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Sample records for pwr blowdown heat

  1. 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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  4. 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

  5. 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

  6. 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

  7. 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.)

  8. 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)

  9. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

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

  10. 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.)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  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. 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)

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  6. 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

  7. 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

  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. 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

  10. 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

  11. A concept of PWR using plate and shell heat exchangers

    International Nuclear Information System (INIS)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de

    2015-01-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  12. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  13. 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)

  14. Improving MODPRESS heat loss calculations for PWR pressurizers

    International Nuclear Information System (INIS)

    Ramos, Natalia V.; Lira, Carlos A. Brayner O.; Castrillho, Lazara S.

    2009-01-01

    The improvement of heat loss calculations in MODPRESS transient code for PWR pressurizer analysis is the main focus of this investigation. Initially, a heat loss model was built based on heat transfer coefficient (HTC) correlations obtained in handbooks of thermal engineering. A hand calculation for Neptunus experimental test number U47 yielded a thermal power loss of 11.2 kW against 17.3 kW given by MODPRESS at the same conditions, while the experimental estimate is given as 17 kW. This comparison is valid only for steady state or before starting the transient experiment, because MODPRESS does not update HTC's when the transient phase begins. Furthermore, it must be noted that MODPRESS heat transfer coefficients are adjusted to reproduce the experimental value of the specific type of pressurizer. After inserting the new routine for HTC's into MODPRESS, the heat loss was calculated as 11.4 kW, a value very close to the first estimate but far below 17 kW found in the U47 experiment. In this paper, the heat loss model and results will be described. Further research is being developed to find a more general HTC that allows the analysis of the effects of heat losses on transient behavior of Neptunus and IRIS pressurizers. (author)

  15. Hydrodynamical tests with an original PWR heat removal pump

    International Nuclear Information System (INIS)

    Wietstock, P.

    1984-01-01

    GKSS-Forschungszentrum performes hydrodynamical tests with an original PWR heat removal pump to analyse the influences of fluid parameters on the capacity and cavitation behavior of the pump in order to get further improvements of the quantification of the reached safety-level. It can be concluded, that in case of the tested heat removal pump the additional loads during transition from cavitation free operation into fully cavitation for the investigated operation point with 980 m 3 /h will be smaller than the alteration of loads during passing through the total characteristic. The results from cavitation tests for other operation points indicate, that this very important consequence especially for accident operation will be valid for the total specified pump flow area. (orig.)

  16. 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

  17. 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

  18. 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

  19. Evaluation of uncertainties in the gamma-ray heating analysis of a PWR

    International Nuclear Information System (INIS)

    West, J.T.

    1977-01-01

    The limits of accuracy in a PWR gamma heating analysis, which used conventional one- and two-dimensional discrete ordinate transport methods, by utilizing three-dimensional Monte Carlo methods are determined

  20. 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

  1. Simplified analysis of passive residual heat removal systems for small size PWR's

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1992-02-01

    The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

  2. 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)

  3. Experimental research on passive residual heat remove system for advanced PWR

    International Nuclear Information System (INIS)

    Huang Yanping; Zhuo Wenbin; Yang Zumao; Xiao Zejun; Chen Bingde

    2003-01-01

    The experimental and qualified results of MISAP in the research of passive residual heat remove system of advanced PWR performed in the Bubble physics and natural circulation laboratory in Nuclear Power Institute of China in the past ten years is overviewed. Further researches for engineering research and design are also suggested

  4. Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach

    Energy Technology Data Exchange (ETDEWEB)

    Dourado, Eneida Regina G. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Cotta, Renato M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Mecanica; Jian, Su, E-mail: eneidadourado@gmail.com, E-mail: sujian@nuclear.ufrj.br, E-mail: cotta@mecanica.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)

  5. Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach

    International Nuclear Information System (INIS)

    Dourado, Eneida Regina G.; Cotta, Renato M.; Jian, Su

    2017-01-01

    This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)

  6. 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

  7. 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

  8. Some factors affecting radiative heat transport in PWR cores

    International Nuclear Information System (INIS)

    Hall, A.N.

    1989-04-01

    This report discusses radiative heat transport in Pressurized Water Reactor cores, using simple models to illustrate basic features of the transport process. Heat transport by conduction and convection is ignored in order to focus attention on the restrictions on radiative heat transport imposed by the geometry of the heat emitting and absorbing structures. The importance of the spacing of the emitting and absorbing structures is emphasised. Steady state temperature distributions are found for models of cores which are uniformly heated by fission product decay. In all of the models, a steady state temperature distribution can only be obtained if the central core temperature is in excess of the melting point of UO 2 . It has recently been reported that the MIMAS computer code, which takes into account radiative heat transport, has been used to model the heat-up of the Three Mile Island-2 reactor core, and the computations indicate that the core could not have reached the melting point of UO 2 at any time or any place. We discuss this result in the light of the calculations presented in this paper. It appears that the predicted stabilisation of the core temperatures at ∼ 2200 0 C may be a consequence of the artificially large spacing between the radial rings employed in the MIMAS code, rather than a result of physical significance. (author)

  9. 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

  10. 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

  11. 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

  12. Critical heat flux correlation analysis for PWR reactors with low mass flow

    International Nuclear Information System (INIS)

    Carajilescov, Pedro

    1996-01-01

    The major limit in the thermalhydraulic design of water cooled reactors consists in the occurrence of critical heat flux, which is verified by correlation of large range of validity. In the present work, the major design correlations were analyzed, through comparisons with experimental data, for utilization in PWR with low mass flux in the core. The results show that the EPRI correlation, with modifications, gives conservative results, from the safety point of view, with lower data spreading, being the most indicated for the reactor thermal design. (author)

  13. Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

    International Nuclear Information System (INIS)

    Su, Jian; Cotta, Renato M.

    2000-01-01

    In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

  14. Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

    International Nuclear Information System (INIS)

    Shimooke, Takanori; Yoshida, Kazuo

    1979-05-01

    The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

  15. PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

    International Nuclear Information System (INIS)

    1991-03-01

    An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

  16. 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)

  17. Experimental investigation of reflux condensation heat transfer in PWR steam generator tubes in the presence of noncondensible gases

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen; Wu, Tiejun [Purdue Univ., West Lafayette (United States); Nagae, Takashi [Institute of Nuclear Safety System, Tokyo (Japan)

    2003-07-01

    Under certain circumstances in a Pressurized Water Reactor (PWR), the coolant system may be in a partially drained state and reflux condensation in the steam generator U-tubes can be the major heat removal mechanism. Noncondensable gases may be present and would degrade the heat transfer rate. If heat removal rates are insufficient, this situation could lead to core boil-off, fuel rod heatup, and eventually core damage. The Institute of Nuclear Safety System, Inc. (INSS) and the Nuclear Heat Transfer Systems Laboratory at Purdue University have begun a cooperative research program to investigate the effectiveness of reflux condensation in PWR steam generator U-tubes in the presence of noncondensable gases. The final objectives are to provide local heat transfer data for development of methods to analyze reflux condensation in PWR steam generator U-tubes and to investigate the potential for flooding. Key features of the experimental data reported herein are that they are local data under laminar steam/gas mixture and condensate film flow and they are taken from a test section with dimensions similar to an actual steam generator tube. Steady state data were obtained under various steam and air inlet flow rates and pressures. The data show the significant degrading effect of noncondensable gas on heat transfer coefficients. From the data, correlations for the reflux condensation local heat transfer coefficient and the local Nusselt number under laminar conditions were derived. These experiments are providing essential and unique fundamental data for development of methods to analyze reflux condensation.

  18. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  19. 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

  20. 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

  1. 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

  2. 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

  3. Earthquake resistance of residual heat removed (RHR) pump for pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, K.; Honma, T.; Matsubayashi, H.; Inazuka, H.

    1980-01-01

    The present paper deals with the earthquake resistance of the residual heat removed (RHR) pump of single stage double volute type, which is one of the structurally simplest pumps used for pressurized water reactors (PWR). The results of the study can be summarized as follows: (1) Any trouble which can give effect on the functions of the pump at earthquake does not become a problem so long as each part of the pump is of aseismatically rigid structure. (2) Aseismatic tolerance test in the pump's operating condition has shown that the earthquake resistance of the pump at its location has a tolerance about five times the dynamic design acceleration. (3) The pump is provided with an impeller-casing wear ring at the pressure boundary between the suction side pressure and discharge side pressure. This wear ring acts as an underwater bearing when the pump is in operation, and improves the vibration characteristics, particularly damping ratio, of the pump shaft to a great extent to make the pump more aseismatic. (4) In the evaluation of the underwater bearing characteristics of the wear ring, the evaluation accuracy of the vibration characteristics of the pump shaft can be improved by taking into consideration the pressure loss in the wear ring part from the head of the single stage of the pump due to the rotation of the impeller. (author)

  4. NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test

    International Nuclear Information System (INIS)

    Richner, M.; Analytis, G.Th.; Aksan, S.N.

    1993-01-01

    1 - Description of test facility: NEPTUN is designed to perform PWR LOCA simulation experiments, which provide the full length emergency cooling heat transfer tests for LOFT. Therefore the NEPTUN heater bundle with 33 electrical heater elements and 4 guide tubes simulates a section of the LOFT nuclear core. The main test loop also contains measuring systems for the carry-over rate and for the steam expelled, and a back-pressure control system. A water loop brings the water to the initial reflooding conditions. In addition, auxiliary systems maintain normal operating conditions. 2 - Description of test: Test 5052 is one of a series of 40 reflood tests performed in NEPTUN. Before the start of the test, the flooding water in its circuit is brought to the following conditions: pressure = 4.1 bar; velocity = 2.5 cm/sec; subcooling temperature = 78 C; single rod power = 2.45 kW; maximal initial cladding temperature = 867 C. 3 - Status: CSNI1013/01, 21-Jul-1993 Arrived at NEADB

  5. 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

  6. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Kao, Lain-Su, E-mail: lskao@iner.gov.tw

    2013-10-15

    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to

  7. 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)

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. Assessment of a non-uniform heat flux correction model to predicting CHF in PWR rod bundles

    International Nuclear Information System (INIS)

    Dae-Hyun, Hwang; Sung-Quun, Zee

    2001-01-01

    The full text follows. The prediction of CHF (critical heat flux) has been, in most cases, based on the empirical correlation. For PWR fuel assemblies the local parameter correlation requires the local thermal-hydraulic conditions usually calculated by a subchannel analysis code. The cross-sectional averaged fluid conditions of the subchannel, however, are not sufficient for determining CHF, especially for the cases of non-uniform axial heat flux distributions. Many investigators have studied the effect of the upstream heat flux on the CHF. In terms of the upstream memory effect, two different approaches have been considered as the limiting cases. The 'local conditions' hypothesis assumes that there is a unique relationship between the CHF and the local thermal-hydraulic conditions, and consequently there is no memory effect. In the 'overall power' hypothesis, on the other hand, it is assumed that the total power which can be fed into the tube with nonuniform heating will be the same as that for a uniformly heated tube of the same heated length with the same inlet conditions. Thus the CHF is totally influenced by the upstream heat flux distribution. In view of some experimental investigations such as the DeBortoli's test, it revealed that the two approaches are inadequate in general. It means that the local critical heat flux may be affected to some extent by the heat flux distribution upstream of the CHF location. Some correction-factor models have been suggested to take into account the upstream memory effect. Typically, Tong devised a correction factor on the basis of the heat balance of the superheated liquid layer that is spread underneath a highly viscous bubbly layer along the heated surface. His physical model suggested that the fluid enthalpy obtained from an energy balance of the superheated liquid layer is a representative quantity for the onset of DNB (departure nucleate boiling). A theoretically based correction factor model has been proposed by the

  16. Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

  17. 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

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

    International Nuclear Information System (INIS)

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

    1993-03-01

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

  19. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method

    International Nuclear Information System (INIS)

    Silva, Alonso F.O.

    2007-01-01

    From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

  20. Numerical simulation of the heating and start-up of PWR nuclear power station

    International Nuclear Information System (INIS)

    Faraco-Medeiros, M.A.; Leite, C.A.T.; Ramalho, F.P.

    1992-01-01

    The start-up of a PWR nuclear power plant must be done within safety criteria and requires a simulation. The design of some equipment, cost and time can be optimized. A computer simulator, which allows control of all the equipment and variables into the operation, has been developed and is presented in this paper. The KWU procedure and an alternative for Angra II were simulated. The results are showed up. 09 refs, 03 figs. (B.C.A.)

  1. 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.)

  2. 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

  3. Probabilistic reliability analyses to detect weak points in secondary-side residual heat removal systems of KWU PWR plants

    International Nuclear Information System (INIS)

    Schilling, R.

    1984-01-01

    Requirements made by Federal German licensing authorities called for the analysis of the second-side residual heat removal systems of new PWR plants with regard to availability, possible weak points and the balanced nature of the overall system for different incident sequences. Following a description of the generic concept and the process and safety-related systems for steam generator feed and main steam discharge, the reliability of the latter is analyzed for the small break LOCA and emergency power mode incidents, weak points in the process systems are identified, remedial measures of a system-specific and test-strategic nature are presented and their contribution to improving system availability is quantified. A comparison with the results of the German Risk Study on Nuclear Power Plants (GRS) shows a distinct reduction in core meltdown frequency. (orig.)

  4. Reliability analyses to detect weak points in secondary-side residual heat removal systems of KWU PWR plants

    International Nuclear Information System (INIS)

    Schilling, R.

    1983-01-01

    Requirements made by Federal German licensing authorities called for the analysis of the secondary-side residual heat removal systems of new PWR plants with regard to availability, possible weak points and the balanced nature of the overall system for different incident sequences. Following a description of the generic concept and the process and safety-related systems for steam generator feed and main steam discharge, the reliability of the latter is analyzed for the small break LOCA and emergency power mode incidents, weak points in the process systems identified, remedial measures of a system-specific and test-strategic nature presented and their contribution to improving system availability quantified. A comparison with the results of the German Risk Study on Nuclear Power Plants (GRS) shows a distinct reduction in core meltdown frequency. (orig.)

  5. 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)

  6. 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

  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. Development, verification and validation of an FPGA-based core heat removal protection system for a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Yichun, E-mail: ycwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China); Shui, Xuanxuan, E-mail: 807001564@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Cai, Yuanfeng, E-mail: 1056303902@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Zhou, Junyi, E-mail: 1032133755@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Wu, Zhiqiang, E-mail: npic_wu@126.com [State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Zheng, Jianxiang, E-mail: zwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China)

    2016-05-15

    Highlights: • An example on life cycle development process and V&V on FPGA-based I&C is presented. • Software standards and guidelines are used in FPGA-based NPP I&C system logic V&V. • Diversified FPGA design and verification languages and tools are utilized. • An NPP operation principle simulator is used to simulate operation scenarios. - Abstract: To reach high confidence and ensure reliability of nuclear FPGA-based safety system, life cycle processes of discipline specification and implementation of design as well as regulations verification and validation (V&V) are needed. A specific example on how to conduct life cycle development process and V&V on FPGA-based core heat removal (CHR) protection system for CPR1000 pressure water reactor (PWR) is presented in this paper. Using the existing standards and guidelines for life cycle development and V&V, a simplified FPGA-based CHR protection system for PWR has been designed, implemented, verified and validated. Diversified verification and simulation languages and tools are used by the independent design team and the V&V team. In the system acceptance testing V&V phase, a CPR1000 NPP operation principle simulator (OPS) model is utilized to simulate normal and abnormal operation scenarios, and provide input data to the under-test FPGA-based CHR protection system and a verified C code CHR function module. The evaluation results are applied to validate the under-test FPGA-based CHR protection system. The OPS model operation outputs also provide reasonable references for the tests. Using an OPS model in the system acceptance testing V&V is cost-effective and high-efficient. A dedicated OPS, as a commercial-off-the-shelf (COTS) item, would contribute as an important tool in the V&V process of NPP I&C systems, including FPGA-based and microprocessor-based systems.

