disc is mounted within a supporting and stopping holding and the rated breaking point is covered by a supporting body. Its outer diameter sufficiently overlaps the recesses on both sides of the rated breaking point. It absorbs the total load given by the operating pressure. Only a release mechanism with slide wedge, eccentric disc, magnet, and rocker arm releases the supporting body, e.g. if the blow-down pressure is reached, so that the operating pressure may work on the bursting disc. An insulated copper wire layed in the breaking region within the bursting disc in case of shearing off signalizes the instant of failing of the breaking point because of current interruption. (DG) [de

  3. CHEMCON User's Manual, Version 3.1

    International Nuclear Information System (INIS)

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

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1977-10-01

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

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

    International Nuclear Information System (INIS)

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

    1984-02-01

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

  7. Cooling tower water conditioning study. [using ozone

    Science.gov (United States)

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

    1979-01-01

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

  8. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  9. Accident loads for a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Techy, Z. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Lajtha, G. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Taubner, R. [Institute for Electric Power Research (VEIKI), Budapest (Hungary)

    1995-08-01

    Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system. (orig.).

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

    Science.gov (United States)

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

    1975-12-19

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

  11. Apparatus for monitoring two-phase flow

    Science.gov (United States)

    Sheppard, John D.; Tong, Long S.

    1977-03-01

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

  12. 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)

    Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

    1991-09-01

    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)

  13. Analysis of ATLAS 6-inch cold leg break simulation with MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Se Yun; Jun, Hwang Yong; Ha, Sang Jun [Korea Electric Power Company, Daejeon (Korea, Republic of)

    2011-05-15

    A Domestic Standard Problem (DSP) exercise using ATLAS facility has been organized by KAERI. As the second DSP exercise, the 6-inch cold leg bottom break was determined. This experiment is the counterpart test to the DVI line break to verify the safety performance of the DVI method over the traditional CLI method. Compared with the large break LOCA, the phases of the small break LOCA prior to core recovery occur over a long period. The blowdown, natural circulation, loop seal clearance, boil-off, and core recovery phase should be investigated minutely with relevant models of safety analysis codes in order to predict these thermal hydraulic phenomena correctly. To investigate the ECC bypass phenomena, a finer study on the thermalhydraulic behavior in upper annulus downcomer was carried out

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

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

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

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

    International Nuclear Information System (INIS)

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

    1978-04-01

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

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

    International Nuclear Information System (INIS)

    Rajan, Manju B.; Roy, Ankan; Ravi, K.V.

    2015-01-01

    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)

  17. Transition phase in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-07-01

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

  19. Reducing resin use in floor drain processing system

    International Nuclear Information System (INIS)

    Flint, W.; Hobart, S.A.; Miller, A.D.

    1995-01-01

    The Kewaunee Nuclear Power Plant utilizes two mixed bed demineralizers for processing floor drain wastes. These demineralizers were originally designed for stream generator blowdown treatment, but were not needed for that purpose. Effluent from the resin beds is monitored for radioactivity and released for discharge. Plant radwaste inleakage volumes and resin disposal volumes were low in comparison with industry averages, but decontamination factors through the treatment system were less than desirable. Release criteria for discharges always had been met, but plant personnel wished to decrease their already low discharges of radioactive species, reduce their resin disposal costs, and provide a margin of safety in the unlikely event that fuel damage would be experienced during an operating cycle. This paper describes the study initiated to address those issues, the findings of the study, and results of implementing some of the study recommendations

  20. 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)

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

    1982-05-01

    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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  2. 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.

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  4. Vessel coolant mass depletion during a 5% SBLOCA in the Semiscale Mod-2C facility

    International Nuclear Information System (INIS)

    Shaw, R.A.; Loomis, G.G.

    1985-01-01

    Experimental results are presented from two 5% small-break loss-of-coolant accident (SBLOCA) simulations in the Semiscale Mod-2C facility. In performing the simulated 5% SBLOCAs, boundary conditions scaled from a pressurized water reactor (PWR) were used. The experiment was run with initial conditions typical of a PWR (15.6 MPa pressure and 35 K core differential temperature). The Mod-2C facility represents the state-of-the-art in small facilities scaled from PWRs. Phenomena which occurred during the transient included: primary fluid saturation (change from subcooled to saturated blowdown), break uncovery (a centerline break was simulated), condensation-induced liquid hold-up in the steam generator primary tubes, pump suction liquid seal formation and core level depression with resulting core rod temperature excursion, pump suction liquid seal clearance, loop fluid mass redistribution, and gradual core rewet. The influence of core bypass flow is also discussed. 11 refs., 13 figs

  5. Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

    1978-01-01

    The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

  6. Fuel-rod response during the large-break LOCA Test LOC-6

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

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

    International Nuclear Information System (INIS)

    Sasaki, Shinobu; Araya, Fumimasa

    1980-03-01

    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)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  9. Structural test and analysis of a model of a BWR suppression chamber support in the plastic regime

    International Nuclear Information System (INIS)

    Blumer, U.R.; Klaeui, E.; Bosshard, E.P.

    1991-01-01

    A BWR Mark I suppression pool support has been analysed and tested in the laboratory. The aim was the demonstration of the acceptability of hypothetical dynamic loadings resulting from simultaneous steam blowdown through all safety relief valves. The analysis has shown that plastic deformation will locally occur, which is difficult to assess purely theoretical. Therefore tests in reduced scale were performed that show the amount and distribution of plastic flow in the supports. The paper describes the elastic analysis, the theory of the scaling laws for the reduced scale test, the test and its results. It also shows the thermographical method that has been used to determine the plastic material flow in the support structure. (author)

  10. 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.

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

    International Nuclear Information System (INIS)

    Perneczky, L.; Toth, I.; Szabados, L.; Ezsoel, Gy.

    1986-12-01

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

  12. INCAS TRISONIC WIND TUNNEL

    Directory of Open Access Journals (Sweden)

    Florin MUNTEANU

    2009-09-01

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

  13. The development and chemistry of high efficiency combined cycle plants

    International Nuclear Information System (INIS)

    Svoboda, Robert

    1999-01-01

    This paper presents a boiler concept based on the combination of a low-pressure drum-type boiler with high-pressure once-through boiler and the appropriate water/steam cycle. An all volatile treatment is used in the low-pressure boiler and oxygenated treatment for the once-through high pressure system. Impurity control is achieved by adapted system design and materials, high quality make-up, an appropriate cleanliness concept and clean-up procedures for a cold start. Cycle refreshing is realized by blowdown from the high-pressure water-separator. This concept utilizes simper and less equipment than traditional solutions, resulting in increased power plant reliability and less requirement on maintenance and on capital cost [it

  14. Kinetic---a system code for analyzing nuclear thermal propulsion rocket engine transients

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Science.gov (United States)

    Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans

    1993-01-01

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

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

    Science.gov (United States)

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

    1993-07-01

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

  17. Assessment of RELAP5/MOD3.1 using LOFT L2-3 experiment data

    International Nuclear Information System (INIS)

    Lee, Sang Yong; Ban, Chang Hwan; Chung, Bob Dong

    1994-06-01

    The capability of RELAP5/MOD3.1 to predict overall LOCA thermal hydraulic phenomena was assessed utilizing the data of LOFT L2-3 experiment. Loop behaviors such as mass flow rate, water density, momentum flux, and the heating-up and rewetting of the fuel rod cladding during blowdown were well calculated. Reflood heat-up of the fuel rod cladding at the high power region of the core was reasonably predicted. But in the upper part of the core, cladding heat-up was calculated incorrectly since present code has no capability to calculate the top-down quenching which of highly multi-dimensional behavior. (Author) 10 refs., 46 figs., 2 tabs

  18. Sensor for measurement of fuel rod gas pressure during loss-of-fluid-tests

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-05-01

    Qualification tests have been conducted of a measurement system for determining the pressure of certain fuel rods in the loss-of-fluid-test (LOFT) reactor. Because of physical size (0.35-in. OD by 5.5-in length) and operational characteristics, an eddy current device was selected as the most promising measurement transducer for the application. The sensor must operate at pressure up to 17.2 MPa (2500 psig) and at temperatures up to 800 0 F. During the reactor transient caused by loss of coolant flow, sensor temperature and applied pressure will vary rapidly and significantly. Consequently, qualification tests included subjection of the sensor to rapid depressurization, temperature transients, and blowdowns in an autoclave, as well as to calibrations and various slow temperature cycles

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

    International Nuclear Information System (INIS)

    Costa, J.R.