  9. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    International Nuclear Information System (INIS)

    Kim, In Young; Lee, Un Chul

    2011-01-01

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  10. Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

  11. 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

  12. A new correlation for convective heat transfer coefficient of water–alumina nanofluid in a square array subchannel under PWR condition

    International Nuclear Information System (INIS)

    Shamim, Jubair A.; Bhowmik, Palash K.; Xiangyi, Chen; Suh, Kune Y.

    2016-01-01

    Highlights: • Thermo-hydrodynamic properties of water–Al_2O_3 nanofluid at PWR condition is analyzed. • Details of CFD simulation and validation procedure is outlined. • Augmented heat transfer capacity of nanofluid is governed by larger pumping power. • A new correlation for nanofluid Nusselt number in subchannel geometry is proposed. - Abstract: The computational fluid dynamic (CFD) simulation is performed to determine on the thermo- and hydrodynamic performance of the water–alumina (Al_2O_3) nanofluid in a square array subchannel featuring pitch-to-diameter ratios of 1.25 and 1.35. Two fundamental aspects of thermal hydraulics, viz. heat transfer and pressure drop, are assessed under typical pressurized water reactor (PWR) conditions at various flow rates (3 × 10"5 ⩽ Re ⩽ 6 × 10"5) using pure water and differing concentrations of water–alumina nanofluid (0.5–3.0 vol.%) as coolant. Numerical results are compared against predictions made by conventional single-phase convective heat transfer and pressure loss correlations for fully developed turbulent flow. It is observed that addition of tiny nanoparticles in PWR coolant can give rise to the convective heat transfer coefficient at the expense of larger pressure drop. Nevertheless, a modified correlation as a function of nanoparticle volume fraction is proposed to estimate nanofluid Nusselt number more precisely in square array subchannel.

  13. A new correlation for convective heat transfer coefficient of water–alumina nanofluid in a square array subchannel under PWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, Jubair A. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Department of Mechanical Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Bhowmik, Palash K. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Department of Nuclear Engineering, Missouri University of Science and Technology, 1201 N. State St., Rolla, MO 65409 (United States); Xiangyi, Chen [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Suh, Kune Y., E-mail: kysuh@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of)

    2016-11-15

    Highlights: • Thermo-hydrodynamic properties of water–Al{sub 2}O{sub 3} nanofluid at PWR condition is analyzed. • Details of CFD simulation and validation procedure is outlined. • Augmented heat transfer capacity of nanofluid is governed by larger pumping power. • A new correlation for nanofluid Nusselt number in subchannel geometry is proposed. - Abstract: The computational fluid dynamic (CFD) simulation is performed to determine on the thermo- and hydrodynamic performance of the water–alumina (Al{sub 2}O{sub 3}) nanofluid in a square array subchannel featuring pitch-to-diameter ratios of 1.25 and 1.35. Two fundamental aspects of thermal hydraulics, viz. heat transfer and pressure drop, are assessed under typical pressurized water reactor (PWR) conditions at various flow rates (3 × 10{sup 5} ⩽ Re ⩽ 6 × 10{sup 5}) using pure water and differing concentrations of water–alumina nanofluid (0.5–3.0 vol.%) as coolant. Numerical results are compared against predictions made by conventional single-phase convective heat transfer and pressure loss correlations for fully developed turbulent flow. It is observed that addition of tiny nanoparticles in PWR coolant can give rise to the convective heat transfer coefficient at the expense of larger pressure drop. Nevertheless, a modified correlation as a function of nanoparticle volume fraction is proposed to estimate nanofluid Nusselt number more precisely in square array subchannel.

  14. Study of transient heat transfer in a fuel rod 3D, in a situation of unplanned shutdown of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Martins, Rodolfo Ienny; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: rodolfoienny@gmail.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The study, in situations involving accidents, of heat transfer in fuel rods is of known importance, since it can be used to predict the temperature limits in designing a nuclear reactor, to assist in making more efficient fuel rods, and to increase the knowledge about the behavior of the reactor's components, a crucial aspect for safety analysis. This study was conducted using as parameter the fuel rod that has the highest average power in a typical PWR reactor. For this, we developed a program (Fuel{sub R}od{sub 3}D) in Fortran language using the Finite Elements Method (FEM) for the discretization of a fuel rod and coolant channel, in order to study the temperature distribution in both the fuel rod and the coolant channel. Transient parameters were coupled to the heat transfer equations in order to obtain details of the behavior of the rod and the channel, which allows the analysis of the temperature distribution and its change over time. This work aims to present a study case of an accident where there is a lack of energy in the reactor's coolant pumps and in the diesel engines, resulting in an unplanned shutdown of the reactor. In order to achieve the intended goal, the present work was divided as follows: a short introduction about heat transfer, including the equations concerning the fuel rod and the energy equation in the channel, an explanation about how the verification of the Fuel{sub R}od{sub 3}D program was made, and the analysis of the results. (author)

  15. Study of passive residual heat removal system of a modular small PWR reactor

    International Nuclear Information System (INIS)

    Araujo, Nathália N.; Su, Jian

    2017-01-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

  16. An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements

    International Nuclear Information System (INIS)

    Bowditch, F.H.; Mogford, D.J.

    1987-02-01

    This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m 2 s, and coolant inlet temperature between 150 and 320 0 C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)

  17. Experimental study on heat transfer characteristic of T-junction in cooling system of PWR

    International Nuclear Information System (INIS)

    Wang Haijun; Lu Donghua; Luo Yushan; Chen Tingkuan

    2002-01-01

    The local near wall fluid temperature and heat transfer coefficient distribution in three different areas were measured for T-junction with and without thermal sleeve where jet injected vertically down to the main flow. Flow with velocity ratio between 0.01 and 1.5 were measured. The diameter ratio of jet pipe to main pipe is 0.16. It was found at that downstream of the jet, the thermal sleeve has little effect on heat transfer. At the inlet of the jet and inside the jet tube, the thermal sleeve distinctly affect heat transfer coefficient and near wall fluid temperature. The thermal sleeve's effect is more marked when the velocity ratio becomes larger

  18. ORINC: a one-dimensional implicit approach to the inverse heat conduction problem. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ott, L.J.; Hedrick, R.A.

    1977-10-18

    The report develops an implicit solution technique to determine both the transient surface temperature and the transient surface heat flux of electrically heated rods given the power input and an ''indicated'' internal temperature during a simulated loss-of-coolant accident. A digital computer program ORINC (ORNL Inverse Code) is developed which solves a one-dimensional, transient, lumped parameter, implicit formulation of the conduction equation at each bundle thermocouple position in the Thermal-Hydraulic Test Facility (THTF).

  19. Heat exchanger technologies for the global environment. PWR-Vol. 25

    International Nuclear Information System (INIS)

    Maurer, J.R.

    1994-01-01

    This is the Committee's largest compilation of papers in the fourteen years that ASME has sponsored one or more sessions at the Joint Power Generation Conference. In addition to the twenty-two papers offered at the conference, one session contained a panel discussion during which five industry specialists related their experiences and interpretations. Many of the presentations of the past few years have been made by authors from foreign lands as the conference has expanded to an international forum. This years' theme of global impact is included in the heat exchanger presentations from the USA and abroad. In addition, ASME has encouraged authors to make the subjects of their presentations more apropos to current concerns. New concepts and designs, particularly with PC programs in mind, is one specific area that is addressed. Other topics such as materials selection, coatings to extend performance life, minimizing the effects of microbiologically influenced corrosion, zebra mussel control and new approaches to heat transfer design are some of the subjects covered at the conference and in this volume. Nineteen papers have been processed separately for inclusion in the appropriate data bases

  20. Effectiveness of selected dispersants on magnetite deposition at simulated PWR heat-transfer surfaces

    International Nuclear Information System (INIS)

    Burgmayer, P.; Crovetto, R.; Turner, C.; Klimas, S.J.

    1999-07-01

    The effectiveness of 3 different dispersants-a polyphosphonic acid (PIPPA), a polymethacrylic acid (PMA), and a hydroxyethylidene methacrylic acid (HEME)-at controlling magnetite deposition was examined under steam generator operating conditions. Tests in a cycling research model boiler showed that the dispersants resulted in corrosion products of a smaller average size and a bimodal size distribution. At a concentration in the boiler of 10 mg/kg, density weight deposit on heated probes was reduced 4-, 3-, and 2-fold for PMA, PIPPA, and HEME, respectively. PIPPA was the most effective at increasing iron transport out of the boiler. In deposition loop tests using an 59 Fe radiotracer, only PIPPA and HEME were effective at reducing the particle deposition rate under flow-boiling conditions. None of the dispersants had any effect on deposition under single-phase forced-convective flow. (author)

  1. Effectiveness of selected dispersants on magnetite deposition at simulated PWR heat transfer surfaces

    International Nuclear Information System (INIS)

    Burgmayer, P.; Crovetto, R.; Turner, C.; Klimas, S.

    1998-01-01

    The effectiveness of three different dispersants - a polyphosphonic acid (PIPPA); a polymethacrylic acid (PMA); and a hydroxyethylidene methacrylic acid (HEME) - at controlling magnetite deposition has been examined under steam generator operating conditions. Tests in a cycling research model boiler showed that the dispersants resulted in corrosion products with a smaller average size and a bimodal size distribution. At a concentration in the boiler of 10 mg/kg, density weight deposit on heated probes was reduced 4-, 3-, and 2-fold for PMA, PIPPA, and HEME, respectively. PIPPA was the most effective at increasing iron transport out of the boiler. In deposition loop tests using a 59-Fe radiotracer, only PIPPA and HEME were effective at reducing the particle deposition rate under flow-boiling conditions. None of the dispersants had any impact on deposition under single-phase forced-convective flow. (author)

  2. Effectiveness of selected dispersants on magnetite deposition at simulated PWR heat transfer surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Burgmayer, P.; Crovetto, R. [Betz Dearborn Labs., Revose, PA (United States); Turner, C.; Klimas, S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    The effectiveness of three different dispersants - a polyphosphonic acid (PIPPA); a polymethacrylic acid (PMA); and a hydroxyethylidene methacrylic acid (HEME) - at controlling magnetite deposition has been examined under steam generator operating conditions. Tests in a cycling research model boiler showed that the dispersants resulted in corrosion products with a smaller average size and a bimodal size distribution. At a concentration in the boiler of 10 mg/kg, density weight deposit on heated probes was reduced 4-, 3-, and 2-fold for PMA, PIPPA, and HEME, respectively. PIPPA was the most effective at increasing iron transport out of the boiler. In deposition loop tests using a 59-Fe radiotracer, only PIPPA and HEME were effective at reducing the particle deposition rate under flow-boiling conditions. None of the dispersants had any impact on deposition under single-phase forced-convective flow. (author)

  3. Effectiveness of selected dispersants on magnetite deposition at simulated PWR heat-transfer surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Burgmayer, P.; Crovetto, R. [Betz Dearborn Labs., Revose, PA (United States); Turner, C.; Klimas, S.J

    1999-07-01

    The effectiveness of 3 different dispersants-a polyphosphonic acid (PIPPA), a polymethacrylic acid (PMA), and a hydroxyethylidene methacrylic acid (HEME)-at controlling magnetite deposition was examined under steam generator operating conditions. Tests in a cycling research model boiler showed that the dispersants resulted in corrosion products of a smaller average size and a bimodal size distribution. At a concentration in the boiler of 10 mg/kg, density weight deposit on heated probes was reduced 4-, 3-, and 2-fold for PMA, PIPPA, and HEME, respectively. PIPPA was the most effective at increasing iron transport out of the boiler. In deposition loop tests using an {sup 59}Fe radiotracer, only PIPPA and HEME were effective at reducing the particle deposition rate under flow-boiling conditions. None of the dispersants had any effect on deposition under single-phase forced-convective flow. (author)

  4. 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)

  5. 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.

  6. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  7. LOFT system structural response during subcooled blowdown

    International Nuclear Information System (INIS)

    Martinell, J.S.

    1978-01-01

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

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. Steady-state and transient studies on critical heat flux of a PWR 5 x 5 fuel element bundle with complex spacer wire geometry

    International Nuclear Information System (INIS)

    Fulfs, H.; Katsaounis, A.; Kreubig, M.; Minden, C. von; Orlowski, R.

    1980-01-01

    The results will be described in exemplary presentations completely and concluding. The experimental examination of the steady state simularity of critical heat flux (CHF) in freon 12 and water at identical PWR-5 x 15-rod bundles will show that hot rod/hot channels position as well as CHF can be transformed from model to original fluid with good accuracy. The investigated mass flow and power transients (only in freon 12) point out a definite influence of initial and boundary conditions on CHF and CHF time delay at changing rates higher than 10 to 20%/s. On the contrary simulation of primary pump failure (LOFA) shows no or only small improvement in CHF behaviour while a coupled Scram prevents from reaching the boiling crisis. (orig.) [de

  14. Microstructural characterization of stainless steel 17-4 PH used in the control element of PWR-Type reactors submitted to different heat treatments

    International Nuclear Information System (INIS)

    Ferreira, Douglas F.A.; Rezende, Renato P.; Turcarelli, Tiago

    2017-01-01

    The Control Element is a set of mechanical components of pressurized water cooled nuclear reactors (PWR), with the function of modifying the reactivity of the nucleus by insertion and withdrawal of neutron absorptive rod, in order to change the flow of neutrons (power) to the necessary and desired levels. The control element also has a safety function when there is a need to have negative reactivity available to shut down the reactor in normal operating or accident situations. In this situation, the control element descends instantly and inserts the rods with absorptive material into the fuel element thus shutting down the reactor. The control element consists of control rods, which carry the neutron absorption material and is supported by the spider, pin, spring and spring retainer assembly. The control element has some components that need to have high resistance to impacts when the safety function is activated, so the material of this component must have high mechanical strength and toughness. One of the materials in which can be specified for this application is martensitic stainless steel 17- 4PH (UNS 17400). This steel, when subjected to the aging heat treatment, has its mechanical properties altered due to the precipitation of dispersed intermetallic compounds in the matrix. In all heat treatments performed the predominant microstructure is lath martensite. The heat treatment of the 620 °C / 4 h presented lower hardness when compared to the other treatments and when increase time and temperature the material presents Nb precipitates that increase the hardness. (author)

  15. Microstructural characterization of stainless steel 17-4 PH used in the control element of PWR-Type reactors submitted to different heat treatments

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Douglas F.A.; Rezende, Renato P.; Turcarelli, Tiago, E-mail: ferreira@marinha.mil.br, E-mail: renato.rezende@marinha.mil.br, E-mail: tiago.turcarelli@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (DDNM/CTMSP), São Paulo, SP (Brazil). Diretoria de Desenvolvimento Nuclear da Marinha