    1978-12-01

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

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

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  1. Steamgenerators corrosion monitoring and chemical cleanings

    International Nuclear Information System (INIS)

    Otchenashev, G.

    2001-01-01

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

  2. TRAC-PF1 choked-flow model

    International Nuclear Information System (INIS)

    Sahota, M.S.; Lime, J.F.

    1983-01-01

    The two-phase, two-component choked-flow model implemented in the latest version of the Transient Reactor analysis Code (TRAC-PF1) was developed from first principles using the characteristic analysis approach. The subcooled choked-flow model in TRAC-PF1 is a modified form of the Burnell model. This paper discusses these choked-flow models and their implementation in TRAC-PF1. comparisons using the TRAC-PF1 choked-flow models are made with the Burnell model for subcooled flow and with the homogeneous-equilibrium model (HEM) for two-phae flow. These comparisons agree well under homogeneous conditions. Generally good agreements have been obtained between the TRAC-PF1 results from models using the choking criteria and those using a fine mesh (natural choking). Code-data comparisons between the separate-effects tests of the Marviken facility and the Edwards' blowdown experiment also are favorable. 10 figures

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. 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.

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

    Science.gov (United States)

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

    2018-03-01

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

  6. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

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

  7. ROSA-II test data report, 13

    International Nuclear Information System (INIS)

    1978-07-01

    Results of the ROSA-II test simulating a loss-of-coolant accident (LOCA) in a PWR are presented, including test conditions and interpretations of phenomena observed in test runs 502, 505, 506 and 507. Development tests were performed to find a more effective ECCS injection method than the existing one based on cold leg injection. A combined injection of hot water into upper plenum in early stage of blowdown and subsequent cold water into lower plenum is the most effective method for a cold leg break. A hot leg injection of a low pressure injection system is effective for direct core cooling and early reflooding. The generalization for actual reactors will require analyses with a reliable code. (auth.)

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

    International Nuclear Information System (INIS)

    Brittain, I.; Aksan, S.N.

    1990-08-01

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

  9. Forced convective post CHF heat transfer and quenching

    International Nuclear Information System (INIS)

    Nelson, R.A.

    1980-01-01

    This paper discusses mechanisms in the post-CHF region which provide understanding and qualitative prediction capability for several current forced convective heat transfer problems. In the area of nuclear reactor safety, the mechanisms are important in the prediction of fuel rod quenches for the reflood phase, blowdown phase, and possibly some operational transients with dryout. Results using the mechanisms to investigate forced convective quenching are presented. Data reduction of quenching experiments is discussed, and the way in which the quenching transient may affect the results of different types of quenching experiments is investigated. This investigation provides an explanation of how minimum wall superheats greater than the homogeneous nucleation temperature result, as well as how these may appear to be either hydrodynamically or thermodynamically controlled. Finally, the results of a parametric study of the effects of the mechanisms upon the LOFT L2-3 hotpin calculation are presented

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

    International Nuclear Information System (INIS)

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

    1982-03-01

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

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

    International Nuclear Information System (INIS)

    Walker, B.W.

    1999-01-01

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

  12. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  14. Rod-bundle transient-film boiling of high-pressure water in the liquid-deficient regime

    International Nuclear Information System (INIS)

    Morris, D.G.; Mullins, C.B.; Yoder, G.L.

    1982-01-01

    Results are reported from a recent experiment investigating dispersed flow film boiling of high pressure water in upflow through a rod bundle. The data, obtained under mildly transient conditions, are used to assess correlations currently used to predict heat transfer in these circumstances. In light of the scarcity of similar data, the data should prove useful in the development and assessment of new heat transfer models. The experiment was conducted at the Oak Ridge National Laboratory in the Thermal-Hydraulic Test Facility, a highly instrumented, non-nuclear, pressurized-water loop containing 64, 3.66-m (12-ft) long rods (of which 60 are electrically heated). The rods are arranged in a square array typical of 17 x 17 fuel rod assemblies in late generation PWRs. Data were collected over typical reactor blowdown parameter ranges

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

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

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

  16. Water reclamation and reuse

    International Nuclear Information System (INIS)

    Hrudey, S.E.

    1982-01-01

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

  17. Apparatus for monitoring two-phase flow

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B. H.; Choi, J. H.; Suk, S. D.; Kim, J. M.; Choi, B. H.; Kim, B. H.; Hahn, D. H

    2007-01-15

    The utilization of modular sodium-to-supercritical CO{sub 2} heat exchangers may yield significant improvements for an overall plant energy utilization. The consequences of a failure of the sodium CO{sub 2} heat exchanger boundary, however, would involve the blowdown and intermixing of high-pressure CO{sub 2} 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 CO{sub 2} would likely produce sodium oxides, sodium carbonate, carbon and carbon monoxide. Information on the kinetics of the sodium-CO{sub 2} reaction is virtually non-existent.

  19. Experiment prediction for Loft Nonnuclear Experiment L1-4

    International Nuclear Information System (INIS)

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

    1977-04-01

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

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

    International Nuclear Information System (INIS)

    Edwards, A.R.; Borgartz, B.O.; Goodman, R.M.E.; O'Brien, T.P.; Rawlingson, M.

    1978-06-01

    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 250 0 C 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 10 0 C and 15 0 C while the reservoir was at saturation conditions with a steam dome present. (UK)

  1. Steam Line Break Analysis in CAREM-25 Reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

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

    International Nuclear Information System (INIS)

    1993-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

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

  5. Dispersant Application during SG Wet Layup at SK Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyukchul; Lee, Dooho; Sung, Kibang [KHNP Central Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The corrosion products in the feedwater are deposited onto the steam generators (SGs) despite the effort to control them within limit of impurity. This deposit is one of causes for occurrence of SCC (Stress Corrosion Cracking), water level fluctuation and further corrosion of SGs. To minimize corrosion and remove deposit, the nuclear power plants apply high pH to the secondary system and SG chemical cleaning, respectively. But these methods can be costly and carry risks of extended outages or incomplete cleaning. Another method is an on-line dispersant application. The role of dispersant is to make deposit suspended in the SGs. Then, the suspended deposit is discharged to the blowdown system. The iron removal is increased in the blowdown system during the dispersant application. Additional significant benefit in the form of reduced corrosion product transport may be obtained through applying dispersant in the SGs wet lay operational mode. This method helps to reduce the total SGs loading without affecting critical outage activities and with minimal additional effort on the part of the utilities. This study provides the results of the dispersant application trial during the SG wet layup at SK Unit 1. As the PAA concentrations were increased, the corrosion rates of Alloy 690 and SA 106 Gr.B were increased. The corrosion rate of Alloy 690 was 2 times less than that of SA 106 Gr.B at 100 ppm of PAA based on the electrochemical experimental. There were no significant feasibility problems with application of PAA during the SG wet layup. The reasonable estimation of the additional mass removed by the presence of PAA during SGs wet layup is 460 g. The iron removal depended on PAA concentration injected based on the comparative results of the SK Unit 1 and TMI-1. It is expected that injection of PAA into the SG result in a significant decrease in the amount of iron transported to the SGs during the startup.

  6. 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.