    2017-07-01

    The Control Element is a set of mechanical components of pressurized water cooled nuclear reactors (PWR), with the function of modifying the reactivity of the nucleus by insertion and withdrawal of neutron absorptive rod, in order to change the flow of neutrons (power) to the necessary and desired levels. The control element also has a safety function when there is a need to have negative reactivity available to shut down the reactor in normal operating or accident situations. In this situation, the control element descends instantly and inserts the rods with absorptive material into the fuel element thus shutting down the reactor. The control element consists of control rods, which carry the neutron absorption material and is supported by the spider, pin, spring and spring retainer assembly. The control element has some components that need to have high resistance to impacts when the safety function is activated, so the material of this component must have high mechanical strength and toughness. One of the materials in which can be specified for this application is martensitic stainless steel 17- 4PH (UNS 17400). This steel, when subjected to the aging heat treatment, has its mechanical properties altered due to the precipitation of dispersed intermetallic compounds in the matrix. In all heat treatments performed the predominant microstructure is lath martensite. The heat treatment of the 620 °C / 4 h presented lower hardness when compared to the other treatments and when increase time and temperature the material presents Nb precipitates that increase the hardness. (author)

  16. 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

  17. 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

  18. 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

  19. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  20. 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

  1. 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

  2. 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

  3. DRUFAN-01/MOD2, Transient Thermohydraulics of PWR Primary System LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Burwell, M J; Lerchl, G; Steinhoff, F; Wolfert, K [Gesellschaft fuer Reaktorsicherheit (GRS) mbH, Forschungsgelaende, 8046 Garching (Germany)

    1982-12-13

    1 - Description of problem or function: DRUFAN is an advanced best estimate code for simulation of the transient thermal hydraulic behaviour during PWR-blowdown with large break size. 2 - Method of solution: The code is based on the lumped parameter approach and allows flexible control volume configurations. The physical model takes into account thermodynamic nonequilibrium. Using finite difference techniques a 1-dimensional representation of the discharge flow path including geometrical influences is possible. The physical model is based on separated field equations for liquid and vapour mass and overall field equations for energy and momentum. The mass transfer rates between phases during evaporation and condensation are based on correlations for the controlled growth and shrinkage of vapour bubbles or liquid droplets, respectively. A heat conductor model based on the energy transport equation is available for simulation of structures, electrical heater rods and fuel rods. For the heat transfer between solid structures and the fluid a comprehensive package of flow regime dependent heat transfer and critical heat flux correlations can be used. Simulation of components (valve, pressurizer, accumulator, pump, steam generator) is possible with functions or models. Power generation in solid structures may be simulated by an input time function, an electrical heater model or a neutron kinetics models. As a result of the lumped parameter approach a set of ordinary differential equations is obtained from the field equations. These equations, together with those resulting from the simulation of critical discharge flow near the outlet by a finite difference method, are solved by an explicit/implicit integration method with automatic time step, order and error control. The ordinary differential equations representing heat conductors are solved by an essentially implicit integration method. 3 - Restrictions on the complexity of the problem: - Vapour or liquid phase are

  4. 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

  5. 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

  6. Exxon Nuclear Company WREM-based generic PWR ECCS evaluation model. Appendix C to Volume II. Yankee Rowe example problem

    International Nuclear Information System (INIS)

    1975-01-01

    Core 12, of the Yankee Rowe (YR), plant is to be licensed with the Exxon Nuclear Co. PWR Evaluation Model. This appendix presents methodology and results of an example calculation for the YR plant using the ENC Evaluation Model. This example problem is for the double-ended guillotine cold leg break with a discharge coefficient of 0.6 assuming loss of one emergency diesel. The NSSS supplier has determined this case to give the highest peak cladding temperature (PCT) for Core 11. The YR example problem was performed to determine the maximum acceptable local peak heating rate (Kw/ft). The blowdown was performed with beginning-of-cycle (BOC) ENC fuel at full power with the hot assembly power corresponding to the design peak rod heating rate of 12.9 Kw/ft. The HOT CHANNEL, TOODEE2, and RELAP4-FLOOD runs were made at several reduced hot assembly radial peaks, holding axial peaking constant, until an acceptable PCT was achieved. This procedure results in a PCT of 1834 0 F at a reduced peak linear heating rate of 10.5 Kw/ft for the ENC fuel at BOC. (auth)

  7. 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.)

  8. 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

  9. 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

  10. 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

  11. 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.

  12. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  13. Determination of mass flow rate and quality distributions between the subchannels of a heated bundle. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bayoumi, M.; Charlot, R.; Ricque, R.

    1976-05-01

    For analyzing, correlating and extrapolating experimental burn-out results obtained with LWR rod bundles, it is necessary to know the distributions of mass flow rate and quality between the subchannels. A description is presented of an experimental study in progress at the CEN-Grenoble for determining and adjusting the laws of mixing in the FLICA Code which is used to predict these distributions. The experiments are performed on the FRENESIE loop with Freon 12. The test section, in vertical position, consists of a four rod bundle in a channel with square section. The heat flux is axially uniform. The flow of each subchannel can be sampled in ''isokinetic conditions,'' at the end of the heating length. Thermodynamic quality and mass flow rate of the samplings are measured in steady state conditions by using respectively a calorimeter and a turbine flow meter. The test facility is described and experimental data are presented and discussed.

  14. Contribution to the modelling of flows and heat transfers during the reflooding phase of a PWR core

    International Nuclear Information System (INIS)

    Colas, D.

    1984-01-01

    This thesis contributes to modelise thermohydraulic phenomena occuring in a pressurized water nuclear reactor core during the reflood phase of a LOCA. The reference accident and phenomena occuring during reflooding are described as well as flow regime and heat transfer proposed models. With these models, we developed a code to compute fluid conditions and fuel rods temperatures in a reactor core chanel. In order to test this code, results of computation are compared with experiments (FLECHT Skewed Tests) and a conclusion is drawn [fr

  15. 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.)

  16. 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

  17. 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)

  18. 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)

  19. 3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

    International Nuclear Information System (INIS)

    2012-04-01

    Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

  20. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  1. 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)

  2. 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

  3. 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.

  4. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  5. Heat Transfer Modes and their Coefficients for a Passive Containment Cooling System of PWR using a Multi-Pod Heat Pipe

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Gyeongho; Park, Junseok; Kim, Sangnyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-05-15

    If a reactor core is damaged due to a disaster such as happened at TEPCO's Fukushima nuclear power plant, the inevitable rise of super-heated steam that could potentially convert to hydrogen resulting from unimpeded temperature and pressure rises will threaten the integrity of the containment structure. To prevent this, safety and regulatory standards typically specify that the gas vent and external cooling systems be designed to maintain containment up to the level C limit for 24 hours and integrity for 48 hours after any damage to the core. Furthermore, it is recommended that the installation of the exhaust penetration unit have a minimum diameter of 3ft. However, installation of such cooling measures or penetration units is burdensome in terms of operational and maintenance costs not to mention the need to ensure a fleet of fire trucks to be on standby as well as the need to ensure a plentiful supply of water for cooling and a filtration system to clean the water. Therefore, the development of a reliable passive cooling system will be economically advantageous because the extra cost burdens of the external system can be omitted. The Passive Containment Cooling System (PCCS) using a multi-pod heat pipe proposed in this study satisfies these conditions.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.)

  11. 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.

  12. 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

  13. 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

  14. Study of passive residual heat removal system of a modular small PWR reactor; Estudo do sistema passivo de remoção de calor residual de um reator PWR pequeno modular

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Nathália N., E-mail: nathalianunes@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Departamento de Engenharia Nuclear; Faccini, José L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Su, Jian, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS.

  15. 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

  16. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  17. 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)

  18. 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.)

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  4. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  5. 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)

  6. 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)

  7. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  8. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  9. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  10. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  11. 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

  12. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  13. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  14. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  15. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  16. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  17. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  18. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations

  19. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  20. THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2001-01-01

    . - The most important feature of THYDE-W is the conservation of mass, momentum and energy. 2 - Method of solution: In THYDE-P, a PWR plant is regarded as a net- work of various coolant components which may be classified into nodes and junctions. The one-dimensional mass, momentum and energy equations are suitably integrated in each node and junction. In integrating the resulting equations with respect to time, a non- linear implicit method is used on the basis of the Newton method. The Jacobian matrix of the basic equations can be reduced to a simple form by the network theory, which is one of the characteristics of THYDE-P. To solve the basic equations by the non-linear implicit method, various smoothing functions with respect to time are introduced for mode changes such as phase change and flow re- versal. New models for a steam generator and a pressurizer are implemented. A THYDE-P calculation is started by a steady-state adjustment, where the basic equations are exactly solved without time derivatives. THYDE-P is able to calculate through both blowdown and re- fill-reflood phases without any change of models and physical conditions of the coolant. A model which takes non-equilibrium effects into account is newly implemented. 3 - Restrictions on the complexity of the problem: It is required that the network has at least one mixing junction except for the core heatup calculation mode and that a normal node without heat source (or sink) must be placed at both the top and bottom ends of the core. After so reticulating the plant, we have a number of nodes and junctions separately, strictly in numeric order in accordance with the following rules: (a) Normal nodes (except linkage nodes) should be numbered in numerical order chain-wise from one mixing junction to another according to the direction of the steady state chain flow. (b) Then linkage nodes should be numbered in numeric order chain- wise from the corresponding mixing junction. (c) Special nodes should be numbered

  1. 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

  2. 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.)

  3. PWR auxiliary systems, safety and emergency systems, accident analysis, operation

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1976-01-01

    The author presents a description of PWR auxiliary systems like volume control, boric acid control, coolant purification, -degassing, -storage and -treatment system and waste processing systems. Residual heat removal systems, emergency systems and containment designs are discussed. As an accident analysis the author gives a survey over malfunctions and disturbances in the field of reactor operations. (TK) [de

  4. 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

  5. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  6. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

  7. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    reactor coolant system, combustible gas burning, atmosphere- structure heat transfer, ventilation, containment spray cooling, etc. After the molten core penetrates the reactor bottom head, steam generation, concrete disintegration and noncondensable gas generation are calculated in the reactor cavity or the pedestal. 2 - Method of solution: Each of the THALES member codes first establishes the steady state conditions after reading input data. Then iterative time-dependent calculation is continued, taking account of various phenomena and events and their interactions which will occur in the course of a postulated severe accident. The transient calculations are iterated by the physical times specified by input. Generally the RCS thermal hydraulic analysis with the THALES-PM or THALES-BM code is first carried out and its results are transferred to the following containment analysis with the THALES-CV code. Then both results are transferred to a code for analyzing fission product release and transport behavior. Automatic data transfer is possible in the case the JAERI's ART code is used for fission product behavior analysis. In overall thermal hydraulic analysis, a new method is adopted aiming at sufficiently accurate estimation of mixture levels in the reactor coolant system and the containment in a reasonable computer time. The heat transfer calculation in the core is carried out based on the backward method. 3 - Restrictions on the complexity of the problem: Restrictions relating to storage allocation are: (1) Maximum number of radial regions in the core : 10; (2) Maximum number of axial increments in the fuel rods : 50; (3) Maximum number of loops in the PWR primary system : 4; (4) Maximum number of volumes in the PWR primary system : 11; (5) Number of BWR recirculation loops: 2 (fixed); (6) Number of volumes in the BWR reactor coolant system : 7 (fixed); (7) Maximum number of compartments in the containment : 10. There is another restriction, which relates to time step

  8. A nodal model for the simulation of a PWR core

    International Nuclear Information System (INIS)

    Souza Pinto, R. de.

    1981-06-01

    A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

  9. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  10. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  11. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.) [de

  12. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  13. 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

  14. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  15. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  16. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  17. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  18. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

  19. 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)

  20. 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

  1. 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.)

  2. 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

  3. 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

  4. 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

  5. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  6. Experimental simulation of corium dispersion phenomena in direct containment heating

    International Nuclear Information System (INIS)

    Wu, Q.

    1996-01-01

    In a direct containment heating (DCH) accident scenario, the degree of corium dispersion is one of the most significant factors responsible for the reactor containment heating and pressurization. To study the mechanisms of the corium dispersion phenomenon, a DCH separate effect test facility of 1:10 linear scale for Zion PWR geometry is constructed. Experiments are carried out with air-water and air-woods metal simulating steam and molten core materials. The physical process of corium dispersion is studied in detail through various instruments, as well as with flow visualization at several locations. The accident transient begins with the liquid jet discharge at the bottom of the reactor pressure vessel. Once the jet impinges on the cavity bottom floor, it immediately spreads out and moves rapidly to the cavity exit as a film flow. Part of the discharged liquid flows out of the cavity before gas blowdown, and the rest is subjected to the entrainment process due to the high speed gas stream. The liquid film and droplet flows from the reactor cavity will then experience subcompartment trapping and re-entrainment. Consequently, the dispersed liquid droplets that follow the gas stream are transported into the containment atmosphere, resulting in containment heating and pressurization in the prototypic condition. Comprehensive measurements are obtained in this study, including the liquid jet velocity, liquid film thickness and velocity transients in the test cavity, gas velocity and velocity profile in the cavity, droplet size distribution and entrainment rate, and the fraction of dispersed liquid in the containment building. These data are of great importance for better understanding of the corium dispersion mechanisms. (orig.)

  7. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

    Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

    2007-01-01

    In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

  8. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  9. Development on design methodology of PWR passive containment system

    International Nuclear Information System (INIS)

    Lee, Seong Wook

    1998-02-01

    The containment is the most important barrier against the release of radioactive materials into the environment during accident conditions of nuclear power plants. Therefore the development of a reliable containment cooling system is one of key areas in advanced reactor development. To enhance the safety of the containment system, many new containment system designs have been proposed and developed in the world. Several passive containment cooling system (PCCS) concepts for both steel and concrete containment systems are overviewed and assessed comparatively. Major concepts considered are: (a) the spray of water on the outer surface of a steel containment from an elevated tank, (b) an external moat for a steel containment, (c) a suppression pool for a concrete containment, and (d) combination of the internal spray and internal or external condensers for a concrete containment. Emphasis is given to the heat removal principles, the required heat transfer area, system complexity and operational reliability. As one of conceptual design steps of containment, a methodology based on scaling principles is proposed to determine the containment size according to the power level. The AP600 containment system is selected as the reference containment to which the scaling laws are applied. Governing equations of containment pressure are set up in consideration of containment behavior in accident conditions. Then, the dimensionless numbers, which characterize the containment phenomena, are derived for the blowdown dominant and decay heat dominant stage, respectively. The important phenomena in blowdown stage are mass and energy sources and their absorption in containment atmosphere or containment structure, while heat transfer to the outer environment becomes important in decay heat stage. Based on their similarity between the prototype and the model, the containment sizes are determined for higher power levels and are compared with the SPWR containment design values available

  10. Shutdown decay heat removal analysis of a Babcock and Wilcox pressurized water reactor: Case study

    International Nuclear Information System (INIS)

    Cramond, W.R.; Ericson, D.M. Jr.; Sanders, G.A.

    1987-03-01

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

  11. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

    Souza, A.L.; Faya, A.J.G.

    1988-07-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analyzed by a nodal model. Average and hot channels are treated so that bulk response of the core and DNBR can be evaluated. A homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  12. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. 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.)

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. 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

  4. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  5. Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

    International Nuclear Information System (INIS)

    Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

    2002-01-01

    The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

  6. 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)

  7. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  8. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  9. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  10. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  11. 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...

  12. 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...