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

    International Nuclear Information System (INIS)

    Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

    2016-01-01

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

  8. The effect of nodalization and temperature of reactor upper region: Sensitivity analysis for APR-1400 LBLOCA

    International Nuclear Information System (INIS)

    Kang, Dong Gu

    2017-01-01

    Highlights: • The nodalization of APR-1400 was modified to reflect the characteristic of upper region temperature. • The effect of nodalization and temperature of reactor upper region on LBLOCA consequence was evaluated. • The modification of nodalization is an essential prerequisite in APR-1400 LBLOCA analysis. - Abstract: In best estimate (BE) calculation, the definition of system nodalization is important step influencing the prediction accuracy for specific thermal-hydraulic phenomena. The upper region of reactor is defined as the region of the upper guide structure (UGS) and upper dome. It has been assumed that the temperature of upper region is close to average temperature in most large break loss of coolant accident (LBLOCA) analysis cases. However, it was recently found that the temperature of upper region of APR-1400 reactor might be little lower than or similar to hot leg temperature through the review of detailed design data. In this study, the nodalization of APR-1400 was modified to reflect the characteristic of upper region temperature, and the effect of nodalization and temperature of reactor upper region on LBLOCA consequence was evaluated by sensitivity analysis including best estimate plus uncertainty (BEPU) calculation. In basecase calculation, in case of modified version, the peak cladding temperature (PCT) in blowdown phase became higher and the blowdown quenching (or cooling) was significantly deteriorated as compared to original case, and as a result, the cladding temperature in reflood phase became higher and the final quenching was also delayed. In addition, thermal-hydraulic parameters were compared and analyzed to investigate the effect of change of upper region on cladding temperature. In BEPU analysis, the 95 percentile PCT used in current regulatory practice was increased due to the modification of upper region nodalization, and it occurred in the reflood phase unlike original case.

  9. Development of Realistic Safety Analysis Technology for CANDU Reactors

    International Nuclear Information System (INIS)

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

    2010-04-01

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

  10. 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)

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

    International Nuclear Information System (INIS)

    Tanskanen, V.; Puustinen, M.

    2008-07-01

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

  12. 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.

  13. Development of loca calculation capability with relap5-3D in accordance with the evaluation model methodology

    International Nuclear Information System (INIS)

    Liang, T.K.S.; Huan-Jen, Hung; Chin-Jang, Chang; Lance, Wang

    2001-01-01

    In light water reactors, particularly the pressurized water reactor (PWR), the severity of a LOCA (loss of coolant accident) will limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the PCT (peak cladding temperature) evaluation during LOCA, it generally takes more resources to develop. Instead, implementation of evaluation models required by the Appendix K of 10 CFR 50 upon an advanced thermal-hydraulic platform can also enlarge significant margin between the highest calculated PCT and the safety limit of 2200 F. The compliance of the current RELAP5-3D code with Appendix K of 10 CFR50 has been evaluated, and it was found that there are ten areas where code assessment and/or further modifications were required to satisfy the requirements set forth in the Appendix K of 10 CFR 50. The associated models for LOCA consequent phenomenon analysis should follow the major concern of regulation and be expected to give more conservative results than those by the best-estimate methodology. They were required to predict the decay power level, the blowdown hydraulics, the blowdown heat transfer, the flooding rate, and the flooding heat transfer. All of the ten areas included in above classified simulations have been further evaluated and the RELAP5-3D has been successfully modified to fulfill the associated requirements. In addition, to verify and assess the development of the Appendix K version of RELAP5-3D, nine separate-effect experiments were adopted. Through the assessments against separate-effect experiments, the success of the code modification in accordance with the Appendix K of 10 CFR 50 was demonstrated. We will apply another six sets of integral-effect experiments in the next step to assure the integral conservatism of the Appendix K version of RELAP5-3D on LOCA licensing evaluation. (authors)

  14. Experiment data report for semiscale Mod-1, test S-02-7. Blowndown heat transfer test

    International Nuclear Information System (INIS)

    Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

    1975-11-01

    Recorded test data are presented for Test S-02-7 of the Semiscale Mod-1 blowdown heat transfer test series conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident (LOCA) in a water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Test S-02-7 was conducted from an initial cold leg fluid temperature of 543 0 F 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 transient with full design core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core with power set to provide a flat radial power profile. System flow was set to achieve the full design core temperature differential of 66 0 F. Blowdown to the pressure suppression system was accomplished without simulated emergency core cooling injection or pressure suppression system coolant spray. The uninterpreted data from Test S-02-7 are presented for future data analysis and test results reporting activities. 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. Also included as an appendix are selected data from a test identified as Test S-02-7C. This test was an initial attempt at Test S-02-7 in which an inadvertent power trip occurred at 2.3 seconds after rupture. Selected data comparisons of the results from Test S-02-7 and S-02-7C are presented to indicate the repeatability of system behavior

  15. Runout distance and dynamic pressure of pyroclastic density currents: Evidence from 18 May 1980 blast surge of Mount St. Helens

    Science.gov (United States)

    Gardner, J. E.; Andrews, B. J.