  13. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M H; Yu, K J; Lee, D J; Cho, B H; Kim, H Y; Yoon, J H; Lee, Y J; Kim, J P; Park, C T; Seo, J K; Kang, H S; Kim, J I; Kim, Y W; Kim, Y H

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  14. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  15. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  1. Reduced scaling of thermal-hydraulic circuits for studies of PWR reactors natural circulation

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1993-01-01

    The Ishii et al. hydrodynamic similarity criteria for natural circulation were used for scaling reduced models of prototype passive residual heat removal system of a 600 M We PWR. The physical scales of the thermohydraulic parameters obtained presented a reasonable agreement when compared with simplified analytic models of the systems. (author)

  2. Corrosion fatigue crack growth of pressure vessel welds in PWR environment

    International Nuclear Information System (INIS)

    Bamford, W.H.; Ceschini, L.J.; Moon, D.M.

    1983-01-01

    The fatigue crack growth rate behavior of several pressure vessel steel welds in PWR environment is discussed. The behavior is compared with associated heat-affected zone behavior, and with comparable base metal results. The welds show different degrees of susceptibility to the environmental influence, and this is discussed in some detail, along with fractographic observations on the tested specimens

  3. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  4. Optimization of control area ventilation systems for Japanese PWR plants

    International Nuclear Information System (INIS)

    Naitoh, T.; Nakahara, Y.

    1987-01-01

    The nuclear power plant has been required to reduce the cost for the purpose of making the low-cost energy since several years ago in Japan. The Heating, Ventilating and Air Conditioning system in the nuclear power plant has been also required to reduce its cost. On the other hand the ventilation system should add the improvable function according to the advanced plant design. In response to these different requirements, the ventilation criteria and the design of the ventilation system have been evaluated and optimized in Japanese PWR Plant design. This paper presents the findings of the authors' study

  5. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  6. Natural-circulation-cooling characteristics during PWR accident simulations

    International Nuclear Information System (INIS)

    Adams, J.P.; McCreery, G.E.; Berta, V.T.

    1983-01-01

    A description of natural circulation cooling characteristics is presented. Data were obtained from several pressurized water reactor accident simulations in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). The reliability of natural circulation cooling, its cooling effectiveness, and the effect of changing system conditions are described. Quantitative comparison of flow rates and time constants with theory for both single- and two-phase fluid conditions were made. It is concluded that natural circulation cooling can be relied on in plant recovery procedures in the absence of forced convection whenever the steam generator heat sink is available

  7. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  8. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  9. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

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

    International Nuclear Information System (INIS)

    Connor, W.M.; Richards, J.E.

    1988-01-01

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

  11. UPTF experiment: Effect of full-scale geometry on countercurrent flow behaviour in PWR downcomer

    International Nuclear Information System (INIS)

    Liebert, J.; Weiss, P.

    1989-01-01

    Four separate effects tests (13 runs) have been performed at UPTF - a 1:1 scale test facility - to investigate the thermal-hydraulic phenomena in the full-scale downcomer of a PWR during end-of-blowdown, refill and reflood phases. Special attention has been paid to the effects of geometry - cold leg arrangement - and ECC-water subcooling on downcomer countercurrent flow and ECC bypass behaviour. A synopsis of the most significant events and a comparison of countercurrent flow limitation (CCFL) data from UPTF and 1/5 scale test facility of Creare are given. The CCFL results of UPTF are compared to data predicted by an empirical correlation developed at Creare, based on the modified dimensionless Wallis parameter J * . A significant effect of cold leg arrangement on CCFL was observed leading to strongly heterogeneous flow condition in the downcomer. CCFL in front of cold leg 1 adjacent to the broken loop exists even for very low steam flow rates. Therefore the benefit of strong water subcooling is not as much as expected. The existing flooding correlation of Creare predicts the full-scale downcomer CCFL insufficiently. New flooding correlations are required to describe the CCFL process adequately. (orig.)

  12. 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

  13. 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

  14. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    Fernandez, J. L.; Lopez, J.

    1986-01-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

  15. 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)

  16. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  17. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  18. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  19. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  20. An analysis of transients in the PWR downcomer

    International Nuclear Information System (INIS)

    Jovanovic, A.

    1981-01-01

    The paper deals with the problem of determining non-stationary temperature field in the downcomer of a PWR type reactor. For this purpose, an analytical model has been developed. The model covers five components of (PWR - Krsko) downcomer: the core-barrel, floor between the core-barrel and the thermal shield, the thermal shield, flow between the thermal shield and the reactor vessel wall, the reactor vessel wall. The model includes internal heat generation in metal structures. The governing equations of the model have been written in the finite difference explicit form. The system of resulting algebraic equations was solved bu Gauss-Seidel method, using a modular computer code. Several characteristic transients were examined (step and continuous change of fluid temperature at the inlet nozzle). Also, an analysis of main parameters (heat transfer coefficient and flow rate) has been performed. The model is intended to be used as basics for further development of a more realistic model that could be used for practical safety analysis. (author)

  1. Four-fluid model of PWR degraded cores

    International Nuclear Information System (INIS)

    Dearing, J.F.

    1985-01-01

    This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

  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. 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.

  5. The physical and chemical degradation of PWR fuel rods in severe accident conditions

    International Nuclear Information System (INIS)

    Parsons, P.D.; Mowat, J.A.S.; Dewhurst, D.W.F.; Hughes, T.E.

    1983-01-01

    An experimental study of the interaction between Zircaloy-4 cladding and UO 2 in PWR fuel rods heated to high temperatures with a negligible differential pressure across the cladding wall is described. The fuel rods were of dimensions appropriate to the 17x17 PWR fuel sub-assembly and were heated in a non-oxidising environment (vacuum) up to approx. 1850 deg. C either isothermally or through heating ramps. Observations were made concerning the extent and nature of the reaction zone between Zircaloy-4 and UO 2 over the temperature range 1500-1850 deg. C for times ranging from 1 min to 125 min. The location, morphology and the chemical composition of the phases formed are described along with the kinetics of their formation. (author)

  6. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  7. Measurement of mist cooling of PWR during LOCA by LDA

    International Nuclear Information System (INIS)

    Lee, S.L.; Sheen, H.J.; Issapour, I.

    1985-01-01

    The prediction of temperature distribution and heat transfer within rod bundles during the refill and reflood phase of a LOCA (loss of coolant accident) is of critical importance for determining the location and size of blockages due to clad deformation in a pressurized water reactor (PWR). Mist cooling by small droplets generated from large droplets on hitting grid spacers has been suggested as one of the most important heat transfer mechanisms which are responsible for the development of this temperature transient. The questions to be asked are whether such small droplets indeed exist and, if so, how are they related to the cooling of the fuel rods. Hereby reported is the result of a direct experimental investigation on these questions by a special laser-Doppler anemometry (LDA) particle sizing technique together with temperature measurements of the rod claddings and flow in the subchannel

  8. Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code

    International Nuclear Information System (INIS)

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

    1987-05-01

    One of the most important unresolved issues governing risk in many nuclear power plants involves the phenomenon called direct containment heating (DCH), in which it is postulated that molten corium ejected under high pressure from the reactor vessel is dispersed into the containment atmosphere, thereby causing sufficient heating and pressurization to threaten containment integrity. Models for the calculation of potential DCH loads have been developed and incorporated into the CONTAIN code for severe accident analysis. Using CONTAIN, DCH scenarios in PWR plants having three different representative containment types have been analyzed: Surry (subatmospheric large dry containment), Sequoyah (ice condenser containment), and Bellefonte (atmospheric large dry containment). A large number of parameter variation and phenomenological uncertainty studies were performed. Response of DCH loads to these variations was found to be quite complex; often the results differ substantially from what has been previously assumed concerning DCH. Containment compartmentalization offers the potential of greatly mitigating DCH loads relative to what might be calculated using single-cell representations of containments, but the actual degree of mitigation to be expected is sensitive to many uncertainties. Dominant uncertainties include hydrogen combustion phenomena in the extreme environments produced by DCH scenarios, and factors which affect the rate of transport of DCH energy to the upper containment. In addition, DCH loads can be aggravated by rapid blowdown of the primary system, co-dispersal of moderate quantities of water with the debris, and quenching of de-entrained debris in water; these factors act by increasing steam flows which, in turn, accelerates energy transport. It may be noted that containment-threatening loads were calculated for a substantial portion of the scenarios treated for some of the plants considered

  9. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  10. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  11. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  12. SCC of Alloy 600 components in PWR primary loop

    International Nuclear Information System (INIS)

    Gomez-Briceno, Dolores; Lapena, Jesus; Castano, M. Luisa; Blazquez, Fernando

    2002-01-01

    Full text: Cracking due to PWSCC in PWR CRDM nozzles and other VHP nozzles fabricated from Alloy 600 is not a new issue. In 1991, a leak was discovered on one CRDM nozzle at Bugey 3 PWR plant in France. The cause of the cracking was identified as primary water stress corrosion cracking. From then, similar cracks have been found in other European and USA PWR plants. The cracks were predominantly axial in orientation and it was accepted that CRDM nozzles and weld cracking in PWR was not a immediate safety concern. However, this consideration has to be reassessed in light of the recent identification of circumferential cracking in CRDM nozzles at Oconee Nuclear Station Unit 2 and 3 along with axial cracking in the Alloy 182 J-groove welds at these two units and at Oconee Nuclear Station 1 and Arkansas Nuclear One Unit 1. Alloy 600 susceptibility in primary water has received an enormous research effort for many years since the Alloy 600 steam generators tube degradation started. A significant amount of information is available to characterise the susceptibility of Alloy 600. However, Alloy 600 susceptibility is strongly dependent on the heat thermomechanical history and both the crack initiation time and the crack growth rate data obtained from representative materials of the VHP nozzles seem to be necessary for the structural integrity assessment of cracking nozzles. An extensive experimental program has been performed at CIEMAT, to study the behaviour of Alloy 600 VHP nozzles in PWR primary conditions. Crack initiation and crack propagation tests have been performed using different types of products (forged bar, tube, plate and steam generator tubing). Long duration crack initiation tests have been carried out, at 330 deg. C and 360 deg. C in water and at 400 deg. C in steam, using ten Alloy 600 heats with yield strength ranging from 291 MPa to 489 MPa. The influence of several parameters (grain boundary carbide distribution, grain size and yield strength) on crack

  13. An integrated PWR for marine propulsion

    International Nuclear Information System (INIS)

    Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

    2008-01-01

    Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

  14. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  15. 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

  16. 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

  17. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  18. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  19. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  20. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Energy Technology Data Exchange (ETDEWEB)

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.

    2014-07-01

    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  1. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

    International Nuclear Information System (INIS)

    Preece, R.J.; Putney, J.M.

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

  2. Heat transfer--Orlando (Symposium), 1980

    International Nuclear Information System (INIS)

    Stein, R.P.

    1980-01-01

    This conference proceedings contains 36 papers of which 3 appear as abstracts. 23 papers are indexed separately. Topics covered include: thermodynamics of PWR and LMFBR Steam Generators; two-phase flow in parallel channels; geothermal heat transfer; natural circulation in complex geometries; heat transfer in non-Newtonian systems; and process heat transfer

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

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200% double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding temperature excursion was terminated at 175 s when complete core quenching occurred. Accumulator flows were initiated at 10 s when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching

  4. A new model for simulation of pressurizers in PWR power plants

    International Nuclear Information System (INIS)

    Madeira, A.A.

    1981-02-01

    The pressurizer of a PWR type reactor was simulated as a thermodynamical system made up of three regions with movable boundaries. The mechanisms of normal condensation, condensation induced by spray, flashing and heat exchange across the water - steam interface, were studied. Various tests have been carried out and satisfactory results were obtained when compared with those from other models and also with some available experimental data. (E.G.) [pt

  5. 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.)

  6. 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

  7. 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.)

  8. Design of large steam turbines for PWR power stations

    International Nuclear Information System (INIS)

    Hobson, G.

    1984-01-01

    The authors review the thermodynamic cycle requirements for use with pressurized-water reactors, outline the way thermal efficiency is maximized, and discuss the special nature of the wet-steam cycle associated with turbines for this type of reactor. Machine and cycle parameters are optimized to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range both of full-speed turbines running at 3000 rev/min on 50 Hz systems and of half-speed turbines running at 1800 rev/min on 60 Hz systems. The importance of service experience with nuclear wet-stream turbines, and its relevance to the design of modern turbines for PWR applications, is discussed. (author)

  9. Analysis of reactivity insertion accidents in PWR reactors

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1978-06-01

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt

  10. Simplified model of a PWR primary coolant circuit

    International Nuclear Information System (INIS)

    Souza, A.L. de; Faya, A.J.G.

    1988-01-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analysed by a nodal model. Average and hot channels are treated so that the bulk response of the core and DNBR can be evaluated. A Homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  11. The design of a compact integral medium size PWR

    International Nuclear Information System (INIS)

    Shirvan, Koroush; Hejzlar, Pavel; Kazimi, Mujid S.

    2012-01-01

    Highlights: ► We model the IRIS reactor in RELAP5 and VIPRE codes. ► We use Printed Circuit Heat Exchangers and internally and externally cooled fuel pins in IRIS. ► We increase the IRIS power by 50% and demonstrate adequate safety performance. ► We show significant potential gain in economics for any integral PWR reactor design. - Abstract: Integral reactors have been proposed in recent years as a means to eliminate loss of coolant events, and reduce the number of large vessels of a nuclear power plant. In this paper the focus on how to further increase the power that can be derived from a given vessel volume. The example is applied to the International Reactor Innovative and Secure (IRIS), a medium size, light water reactor rated at 1000 MWt. The IRIS is an integral design containing all pumps and steam generators along with a traditional PWR core inside the reactor vessel. IRIS was designed with 8 Once-Through Helically Coiled Steam Generators (OTHSG), located above the core, in an annular region between the riser and the pressure vessel wall. This work examines ideas to increase its power output in the same vessel size while maintaining or improving the safety margins. The combination of Printed Circuit Heat Exchangers (PCHE) and Internally and EXternally cooled Annular Fuel (IXAF) is proposed to implement such improvement in otherwise the reference IRIS design. Safety implications of such steam generator and fuel design changes for the same reactor size are examined, under both steady state and transients, using the RELAP5 and VIPRE codes. It is found that the IRIS reactor power can be increased by 50% by using the PCHE and IXAF. The proposed design is found to be less expensive per unit electric power produced, these improvements and analyses can be applied to any integral reactor design.

  12. Thermo-mechanical analysis of PWR bolts susceptible to IASCC

    International Nuclear Information System (INIS)

    Matteoli, C.; Hannink, M.H.C.; Blom, F.J.; Marck, S.C. van der; Charpin-Jacobs, F.

    2015-01-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) is considered a primary ageing issue for the Reactor Pressure Vessel (RPV) internals of Pressurized Water Reactors (PWR). In particular, this complex phenomenon which develops in an environment featuring thermal and mechanical stresses, interaction with corrosive compounds and irradiation, is affecting the bolts connecting the baffles and the formers in the Nuclear Power Plants' RPVs. The baffle-former assembly is the structure that borders the fuel assemblies region, contributing to keep them in position and separating in the radial direction, the core region from the downcomer region. An evaluation of the stresses and temperatures reached in the baffle-former bolts during normal operation was performed by means of a coupled thermo-mechanical study which uses reactor physics calculations to obtain the fluence in the reactor core and as a consequence the heat deposition in the RPV internals. The heat deposition data are coupled with a finite element model of the bolts and the RPV internals in order to perform a complete analysis taking in account thermal, mechanical and radiation loadings. The study is first carried out focusing on a section of the RPV internals, showing a single row of baffle-former bolts. Then the work is extended to the full core height. The model set up in this work, includes an in-depth study of the behavior of the core internals, in particular baffle-former bolts. The model has the capability of understanding the mechanical and thermal behavior of essential internal components in a PWR. (authors)

  13. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  14. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

  15. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  16. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  17. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  18. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  19. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  20. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  1. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  2. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  3. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J L; Lopez, J

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  4. PWR steam generator chemical cleaning. Phase I: solvent and process development. Volume II

    International Nuclear Information System (INIS)

    Larrick, A.P.; Paasch, R.A.; Hall, T.M.; Schneidmiller, D.

    1979-01-01

    A program to demonstrate chemical cleaning methods for removing magnetite corrosion products from the annuli between steam generator tubes and the tube support plates in vertical U-tube steam generators is described. These corrosion products have caused steam generator tube ''denting'' and in some cases have caused tube failures and support plate cracking in several PWR generating plants. Laboratory studies were performed to develop a chemical cleaning solvent and application process for demonstration cleaning of the Indian Point Unit 2 steam generators. The chemical cleaning solvent and application process were successfully pilot-tested by cleaning the secondary side of one of the Indian Point Unit 1 steam generators. Although the Indian Point Unit 1 steam generators do not have a tube denting problem, the pilot test provided for testing of the solvent and process using much of the same equipment and facilities that would be used for the Indian Point Unit 2 demonstration cleaning. The chemical solvent selected for the pilot test was an inhibited 3% citric acid-3% ascorbic acid solution. The application process, injection into the steam generator through the boiler blowdown system and agitation by nitrogen sparging, was tested in a nuclear environment and with corrosion products formed during years of steam generator operation at power. The test demonstrated that the magnetite corrosion products in simulated tube-to-tube support plate annuli can be removed by chemical cleaning; that corrosion resulting from the cleaning is not excessive; and that steam generator cleaning can be accomplished with acceptable levels of radiation exposure to personnel

  5. Predicting critical heat flux in slug flow regime of uniformly heated ...

    African Journals Online (AJOL)

    Numerical computation code (PWR-DNBP) has been developed to predict Critical Heat Flux (CHF) of forced convective flow of water in a vertical heated channel. The code was based on the liquid sub-layer model, with the assumption that CHF occurred when the liquid film thickness between the heated surface and vapour ...

  6. 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

  7. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  8. Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

    International Nuclear Information System (INIS)

    Zhou, Shengcheng; Wu, Hongchun; Zheng, Youqi

    2018-01-01

    Highlights: •ADS reloading scheme is optimized to raise discharge burnup and lower reactivity loss. •ADS is flexible to be combined with various pyro-chemical reprocessing technologies. •ADS is flexible to transmute MAs from different spent PWR fuels. -- Abstract: A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

  9. A systematic approach for development of a PWR cladding corrosion model

    International Nuclear Information System (INIS)

    Quecedo, M.; Serna, J.J.; Weiner, R.A.; Kersting, P.J.

    2001-01-01

    A new model for the in-reactor corrosion of Improved (low-tin) Zircaloy-4 cladding irradiated in commercial pressurized water reactors (PWRs) is described. The model is based on an extensive database of PWR fuel cladding corrosion data from fuel irradiated in commercial reactors, with a range of fuel duty and coolant chemistry control strategies which bracket current PWR fuel management practices. The fuel thermal duty with these current fuel management practices is characterized by a significant amount of sub-cooled nucleate boiling (SNB) during the fuel's residence in-core, and the cladding corrosion model is very sensitive to the coolant heat transfer models used to calculate the coolant temperature at the oxide surface. The systematic approach to developing the new corrosion model therefore began with a review and evaluation of several alternative models for the forced convection and SNB coolant heat transfer. The heat transfer literature is not sufficient to determine which of these heat transfer models is most appropriate for PWR fuel rod operating conditions, and the selection of the coolant heat transfer model used in the new cladding corrosion model has been coupled with a statistical analysis of the in-reactor corrosion enhancement factors and their impact on obtaining the best fit to the cladding corrosion data. The in-reactor corrosion enhancement factors considered in this statistical analysis are based on a review of the current literature for PWR cladding corrosion phenomenology and models. Fuel operating condition factors which this literature review indicated could have a significant effect on the cladding corrosion performance were also evaluated in detail in developing the corrosion model. An iterative least squares fitting procedure was used to obtain the model coefficients and select the coolant heat transfer models and in-reactor corrosion enhancement factors. This statistical procedure was completed with an exhaustive analysis of the model

  10. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water

    International Nuclear Information System (INIS)

    Deleume, J.

    2007-11-01

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. δ phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  11. Experimental research progress on passive safety systems of Chinese advanced PWR

    International Nuclear Information System (INIS)

    Xiao Zejun; Zhuo Wenbin; Zheng Hua; Chen Bingde; Zong Guifang; Jia Dounan

    2003-01-01

    TMI and Chernobyl accidents, having pronounced impact on nuclear industries, triggered the governments as well as interested institutions to devote much attention to the safety of nuclear power plant and public's requirements on nuclear power plant safety were also going to be stricter and stricter. It is obvious that safety level of an ordinary light water reactor is no longer satisfactory to these requirements. Recently, the safety authorities have recommended the implementation of passive system to improve the safety of nuclear reactors. Passive safety system is one of the main differences between Chinese advanced PWR and other conventional PWR. The working principle of passive safety system is to utilize the gravity, natural convection (natural circulation) and stored energy to implement the system's safety function. Reactors with passive safety systems are not only safer, but also more economical. The passive safety system of Chinese advanced PWR is composed of three independent systems, i.e. passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system. This paper is a summary of experimental research progress on passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system

  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. 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.)

  14. 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)

  15. A probabilistic SSYST-3 analysis for a PWR-core during a large break LOCA

    International Nuclear Information System (INIS)

    Schubert, J.D.; Gulden, W.; Jacobs, G.; Meyder, R.; Sengpiel, W.

    1985-05-01

    This report demonstrates the SSYST-3 analysis and application for a German PWR of 1300 MW. The report is concerned with the probabilistic analysis of a PWR core during a loss-of-coolant accident due to a large break. With the probabilistic analysis, the distribution functions of the maximum temperatures and cladding elongations occuring in the core can be calculated. Parameters like rod power, the thermohydraulic boundary conditions, stored energy in the fuel rods and the heat transfer coefficient were found to be the most important. The expected value of core damage was determined to be 2.9% on the base of response surfaces for cladding temperature and strain deduced from SSYST-3 single rod results. (orig./HP) [de

  16. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  17. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  18. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  19. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  20. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  1. 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.)

  2. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2009-08-15

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

  3. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  4. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  5. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  6. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  7. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  8. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  9. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  10. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  11. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  12. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  13. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  14. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  15. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  16. An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

    International Nuclear Information System (INIS)

    Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

    2014-01-01

    Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

  17. Zircaloy-4 corrosion in PWR's

    International Nuclear Information System (INIS)

    Fyfitch, S.; Smalley, W.R.; Roberts, E.

    1985-01-01

    Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

  18. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  19. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  20. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  1. Probabilistic analysis of 900 MWe PWR. Shutdown technical specifications

    International Nuclear Information System (INIS)

    Mattei, J.M.; Bars, G.

    1987-11-01

    During annual shutdown, preventive maintenance and modifications which are made on PWRs cause scheduled unavailabilities of equipment or systems which might harm the safety of the installation, in spite of the low level of decay heat during this period. The pumps in the auxiliary feedwater system, component cooling water system, service water system, the water injection arrays (LPIS, HPIS, CVCS), and the containment spray system may have scheduled unavailability, as well as the power supply of the electricity boards. The EDF utility is aware of the risks related to these situations for which accident procedures have been set up and hence has proposed limiting downtime for this equipment during the shutdown period, through technical specifications. The project defines the equipment required to ensure the functions important for safety during the various shutdown phases (criticality, water inventory, evacuation of decay heat, containment). In order to be able to judge the acceptability of these specifications, the IPSN, the technical support of the Service Central de Surete des Installations Nucleaires, has used probabilistic methodology to analyse the impact on the core melt probability of these specifications, for a French 900 MWe PWR

  2. 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

  3. Failure of PWR-RHRS under cold shutdown conditions: Experimental results from the PKL test facility

    International Nuclear Information System (INIS)

    Mandl, R.M.; Umminger, K.J.; Logt, J.V.D.

    1991-01-01

    The Residual Heat Removal System (RHRS) of a PWR is designed to transfer thermal energy from the core after plant shutdown and maintain the plant in cold shutdown or refuelling conditions for extended periods of time. Initial reactor cooling after shutdown is achieved by dissipating heat through the steam generators (SGs) and discharging steam to the condenser by means of the Turbine Bypass System (TBS). When the reactor coolant temperature has dropped to about 160C and pressure has been reduced to 30 bar the RHRS is placed into operation. it reduces the coolant temperature to 50C within 20 hours after shutdown. The time margin for establishing alternate methods of heat removal following a failure of the RHRS depends on the Reactor Coolant System (RCS) temperature, the decay heat rate and the amount of RCS inventory. During some shutdown operations the RCS may be partially drained (e. g. to perform SG inspections). Decreased primary system inventory can significantly reduce the time available to recover the RHRS's function prior to bulk boiling and possible core uncovery. In the PKL test facility, which simulates a 1,300 MWe 4-loop PWR on a scale 1:145, a failure of RHRS under cold shutdown conditions was performed. This presentation gives a brief description of the test facility followed by the test objectives and results of this experiment

  4. 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)

  5. 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

  6. Layout of the primary circuit with its components for PWR and BWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1981-01-01

    The light water-moderated and cooled pressurized water reactors and boiling water reactors constitute the basis of economic utilization of nuclear energy all over the world. Pressurized water reactors up to capacities of 3,800 MWth are those most used for power generation. However, their potential capacities exceed 3,800 MWth, so that already in the near future PWR are conseivable which readily generate 1,500 to 2,000 MWe. The main problem for starting the next generation of PWRs are of safety measure and licensing questions. Interesting applications of the PWRs are nuclear district heating, generation of process steam of desalination plants, steam injection into the ground for oil production or chemical factories. A new generation of natural circulation boiling water reactors with a capacity of 200 to 400 MW will be used for development of small industrial areas or for countries without an integral grid system. The natural circulation boiling water reactor will be subject of a separate lecture. Due to the fact of the majority of the PWR all over the world this lecture will discuss mainly PWR design aspects. (orig./RW)

  7. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  8. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  9. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  10. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  11. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  12. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1980-10-01

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  13. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  14. Technical specifications for PWR secondary water chemistry

    International Nuclear Information System (INIS)

    Weeks, J.R.; van Rooyen, D.

    1977-08-01

    The bases for establishing Technical Specifications for PWR secondary water chemistry are reviewed. Whereas extremely stringent control of secondary water needs to be maintained to prevent denting in some units, sound bases for establishing limits that will prevent stress corrosion, wastage, and denting do not exist at the present time. This area is being examined very thoroughly by industry-sponsored research programs. Based on the evidence available to date, short term control limits are suggested; establishment of these or other limits as Technical Specifications is not recommended until the results of the research programs have been obtained and evaluated

  15. Technical basis for PWR emergency plans forming

    International Nuclear Information System (INIS)

    L'Homme, A.; Manesse, D.; Gauvain, J.; Crabol, B.

    1989-10-01

    Our speech first summarizes the french approach concerning the management of severe accidents which could occur on PWR stations. Then it defines the source-term which is being used as a general support for elaborating the emergency plans devoted to the protection of the population. It describes next the consequences of this source-term on the site and in the environment, which constitute the technical bases for defining actions of utilities and concerned authorities. It gives lastly information on the present status of the different emergency plans and the complementary work undertaken to improve them [fr

  16. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  17. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Richards, J.E.; Byers, W.A.

    1986-07-01

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  18. Microcomputer simulation of PWR power plant pressurizer

    International Nuclear Information System (INIS)

    Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

    1990-01-01

    It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

  19. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  20. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  1. PWR life time: the EDF project

    International Nuclear Information System (INIS)

    Noel, R.; Reynes, L.; Mercier, J.P.

    1987-01-01

    Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

  2. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  3. Environmental surveillance of PWR power stations

    International Nuclear Information System (INIS)

    Conti, M.

    1980-01-01

    The action of Electricite de France with respect to the environment of PWR nuclear power stations is essentially centred on prevention. Controls are carried out at two levels: - before the power station goes on stream (radioecological study), - when the power station is operational. The purpose of the controls effected on the radioactive effluents and the environment is to check that the maximum discharge rate stipulated in the corresponding orders is complied with and to ensure that there are no anomalies in the environment [fr

  4. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  5. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

    Ocken, H.; Fruzzetti, K.; Frattini, P.; Wood, C.J.

    2002-01-01

    Zinc injection to the reactor coolant system (RCS) of PWRs holds the promise to alleviate two key challenges facing PWR plant operators: (1) reducing degradation of coolant system materials, including nickel-base alloy tubing and lower alloy penetrations due to stress corrosion cracking, and (2) lowering shutdown dose rates. Primary water stress corrosion cracking (PWSCC) is a dominant tube failure mode at many plants. This paper summarizes recent observations from U. S. and international PWRs that have implemented zinc injection, focusing primarily on coolant chemistry and dose rate issues. It also provides a look at the future direction of EPRI-sponsored projects on this topic. (authors)

  6. 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

  7. 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

  8. Simple analysis of very long term proceses without operational and emergency energy supply in the PWR power plant

    International Nuclear Information System (INIS)

    Benedek, S.

    1983-01-01

    Published calculational methods are cited and used for examination of PWR transients after a loss-of-coolant accident. For different sizes of breaks and breakdown of the pumps the long term transients - without operational and emergency power supply - were calculated. The results show the critical time interval until the operational or emergency/safety water pump/supply should be made into operation to avoid the core heat-up, melt down and the large radioactive issue. (orig.)

  9. An homogeneous model of steam generator to simulate operational transiento and accidents in PWR nuclear power plants

    International Nuclear Information System (INIS)

    Souza, A.L. de.

    1981-07-01

    GEVAP - A digital computer code was developed to simulate the thermodynamic transient behaviour of steam generators. The steam generator is divided in heating sections. In each section, the conservation equations of mass and energy are integrated numerically, using a predictor-corrector method. As good reslts where obtained, as compared to transients simulated using more detainled codes, it is concluded that GEVAP can be included as the steam generator module of a more complete systems simulation code for PWR's. (E.G.) [pt

  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. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

    A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

  12. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  13. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  14. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2003-01-01

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  15. 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.

  16. Study on external reactor vessel cooling capacity for advanced large size PWR

    International Nuclear Information System (INIS)

    Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

    2014-01-01

    External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

  17. The significance of thermohydraulic conditions for the corrosion safety of PWR steam generators

    International Nuclear Information System (INIS)

    Gulich, J.F.