    2016-12-01

    Pyroclastic density currents (flows and surges) are one of the most deadly hazards associated with volcanic eruptions. Understanding what controls how far such currents will travel, and how their dynamic pressure evolves, could help mitigate their hazards. The distance a ground hugging, pyroclastic density current travels is partly limited by when it reverses buoyancy and lifts off into the atmosphere. The 1980 blast surge of Mount St. Helens offers an example of a current seen to lift off. Before lofting, it had traveled up to 20 km and leveled more than 600 km3 of thick forest (the blowdown zone). The outer edge of the devastated area - where burned trees that were left standing (the singe zone) - is where the surge is thought to have lifted off. We recently examined deposits in the outer parts of the blowdown and in the singe zone at 32 sites. The important finding is that the laterally moving surge travelled into the singe zone, and hence the change in tree damage does not mark the run out distance of the ground hugging surge. Eyewitness accounts and impacts on trees and vehicles reveal that the surge consisted of a fast, dilute "overcurrent" and a slower "undercurrent", where most of the mass (and heat) was retained. Reasonable estimates for flow density and velocity show that dynamic pressure of the surge (i.e., its ability to topple trees) peaked near the base of the overcurrent. We propose that when the overcurrent began to lift off, the height of peak dynamic pressure rose above the trees and stopped toppling them. The slower undercurrent continued forward, burning trees but it lacked the dynamic pressure needed to topple them. Grain-size variations argue that it slowed from 30 m/s when it entered the singe zone to 3 m/s at the far end. Buoyancy reversal and liftoff are thus not preserved in the deposits where the surge lofted upwards.

  16. Liftoff of the 18 May 1980 surge of Mount St. Helens (USA) and the deposits left behind

    Science.gov (United States)

    Gardner, James E.; Andrews, Benjamin J.; Dennen, Robert

    2017-01-01

    The distance that ground-hugging pyroclastic density currents travel is limited partly by when they reverse buoyancy and liftoff into the atmosphere. It is not clear, however, what deposits are left behind by lofting flows. One current that was seen to liftoff was the surge erupted from Mount St. Helens on the morning of 18 May 1980. Before lofting, it had leveled a large area of thick forest (the blowdown zone). The outer edge of the devastated area—where trees were scorched but left standing (the scorched zone)—is where the surge is thought to have lifted off. Deposits in the outer parts of the blowdown and in the scorched zone were examined at 32 sites. The important finding is that the laterally moving surge traveled through the scorched zone, and hence, the change in tree damage does not mark the runout distance of the surge. Buoyancy reversal and liftoff are thus not preserved in the deposits where the surge lofted upwards. We propose, based on interpretation of eyewitness accounts and the impacts of the surge on trees and vehicles, that the surge consisted of a faster, dilute "overcurrent" and a slower "undercurrent," where most of the mass (and heat) was retained. Reasonable estimates for flow density and velocity show that dynamic pressure of the surge (i.e., its ability to topple trees) peaked near the base of the overcurrent. We propose that where the overcurrent began to liftoff, the height of peak dynamic pressure rose above the trees and stopped toppling them. The slower undercurrent continued forward, however, scorching trees, but lacked the dynamic pressure needed to topple them. Grain-size variations argue that it slowed from ˜30 m s-1 when it entered the scorched zone to ˜3 m s-1 at the far end.

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

    International Nuclear Information System (INIS)

    Fuller, R.; Harrell, J.

    1996-01-01

    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

  18. Fast reactor primary cover gas system proposals for CDFR

    International Nuclear Information System (INIS)

    Harrison, L.M.T.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Robbe, M.F.

    2003-12-01

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