    1975-04-01

    In several PWR nuclear power plants leakages have occurred in the steam generator which were caused by localised corrosion attack. While the attention of manufacturers and operators is focused on the influences of feedwater chemistry and tube material, the present work highlights the fact that the damage always occurred in those places where flow regimed are poorly defined. The investigation leads to the result that local dry out of the heating surface can be contributing cause of damage. A method is indicated for estimating the thermohydraulic conditions in the inflow region over the tube plate and measures to improve corrosion safety are discussed. (author)

  18. Sensitivity analysis on hot channel of PWR type reactors using matricial formalism

    International Nuclear Information System (INIS)

    Maciel, Edisson Savio G.; Andrade Lima, Fernando Roberto de; Lira, Carlos Alberto B.O.

    1995-01-01

    The matricial formalism of the perturbation theory is applied in a simplified model to study the hot channel of PWR reactors. Mass, linear momentum and energy conservation equations and appropriated heat transfer and fluid mechanics correlations describe the discretized system. After calculating system's thermalhydraulic properties, the matricial formalism is applied and the sensitivity coefficients are determined for each case of interest. Comparisons between perturbative method and direct results of the model have shown good agreement which demonstrates that the matricial formalism is an important tool for discretized system analysis. (author). 6 refs, 2 tabs

  19. ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA

    International Nuclear Information System (INIS)

    1978-01-01

    1 - Nature of the physical problem solved: ASCOT-1 is used to analyze the thermo-hydraulic behaviour in a PWR core during a loss-of-coolant accident. 2 - Method of solution: The core is assumed to be axisymmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of fuel in the annular regions into which the core is divided, the heat conduction equations are solved by an explicit method with averaged flow conditions. 3 - Restrictions on the complexity of the problem: Axisymmetric two-dimensional homogeneous flows

  20. The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

  1. A feasibility assessment for incorporating of passive RHRS into large scale active PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S O; Sub, S Y; Kim, Y S; Chang, M H; Park, J K [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-12-01

    A feasibility study was carried out for the possible incorporation of passive RHRS (Residual Heat Removal System) into a large-scale of active PWR plant. Four kinds of system configurations were considered. For each case its performance and impacts on plant safety, cost, licensing, operation and maintenance were evaluated. The evaluation came up with a finding of PRHRS with a gravity feed tank as most probable design concept. However, considering rearrangement of structure and pipe routing inside and outside containment, it is concluded that implementation of the PRHRS concept into well developed active plants is not desirable at present. (author). 6 refs, 7 figs, 1 tab.

  2. Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor

    International Nuclear Information System (INIS)

    Braga, C.V.M.; Carajilescov, P.

    1981-01-01

    Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt

  3. A feasibility assessment for incorporating of passive RHRS into large scale active PWR

    International Nuclear Information System (INIS)

    Kim, S.O.; Sub, S.Y.; Kim, Y.S.; Chang, M.H.; Park, J.K.

    1996-01-01

    A feasibility study was carried out for the possible incorporation of passive RHRS (Residual Heat Removal System) into a large-scale of active PWR plant. Four kinds of system configurations were considered. For each case its performance and impacts on plant safety, cost, licensing, operation and maintenance were evaluated. The evaluation came up with a finding of PRHRS with a gravity feed tank as most probable design concept. However, considering rearrangement of structure and pipe routing inside and outside containment, it is concluded that implementation of the PRHRS concept into well developed active plants is not desirable at present. (author). 6 refs, 7 figs, 1 tab

  4. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  5. 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

  6. Optimal design of passive containment cooling system for innovative PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon [Central Research Institute, Korea Hydro and Nuclear Power, Ltd., Daejeon (Korea, Republic of)

    2017-08-15

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  7. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  8. Advanced methods for the study of PWR cores

    International Nuclear Information System (INIS)

    Lambert, M.; Salvatores, St.; Ferrier, A.; Pelet, J.; Nicaise, N.; Pouliquen, J.Y.; Foret, F.; Chauliac, C.; Johner, J.; Cohen, Ch.

    2003-01-01

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in October 2003. The transparencies of the annual meeting are presented in the introductive part: 1 - status of the French nuclear park: nuclear energy results, management of an exceptional climatic situation: the heat wave of summer 2003 and the power generation (J.C. Barral); 2 - status of the research on controlled thermonuclear fusion (J. Johner). Then follows the technical session about the advanced methods for the study of PWR reactor cores: 1 - the evolution approach of study methodologies (M. Lambert, J. Pelet); 2 - the point of view of the nuclear safety authority (D. Brenot); 3 - the improved decoupled methodology for the steam pipe rupture (S. Salvatores, J.Y. Pouliquen); 4 - the MIR method for the pellet-clad interaction (renovated IPG methodology) (E. Baud, C. Royere); 5 - the improved fuel management (IFM) studies for Koeberg (C. Cohen); 6 - principle of the methods of accident study implemented for the European pressurized reactor (EPR) (F. Foret, A. Ferrier); 7 - accident studies with the EPR, steam pipe rupture (N. Nicaise, S. Salvatores); 8 - the co-development platform, a new generation of software tools for the new methodologies (C. Chauliac). (J.S.)

  9. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  10. Optimal design of passive containment cooling system for innovative PWR

    International Nuclear Information System (INIS)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon

    2017-01-01

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed

  11. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

    Mann, C.A.; Hindle, E.D.; Parsons, P.D.

    1982-04-01

    Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 1500 0 C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

  12. Water chemistry control of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Hino, Yuichi; Makino, Ichiro; Yamauchi, Sumio; Fukuda, Fumihito.

    1992-01-01

    In PWR power plants, the primary system taking heat out of nuclear reactors and the secondary system generating steam and driving turbines are completely separated by steam generators, accordingly, by mutually independent water treatment, both systems are to be maintained in the optimal conditions. Namely, primary system is the closed water circulation circuit of simple liquid phase though under high temperature, high pressure condition, therefore, water shows the stable physical and chemical properties, and the minute water treatment for restraining the corrosion of structural materials and reducing radioactivity can be done. Secondary system is similar to the condensate and feedwater system of thermal power plants, and is the circuit for liquid-vapor two-phase transformation, but due to the local concentration of impurities by evaporation, the strict requirement is set for secondary water quality. However, secondary system can be treated in the state without radioactivity, and this is a great merit. The outline, basic concept and execution of primary water quality control, and the outline, concept, control criteria, facilities and execution of secondary water quality control are reported. (K.I.)

  13. Oxygen control in PWR secondary coolant: Final report

    International Nuclear Information System (INIS)

    Oliker, I.; Shaikh, B.

    1988-12-01

    The objective of the study is to assess the technical aspects of utilizing direct contact heaters in the condensate train and a bubbling deaerating device in the condenser in order to improve the deaeration in the feedwater system, and develop cost estimates for such a plant retrofit. A reference PWR plant has been used to establish a basis for development of retrofit requirements and system costs. Emphasis has been placed on retrofitting the existing low pressure heaters in the condenser neck into direct contact heaters in order to improve the condensate deaeration at the very beginning of the condensate-feedwater cycle. Two basic designs of direct contact heater are discussed and their technical benefits are described. Required plant modifications have been developed and improvement in heat rates have been evaluated. Cost estimates have been developed for such a retrofit. Turbine protection against water induction has been addressed at length including recommendations on various protection measures. Incorporation of a bubbling deaerating device into the reference plant has been evaluated, and its cost estimate has been developed. 4 refs., 18 figs., 1 tab

  14. Method of stopping operation of PWR type reactor

    International Nuclear Information System (INIS)

    Ueno, Takashi; Tsuge, Ayao; Kawanishi, Yasuhira; Onimura, Kichiro; Kadokami, Akira.

    1989-01-01

    In PWR type reactors after long period of l00 % power operation, since boiling is caused in heat conduction pipes and water is depleted within the intergranular corrosion fracture face in the crevis portion to result in a dry-out state, impregnation and concentration of corrosion inhibitors into the intergranular corrosion fracture face are insufficient. In view of the above, the corrosion inhibitor at a high concentration is impregnated into the intergranular corrosion fracture face by keeping to inject the corrosion inhibitor from l00 % thermal power load by way of the thermal power reduction to the zero power state upon operatioin shutdown. That is, if the thermal power is reduced to or near the 0 power upon reactor shutdown, feedwater in the crevis portion is put to subcooled state, by which the steam present in the intergranular corrosion fracture face are condensated and the corrosion inhibitor at high concentration impregnated into the crevis portion are penetrated into the intergranular corrosion fracture face. (K.M.)

  15. Tendency of nuclear pumps for PWR primary system

    International Nuclear Information System (INIS)

    Shibata, Takeshi

    1976-01-01

    At present, large PWR power stations of more than 1,000 MW are successively constructed, and the pumps used there have become large. The progress and tendency of the technical development of main pumps in primary system are described. The increase of the capacity of power stations is accomplished by increasing the circulating coolant quantity per loop or the number of loops. Same standard primary coolant pumps are employed in the plants from 500 to 1,100 MW. The type of primary coolant pumps changed from canned type to shaft seal type, and the advantages of the shaft seal type are cheap production cost, high efficiency, and the easy utilization of inertia force. The bearings and shaft seals are thermally insulated from primary coolant. As for auxiliary pumps, reciprocating filling-up pumps and centrifugal high pressure injection pumps are used for 500 MW plants, but only centrifugal pumps are used for both purposes in 800 MW plants, and in 1,100 MW plants, the pumps of both types for separate purposes and centrifugal pumps for combined purposes are installed. Horizontal or vertical pumps of same type are used as containment vessel-spraying pumps and excess heat-eliminating pumps. The type of boric acid pumps changed from canned type to mechanical seal type. (Kako, I.)

  16. Reentrainment of droplet from grid spacer in mist flow portion of LOCA reflood of PWR

    International Nuclear Information System (INIS)

    Lee, S.L.; Cho, S.K.; Sheen, H.J.

    1983-01-01

    An investigation is made on the influence of a quenched grid spacer on the greatly enhanced heat transfer from heated fuel rods during the mist flow phase of emergency reflood of loss of coolant accident (LOCA) of a pressurized water reactor (PWR). The situation for the case of a dry grid spacer before its quenching has not been covered in this study. The experimental technique used is a relatively simple optical scheme which combines the reference-mode laser-Doppler anemometry making use of the scattering of a light beam from a droplet. The results reveal that the large droplets in the mist flow, which are intercepted by the grid spacer, are responsible for the creation of a large number of smaller droplets. These small droplets, due to their large surface area to mass ratios, can serve as superb evaporative cooling agents to heat transfer downstream of the grid spacer

  17. PRETTA:A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS

    Institute of Scientific and Technical Information of China (English)

    阿谢德; 徐济鋆

    2001-01-01

    A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

  18. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  19. Definition of thermal-hydraulics parameters of a naval PWR via energy balance of a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chaves, Luiz C.; Curi, Marcos F., E-mail: marcos.curi@cefet-rj.br [Centro Federal de Educação Tecnológica Celso Suckow da Fonseca (CEFET-RJ), Rio de Janeiro, RJ (Brazil). Department of Mechanical Engineering

    2017-07-01

    In this work, we used the operational parameters of the Angra 1 nuclear power plant, designed by Westinghouse, to estimate the thermal-hydraulic parameters for naval nuclear propulsion, focusing on the analysis of the reactor and steam generator. A thermodynamics analysis was made to reach the operational parameters of primary circuit such as pressure, temperature, power generated among others. Previous studies available in literature of 2-loop Westinghouse Nuclear Power Plants, which is based on a PWR and similar to Angra-1, support this analysis in the sense of a correct procedure to deal with many complex processes to energy generation from a nuclear source. Temperature profiles in reactor and steam generator were studied with concepts of heat transfer, fluid mechanics and also some concepts of nuclear systems, showing the behavior into them. In this simulation, the Angra 1 primary circuit was reduced on a scale of 1: 3.5 to fit in a Scorpène-class submarine. The reactor generates 85.7 MW of total thermal power. The maximum power and temperatures reached were lower than the operational safe limits established by Westinghouse. The number of tubes of the steam generator was determined in 990 U-tubes with 6.3 m of average length. (author)

  20. The simulation research for the dynamic performance of integrated PWR

    International Nuclear Information System (INIS)

    Yuan Jiandong; Xia Guoqing; Fu Mingyu

    2005-01-01

    The mathematical model of the reactor core of integrated PWR has been studied and simplified properly. With the lumped parameter method, authors have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, authors have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct. (authors)

  1. Computer code for thermal-hydraulic simulation of heat pressurizer tanks operation (Simterm-H)

    International Nuclear Information System (INIS)

    Sellos, R.F.

    1987-01-01

    It is presented the Simtherm-H computer code, developed for calculating the thermodynamic properties of the high pressure heating system and the feedwater tank in transient state for PWR nuclear power plants (1300 MWe). (E.G.) [pt

  2. Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

  3. Stress corrosion cracking of nickel base alloys in PWR primary water

    International Nuclear Information System (INIS)

    Guerre, C.; Chaumun, E.; Crepin, J.; De Curieres, I.; Duhamel, C.; Heripre, E.; Herms, E.; Laghoutaris, P.; Molins, R.; Sennour, M.; Vaillant, F.

    2013-01-01

    Stress corrosion cracking (SCC) of nickel base alloys and associated weld metals in primary water is one of the major concerns for pressurized water reactors (PWR). Since the 90's, highly cold-worked stainless steels (non-sensitized) were also found to be susceptible to SCC in PWR primary water ([1], [2], [3]). In the context of the life extension of pressurized water reactors, laboratory studies are performed in order to evaluate the SCC behaviour of components made of nickel base alloys and of stainless steels. Some examples of these laboratory studies performed at CEA will be given in the talk. This presentation deals with both initiation and propagation of stress corrosion cracks. The aims of these studies is, on one hand, to obtain more data regarding initiation time or crack growth rate and, one the other hand, to improve our knowledge of the SCC mechanisms. The aim of these approaches is to model SCC and to predict components life duration. Crack growth rate (CGR) tests on Alloy 82 with and without post weld heat treatment are performed in PWR primary water (Figure 1). The heat treatment seems to be highly beneficial by decreasing the CGR. This result could be explained by the effect of thermal treatment on the grain boundary nano-scopic precipitation in Alloy 82 [4]. The susceptibility to SCC of cold worked austenitic stainless steels is also studied. It is shown that for a given cold-working procedure, SCC susceptibility increases with increasing cold-work ([2], [5]). Despite the fact that the SCC behaviour of Alloy 600 has been widely studied for many years, recent laboratory experiments and analysis ([6], [7], [8]) showed that oxygen diffusion is not a rate-limiting step in the SCC mechanism and that chromium diffusion in the bulk close the crack tip could be a key parameter. (authors)

  4. Secondary water chemistry control practices and results of the Japanese PWR plants

    International Nuclear Information System (INIS)

    Maeda, Akihiro; Shoda, Yasuhiko; Ishihara, Nobuo; Murata, Kazutoyo; Fujiwara, Hiroyuki; Hayakawa, Hitoshi; Matsuda, Tadashi

    2012-09-01

    In Japan, since the start of the operation of the first PWR plant, Mihama Unit-1 in 1970, 24 PWR plants have been built by 2010, and all of them are in operation. Due to the plant-specific needs of management, and by flexibly incorporating the state-of-the-art insights into the design, the system configurations of the plants vary so many as 15 types. Meanwhile, the geographical feature of Japan makes all the Japanese PWR plants to have condensers cooled by sea water, and all the plants have a common system with a full-flow Condensate Polisher System (CPS). To prevent corrosion, continued improvements of the secondary water chemistry management has been performed like other countries, and one of the major features of the Japanese PWR plants is an enhanced provision for the condenser leakage. The water quality of SG (Steam Generator) has been significantly improved by the provision for the sea water leakage, in combination with other improvements in water chemistry management. Also in Japan, almost all of the treatments of the spent polisher resin and the wastewater are performed within the power plant sites. To facilitate the treatment of the waste water and the regeneration of the spent resins, either ammonia or ETA (Ethanol Amine) is selected as the pH adjustment agent for the secondary system water. Also at the ammonia treatment, high pH accomplishes the inhibition of the piping wall thinning and the lower iron transportation into SGs. In addition, the iron transported into the SG is removed by the chemical conditioning treatment called ASCA (Advanced Scale Conditioning Agent). This provides the effective recovery of the SG heat-transfer performance, and the improved SG support plate BEC (Broached Egg Crate) hole blockage rates. Basically in Japan, the secondary water chemistry management has been improved based on a single basic specification, for the variety of the plant configurations, with the plant-specific investigations and analyses. This paper summarizes

  5. A universal PWR spectral history correction

    International Nuclear Information System (INIS)

    Hutt, P.K.; Nunn, D.L.

    1989-01-01

    The accuracy of a form of universal correction for the difference between depletion conditions assumed in PWR assembly lattice calculations and those experienced in a reactor burn-up is investigated. The correction is based on lattice calculations in which only one such depletion history difference, depletion at two different water densities, is explicitly represented by lattice calculations. The assumption is made that other historical effects bear the same relationship to an appropriate time-average of the two-group neutron flux spectrum. The correction is shown to be accurate for the most important historical effects, depletion with burnable absorbers inserted, control rods inserted or at a different soluble boron level, in addition to density itself. The correction is less accurate for representing depletion at a different fuel or coolant temperature but even in these cases gives an improvement over no correction. In addition it is argued that these historic temperature effects are likely to be of minor importance. (author)

  6. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    Gomes, I.C.

    1983-01-01

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

  7. Chemical cleaning of nuclear (PWR) steam generators

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Mundis, J.A.

    1982-01-01

    This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

  8. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  9. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  10. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  11. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    Cao, L.; Wu, H.; Zheng, Y.

    2012-01-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  12. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  13. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  14. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  15. Development of advanced PWR steam generator

    International Nuclear Information System (INIS)

    Saito, Itaru; Nakamura, Tomomichi

    1999-01-01

    In response to the increased power of the advanced PWR, it is necessary to develop a steam generator (SG) which has a large capacity with high performance and high reliability as well as being economical to produce. In this paper, the development of the design of a new SG for the advanced PWRs is described and compared with the design of a conventional SG. Moreover, an outline of a seismic verification test for the U-bend tube bundle which includes advanced anti-vibration bars (AVB) which are very important is described. As a result, it was verified that the bundle has sufficient strength and a relatively high attenuation to seismic loads. These results will be reflected in the detailed design of advanced AVBs. (author)

  16. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

    1989-01-01

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

  17. Changes in 900 MW PWR alarm processing policy

    Energy Technology Data Exchange (ETDEWEB)

    Pont, M [Electricite de France, Generation and Transmission, Nuclear Power Plant Operations, Paris (France)

    1997-09-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs.

  18. Changes in 900 MW PWR alarm processing policy

    International Nuclear Information System (INIS)

    Pont, M.

    1997-01-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs

  19. Deboration in nuclear stations of the PWR type

    International Nuclear Information System (INIS)

    1978-01-01

    Reactivity control in nuclear power stations of the PWR type is realised with boric acid. A method to concentrate boric acid without an evaporator has been studied. A flow-sheet with reverse osmosis is proposed. (author)

  20. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  1. Dose rate evaluation after accident in a PWR

    International Nuclear Information System (INIS)

    Cladel, C.; Duchemin, B.; Le Dieu de Ville, A.; Nimal, B.; Nimal, J.C.; Evrard, J.M.

    1983-05-01

    A calculation scheme for the gamma radiation dose rate after accident in a PWR is presented. These studies use a fine description of the geometry and of the fission product inventory. Some results are given and some improvements are planned

  2. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  3. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  4. Development of improved SGV480 steel plate for containment vessel in PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, Norioki [Advanced Nuclear Equipment Research Inst., Tokyo (Japan); Morikage, Yasushi; Okayama, Yutaka; Higashikubo, Tomohiro

    2001-01-01

    When a nuclear containment vessel made of steel plate at PWR plants in Japan is produced, SGV480 steel plate made by annealing method according to JIS G3118 is usually used in main. And, when thickness of welding portion of the vessel is larger than 38 mm, as heat treatment after welding is regulated to carry out according to the ministerial ordinance, it is difficult in actual to carry out the heat treatment of the actual welded portions. In a leading plant, approval of welding using a special method without heat treatment less than 47.25 mm of SGV480 carbon steel plate for JIS G3118 middle and ordinary pressure vessel was carried out to supply it for actual use. And, it is required for protection of welding fracture to carry out pre-heat treatment before welding. Because of increasing plate thickness requiring for lower temperature and more seismic resistance in construction condition, in order to produce a containment vessel without heat treatment after welding, more toughness is required for using material and welded portion. Therefore, a new SGV480 steel plate was developed by using TMCP method of modern steel manufacturing technology, to establish lower carbon equivalence and finer texture with upgrading of both toughness and weldability, without heat treatment after welding and pre-heat treatment before welding, at the Shin-Nippon Steel Co, Ltd. and Kawasaki Steel, Co. Ltd., respectively. (G.K.)

  5. Application of the MELCOR code to design basis PWR large dry containment analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  6. Experimental results of the effective water head in downcomer during reflood phase of a PWR LOCA

    International Nuclear Information System (INIS)

    Sudo, Yukio; Murao, Yoshio; Akimoto, Hajime

    1980-08-01

    The results and analysis of an experiment for the effective water head in downcomer with 50mm gap size are described. The main objective of the experiment was to clarify the effect of gap size on reflooding in a PWR LOCA. The effective water head in downcomer is the driving force for feeding emergency coolant into the core during reflood phase of a PWR LOCA. Discussions presented here follow those of a previous report in which experimental results and analysis were described for the case of 200mm gap size. Experimental Conditions were: Initial Wall Temperature = 200 -- 300 0 C, Back Pressure = 1 atm., Coolant Temperature = 71 -- 100 0 C, Extraction Water Velocity = 0 -- 2 cm/s, Gap Size = 50 mm. The effective water head history obtained in the experiment was compared with those predicted with Sudo's void fraction correlation. In the prediction, heat input to coolant was calculated from the response of measured wall temperature with heat condition analysis. The experimental results and analysis reveals that: (1) The effects of the gap size and initial wall temperature are evident, (2) The effect of extraction water velocity is negligible, and (3) The predicted history of effective water head is in good agreement with the experimental results except during the transient period in which the effective water head is descreasing. (author)

  7. A genetic algorithm applied to a PWR turbine extraction optimization to increase cycle efficiency

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Schirru, Roberto

    2002-01-01

    In nuclear power plants feedwater heaters are used to heat feedwater from its temperature leaving the condenser to final feedwater temperature using steam extracted from various stages of the turbines. The purpose of this process is to increase cycle efficiency. The determination of the optimal fraction of mass flow rate to be extracted from each stage of the turbines is a complex optimization problem. This kind of problem has been efficiently solved by means of evolutionary computation techniques, such as Genetic Algorithms (GAs). GAs, which are systems based upon principles from biological genetics, have been successfully applied to several combinatorial optimization problems in nuclear engineering, as the nuclear fuel reload optimization problem. We introduce the use of GAs in cycle efficiency optimization by finding an optimal combination of turbine extractions. In order to demonstrate the effectiveness of our approach, we have chosen a typical PWR as case study. The secondary side of the PWR was simulated using PEPSE, which is a modeling tool used to perform integrated heat balances for power plants. The results indicate that the GA is a quite promising tool for cycle efficiency optimization. (author)

  8. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  9. GO evaluation of a PWR spray system. Final report

    International Nuclear Information System (INIS)

    Long, W.T.

    1975-08-01

    GO is a reliability analysis methodology developed over the years from 1960 to the present by Kaman Sciences Corporation, Colorado Springs, Colorado. In this report the GO methodology is presented and its application demonstrated by performing a reliability analysis of a conceptual PWR Containment Spray System. Certain numerical results obtained are compared with those of a prior fault tree analysis of the same system as documented in the 11 January 1973 draft report, A Fault Tree Evaluation of a PWR Spray System

  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. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  12. Safety Test Report for the PWR S/F Dry Storage System

    Energy Technology Data Exchange (ETDEWEB)

    Seo, K. S.; Lee, J. H.; Koo, K. H.; Lee, J. C.; Choi, W. S.; Bang, K. S.; Park, H. Y.; Jang, S. Y

    2008-10-15

    This is contract report conducted by KAERI under the contract with NETEC for safety test for the PWR S/F dry storage system. Leak Test was performed after drop test and turn-over test, the measured leakage rate was lower than allowable leakage rate. It is revealed that the containment integrity of the dry storage system is maintained. In the seismic test, the moving of the model was measured at SRTH seismic response of 0.4 g and 0.8 g. Therefore the seismic test results can be used fully to the test data for verification of the seismic analysis. In the thermal test, the direction of the inlet and outlet of the air has no effect on the heat transfer performance. The passive heat removal system of the horizontal storage module was designed well.

  13. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  14. Efficacious of estimatives of thermal-hydraulic conditions of the PWR core by measured parameters

    International Nuclear Information System (INIS)

    Camargo, C.T.M.; Pontedeiro, A.C.

    1985-01-01

    Using ALMOD 3W2 and COBRA IIIP computer codes an evaluation of usual methods of estimatives of heat transfer conditions in the PWR core was made, using variables of the monitored processes. It was done a parametric study in conditions of the permanent regim to verify the influence of variables such as, pressure, temperature and power in the value of critical heat flux. Parameters to prevent the DNB phenomenon in KWU power plants and Westinghouse were calculated and implemented in the ALMOD 3W2 program to estimate the DNBR evolution. It was identified a common origin to both methods and comparing with detailed calculations of the COBRA IIIP code, it was settled limitations in the application of parameters. (M.C.K.) [pt

  15. The role of sludge and fouling on local concentration processes in PWR steam generators

    International Nuclear Information System (INIS)

    Millett, D.J.; Paine, J.P.N.; Fenton, J.M.

    1992-01-01

    A detailed model of the transport processes in heated crevices is used to understand the role of corrosion product deposits on local concentration processes in PWR steam generators (SG). The model describes the heat, mass and momentum transfer processes which occur in the porous deposits found in tube support and tube sheet crevices and in the sludge pile on top of the tube sheet. The model is used to predict the concentration of a given specie in the liquid pore solution at steady-state, as a function of time, and may be readily expanded to multi-component solutions by the use of chemical thermodynamic models. In a previous paper the authors investigated the role of SG design and SG operating parameters on the concentration process. In this paper, several parametric studies were performed to investigate the sensitivity of the local concentration process to the corrosion product properties

  16. Steam generator for PWR type reactor

    International Nuclear Information System (INIS)

    Baba, Iwao; Hiyama, Nobuyuki.

    1994-01-01

    A steam generator of the present invention comprises a primary coolant chamber having primary coolants circulating therein, a secondary coolants chamber having secondary coolants and steams circulating therein, which are isolated from each other by a partition wall, and heat pipes disposed being passed through the partition wall. The heat pipes are disposed having an evaporation portion in the primary coolants chamber, a condensation portion in the secondary coolants chamber, and an intermediate heat insulating portion in the partition wall. Since the primary coolants containing radioactivity and the secondary coolants not containing radioactivity does not transfer heat directly by a heat transfer wall, a leakage accident of radioactivity to the secondary coolants can be prevented. Moreover, since the heat pipes are used, a great amount of heat can be transferred by a slight temperature difference by using steams of the heat transfer medium itself, latent heat due to coagulation, and capillary phenomenon. Since neither transferring power nor pumps are required, heat of the primary coolants can effectively be transferred to the secondary coolants. (N.H.)

  17. 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

  18. Effect of water chemistry on deposition for PWR plant operation

    International Nuclear Information System (INIS)

    Le Calvar, Marc; Bretelle, J. L.; Cailleaux, J. P.; Lacroix, R.; Guivarch, M.; Gay, N.; Taunier, S.; Gressier, F.; Varry, P.; Corredera, G.; Alos-Ramos, O.; Dijoux, M.

    2012-09-01

    For Pressurized Water Reactor (PWR) operation, water chemistry guidelines, specifications and associated surveillance programs are key to avoid deposition of oxides. Deposition of oxides can be detrimental by disrupting results of flow measurements, decreasing the thermal exchange capacity, or even by impairing safety. This paper describes the most important cases of deposition, their consequences for operation, and the implemented improvements to avoid their reoccurrence. Deposition that led to a Crud Induced Power Shift (CIPS) is also described. In the primary and in the secondary sides, orifice plates are typically used for measuring feedwater flow rate in nuclear power plants. Feedwater flow rates are used for control purposes and are important safety parameters as they are used to determine the plant's operating power level. Fouling of orifice plates in the primary side has been found during surveillance testing. For reactor coolant pumps, the formation of deposits on the seal No.1 can cause abnormally high or low leak rates through the seal. The leak rate through this seal must be carefully maintained within a prescribed range during plant operation. In the secondary side, orifice plate fouling has been the cause of feedwater flow/reference thermal power drift. For the steam generators (SG), magnetite deposition has led to fouling of the tube bundle, clogging of the quadri-foiled support plate holes and hard sludge formation on the base plate. For the generators, copper hollow conductors are widely used. Buildup of copper oxides on the interior walls of copper conductors has caused insufficient heat transfer. All these deposition cases have received adequate attention, understanding and response via improvement of our surveillance programs. (authors)

  19. Benefits of Low Boron Core Design Concept for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2009-10-15

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in {sup 10}B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts.

  20. Benefits of Low Boron Core Design Concept for PWR

    International Nuclear Information System (INIS)

    Daing, Aung Tharn; Kim, Myung Hyun

    2009-01-01

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in 10 B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts

  1. Prototypical fabrication of PWR spent fuel shipping cask

    International Nuclear Information System (INIS)

    Kwack, Eun Ho; Kim, Byung Ku; Kang, Hee Yung; Lee, Chung Young; Jeon, Kyeong Lak; Lee, Bum Soo

    1985-02-01

    This report describes about the safety analysis for the spent fuel shipping cask, which is used to transfer a single fuel assembly discharged from PWR in operation in Korea. The contents cover the methods and the results of structural, thermal, thermo-hydraulic, radiation shield and criticality detail analysis. The safety evaluation has been made under the normal transportation and hypothetical accident conditions such as 30ft free drop, puncture, fire, immersion, penetration, corner drop, etc,. Some corrections in design are made, and a brief information for fabrication and transportation are obtained by the use of a 1/6 scale model. The design is based on one year cooling time of the spent fuel with 40,000 MWT/MTU maximum burnup, which gives 7.2KW decay heat and 1.6x10 6 ci/hr radiation intensity. The cask is composed of main body with the double closures, impact limiter and fuel basket. The inner shell, inner closure and valves constitute the pressure boundary of the containment. The inner, intermediate and outer shells, upper and lower forgings are made of stainless steel which compose the main body with lead for gamma shield and 50% ethylene glycol for neutron shield. The impact limiters are made of balsa wood on both end sides of the cask to protect the cask from a sudden shocks in accident during the transportation. The analysis results show that the cask is proved to retain its structural integrity within allowable stress and to be safe under the normal and hypothetical accident conditions, and the maximum dose rates of radiation at 2m distance from the surface of the cask are less than the required values. The weight will be 23.2tons in dry and 27.8 tons in wet with fuel loaded. All the design data, calculated results for the structural integrity, shield and thermal analysis are shown in this report with the basic drawings. (Author)

  2. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  3. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  4. Aging effects in PWR power plants components

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  5. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    Okimura, Koji; Hori, Nobuyuki; Kanzaki, Hiroshi; Tokuhisa, Kiichi; Kamo, Kazuhiko; Kurokawa, Masaaki

    2007-01-01

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  6. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  7. MELCOR/VISOR PWR desktop simulator

    International Nuclear Information System (INIS)

    With, Anka de; Wakker, Pieter

    2010-01-01

    Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

  8. Summary of PWR leak detection studies

    International Nuclear Information System (INIS)

    Cho, J.H.; Elia, F.A. Jr.

    1986-01-01

    Thermal-hydraulic analysis can be used to determine the location and magnitude of leaks inside and location of leaks outside a pressurized water reactor (PWR) containment as required by plant technical specifications. The major advantage of this detection method is that it minimizes radiation exposure of maintenance personnel because most of the leak detection process is performed from the control room outside containment. Plant-specific analyses are utilized to predict change in parameters such as local dew point temperature, relative humidity, dry bulb temperature, and flow rate to sump for various leak rates and enthalpies. These parameter responses are then programmed into the plant computer and instrumentation is provided for area monitoring. The actual inputs are continuously monitored and compared to the predicted plant responses to identify the leak location and quantify the leak. This study concludes that a system that monitors dew point (or relative humidity) and dry bulb temperature changes together with the flow rate to the sump will provide the capability to both locate and quantify a leak inside a containment, while a system that monitors dew point temperature (or relative humidity) changes will provide the capability to locate a leak outside a containment

  9. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1984-01-01

    Hydrogen generation during a PWR LOCA has been estimated for design basis accident and for two more severe hypothetical accidents. Hydrogen production during design basis accident is a rather slow mechanism, allowing in the worst case, 15 days to connect a hydrogen recombining unit to the containment atmosphere monitoring system. Hydrogen generated by steam oxidation during more severe hypothetical accidents was found limited by steam availability and fuel melting phenomena. Uncertainty is, however, still remaining on corium-zirconium-steam interaction. In the worst case, calculations lead to the production of 500 kg of hydrogen, thus leading to a volume concentration of 15% in containment atmosphere, assuming homogeneous hydrogen distribution within the reactor building. This concentration is within flammability limits but not within detonation limits. However, hydrogen detonation due to local hydrogen accumulation cannot be discarded. A major uncertainty subsisting on hydrogen hazard is hydrogen distribution during the first hours of the accident. This point determines the effects and consequences of local detonation or deflagration which could possibly be harmful to safeguard systems, or induce missile generation in the reactor building. As electrical supply failures are identified as an important contributor to severe accident risk, corrective actions have been taken in France to improve their reliability, including the installation of a gas turbine on each site to supplement the existing sources. These actions are thus contributing to hydrogen hazard reduction

  10. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    Camous, F.; Chesnel, A.

    1989-12-01

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  11. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  12. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  13. Multicriteria analysis of public protection in PWR's

    International Nuclear Information System (INIS)

    Lombard, J.

    1986-09-01

    In order to manage a risk efficiently and to reach the ALARA level of protection, the best possible protection options must be employed. As the available resources are limited, it is not always possible to choose those options that minimize the risk, therefore a compromise must be made between risks and safety expenses. When the choice is difficult or complex, finding such a compromise can be facilitated by resorting to a decision aiding method which allows the assessment of the respective advantages of the various protection options considered. The multicriteria methods employ successive comparisons. Instead of searching for a final indicator expressing the performance of each option they compare all option pairs in order to determine if the gap between their respective advantages and disadvantages is sufficient to estimate that one option of the each pair is better than the other. Instead of judging each option globally these methods evaluate the advantages and disadvantages associated with the eventual choice of an option as compared with the others. These differential and comparative approach gives more flexibility and allows the introduction of qualitative criteria. The method presented here (Electre 3), one of the most recent ones, allows a multicriteria comparison of a set of options keeping into account the uncertainties associated with the options or the preferences. In order to illustrate this method a simple example (4 options, 4 criteria) dealing with a PWR liquid releases treatment system, is taken up

  14. Modeling of PWR fuel at extended burnup

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael M.; Silva, Antonio Teixeira, E-mail: rmdias@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  15. Operation and maintenance in Genkai PWR Plant

    International Nuclear Information System (INIS)

    Ohta, Shojiro

    1984-01-01

    The No.1 PWR plant with 559 MW capacity in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc., required about 115 days for the regular inspection in fiscal 1982 and thereafter, although more maintenance work was done. But No.2 plant of the same type required not more than 80 days. In most cases, the period of one operation cycle was from 10 to 12 months, but in the third operation cycle of No.2 plant, it is expected to be 13 months. The capacity ratio of the whole power station was 75.2% at the end of fiscal 1983. These operational records all exceeded the Japanese average. The plants are two-loop Westinghouse type PWRs, and No.1 plant started the commercial operation of anti h and the increment of P 0 + . (author) apacity ratio of No.1 plant was 71.6%, and that of No.2 plant was 85.5%. The intergranular attack on steam generator tubes was found first in the fifth regular inspection, and also in the sixth and seventh inspections, and the faulty tubes were plugged. The prevention of its spread is the largest problem. The in-service quality assurance activity, the personnel training program and the effort of upgrading the plant availability are reported. (Kako, I.)

  16. Computer aided information system for a PWR

    International Nuclear Information System (INIS)

    Vaidian, T.A.; Karmakar, G.; Rajagopal, R.; Shankar, V.; Patil, R.K.

    1994-01-01

    The computer aided information system (CAIS) is designed with a view to improve the performance of the operator. CAIS assists the plant operator in an advisory and support role, thereby reducing the workload level and potential human errors. The CAIS as explained here has been designed for a PWR type KLT- 40 used in Floating Nuclear Power Stations (FNPS). However the underlying philosophy evolved in designing the CAIS can be suitably adopted for other type of nuclear power plants too (BWR, PHWR). Operator information is divided into three broad categories: a) continuously available information b) automatically available information and c) on demand information. Two in number touch screens are provided on the main control panel. One is earmarked for continuously available information and the other is dedicated for automatically available information. Both the screens can be used at the operator's discretion for on-demand information. Automatically available information screen overrides the on-demand information screens. In addition to the above, CAIS has the features of event sequence recording, disturbance recording and information documentation. CAIS design ensures that the operator is not overburdened with excess and unnecessary information, but at the same time adequate and well formatted information is available. (author). 5 refs., 4 figs

  17. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  18. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael M.; Silva, Antonio Teixeira

    2015-01-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  19. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

  20. Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

    International Nuclear Information System (INIS)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol

    2015-01-01

    Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of

  1. 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

  2. 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.)

  3. Nuclear district heating

    International Nuclear Information System (INIS)

    Ricateau, P.

    1976-01-01

    An economic study of nuclear district heating is concerned with: heat production, its transmission towards the area to be served and the distribution management towards the consumers. Foreign and French assessments show that the high cost of now existing techniques of hot water transport defines the competing limit distance between the site and township to be below some fifty kilometers for the most important townships (provided that the fuel price remain stationary). All studies converge towards the choice of a high transport temperature as soon as the distance is of some twenty kilometers. As for fossile energy saving, some new possibilities appear with process heat reactors; either PWR of about 1000MWth for large townships, or pool-type reactors of about 100MWth when a combination with an industrial steam supply occurs [fr

  4. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

  5. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  6. PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)

    International Nuclear Information System (INIS)

    1981-01-01

    1 - Description of test facility: PKL-facility simulates the essential primary system components of a typical West German 1300 PWR with regard to their thermohydraulic behaviour. The facility essentially consists of the pressure vessel with the heated bundle, the downcomer simulator, the primary loops with the components steam generator and pump simulator, the injection devices, the break geometry simulator, as well as the separators connected thereto, and the test containment to maintain a back-pressure at the location of break which is expected to be typical for emergency conditions. The number of heater rods and the cross-sections of the testing plant are on a reduced scale 1:134 in comparison with a typical German PWR. The elevations and locations are essentially full scale. Pressure vessel: The space between the pressure vessel and the inner core casing is sealed from the core region and the upper and lower plenum and connected with the upper plenum only by a pressure equalization line. The rod bundle surrounded by the inner core casing consists of 340 rods, 337 of which are indirect electrically heated. The test bundle cross-section as well as a heater element with the measuring elevations, the original-KWU-spacers and the axial power profile (7 power steps) are described. Downcomer: The downcomer is simulated by the downcomer nozzle region and the downcomer U-tube. The cold leg injection takes place both directly in the downcomer nozzle region and in the lines of t he intact single and double loop near to the downcomer nozzle region. A cylindrical insertion and repulsing metal sheets are installed in the downcomer nozzle region in order to avoid the emergency injection points into the broken loop. 2 - Description of test: Test K 9 out of a series PKL-IB was conducted on May 30, 1979 by Kraftwerk Union (KWU) at Erlangen (Germany). The objective of the integral cold leg injection test K 9 (double-ended 200%-break) was to investigate after a LOCA the refill and

  7. UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

    International Nuclear Information System (INIS)

    1998-01-01

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

  8. Statistical analysis of the early phase of SBO accident for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Yaroslav, E-mail: y.kozmenkov@hzdr.de; Jobst, Matthias, E-mail: m.jobst@hzdr.de; Kliem, Soeren, E-mail: s.kliem@hzdr.de; Schaefer, Frank, E-mail: f.schaefer@hzdr.de; Wilhelm, Polina, E-mail: p.wilhelm@hzdr.de

    2017-04-01

    Highlights: • Best estimate model of generic German PWR is used in ATHLET-CD simulations. • Uncertainty and sensitivity analysis of the early phase of SBO accident is presented. • Prediction intervals for occurrence of main events are evaluated. - Abstract: A statistical approach is used to analyse the early phase of station blackout accident for generic German PWR with the best estimate system code ATHLET-CD as a computation tool. The analysis is mainly focused on the timescale uncertainties of the accident events which can be detected at the plant. The developed input deck allows variations of all input uncertainty parameters relevant to the case. The list of identified and quantified input uncertainties includes 30 parameters related to the simulated physical phenomena/processes. Time uncertainties of main events as well as the major contributors to these uncertainties are defined. The uncertainty in decay heat has the highest contribution to the uncertainties of the analysed events. A linear regression analysis is used for predicting times of future events from detected times of occurred/past events. An accuracy of event predictions is estimated and verified. The presented statistical approach could be helpful for assessing and improving existing or elaborating additional emergency operating procedures aimed to prevent severe damage of reactor core.

  9. Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

    International Nuclear Information System (INIS)

    Todorova, N.K.; Ivanov, K.N.

    2003-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement

  10. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  11. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  12. Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

    Science.gov (United States)

    Ga¨rtner, M.; Fischer, G.

    1987-06-01

    To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

  13. Nupec thermal hydraulic test to evaluate post-DNB characteristics for PWR fuel assemblies (1. general test plan and results)

    International Nuclear Information System (INIS)

    Norio, Kono; Kenji, Murai; Kaichiro, Misima; Takayuki, Suemura; Yoshiei, Akiyama; Keiichi, Hori

    2001-01-01

    In the present thermal hydraulic design of Pressurized Water Reactor (PWR), a departure from nucleate boiling (DNB) under anticipated transient conditions is not allowed. However, it is recognized that the DNB dose not cause a fuel rod failure immediately, and a suitable reactor trip can prevent the core from severe damages. If the fuel rod temperature under the post-DNB conditions can be accurately evaluated, the potentially existing margin in the present design method will be quantitatively assessed. To establish the heat transfer evaluation method on post-DNB event for PWR thermal hydraulic design, Nuclear Power Engineering Corporation (NUPEC) started a program, NUPEC Thermal Hydraulic Test to Evaluate Post-DNB Characteristics for PWR Fuel Assemblies (NUPEC-TH-P), in 1995 (hereinafter the year means fiscal year) under the sponsorship of Ministry of Economy, Trade and industry (METI). This program is now under going until 2001. This paper is to show the overall plan and the status of NUPEC-TH-P. (authors)

  14. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom`s N4 plant, KWU`s plants, ABB-CE`s Systems 80+, and Westinghouse`s AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author).

  15. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

  16. Modeling on a PWR power conversion system with system program

    International Nuclear Information System (INIS)

    Gao Rui; Yang Yanhua; Lin Meng

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Daya Bay Power Station, this paper models the thermal-hydraulic systems of primary and secondary loops for PWR by using the PWR best-estimate program-RELAP5. To simulate the full-scope power conversion system, not only the traditional basic system models of nuclear island, but also the major system models of conventional island are all considered and modeled. A comparison between the calculated results and the actual data of reactor demonstrates a fine match for Daya Bay Nuclear Power Station, and manifests the feasibility in simulating full-scope power conversion system of PWR by RELAP5 at the same time. (authors)

  17. Improved emergency elevated air release for simplified PWR

    International Nuclear Information System (INIS)

    Naitoh, T.; Bruce, R.A.; Hirota, K.; Tajiri, Y.

    1992-01-01

    In developing the application of the simplified PWR in Japan, one of the most important areas is to limit post-accident site boundary whole body dose. In addressing this, the concept of Emergency Passive Air Filtration System (EPAFS) and it's feasibility is developed. The efficiency of charcoal filtering and the atmospheric diffusion effect of an elevated air release are important for dose reduction. The performance of these functions was evaluated by confirmatory testing. The test results confirmed a 99 percent efficiency of charcoal filter and an atmospheric diffusion effect higher than that of a conventional plant. The Emergency Passive Air Filtration System (EPAFS) and the atmospheric diffusion effect of elevated air release contribute to making the calculated post-accident site boundary whole body dose of simplified PWR as low as that of the conventional Japanese PWR plant. (author)

  18. Swing-Down of 21-PWR Waste Package

    International Nuclear Information System (INIS)

    A.K. Scheider

    2001-01-01

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

  19. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  20. The advanced main control console for next japanese PWR plants

    International Nuclear Information System (INIS)

    Tsuchiya, A.; Ito, K.; Yokoyama, M.

    2001-01-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  1. Cylindrization of a PWR core for neutronic calculations

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2005-01-01

    In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

  2. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Kondo, Y.

    1987-01-01

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

  3. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    Fairclough, M.P.; Tymons, B.J.

    1985-06-01

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  4. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2007-07-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  5. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Junho; Hong, Jong-Dae; Seo, Myung-Gyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life.

  6. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    International Nuclear Information System (INIS)

    Lee, Junho; Hong, Jong-Dae; Seo, Myung-Gyu; Jang, Changheui

    2015-01-01

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life

  7. The traveller: a new look for PWR fresh fuel packages

    International Nuclear Information System (INIS)

    Bayley, B.; Stilwell, W.E.; Kent, N.A.

    2004-01-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

  8. The study on radioactivity reduction of spent PWR cladding hull

    International Nuclear Information System (INIS)

    Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

    2003-01-01

    Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

  9. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  10. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  11. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  12. Sensitivity of risk parameters to human errors for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.; Hall, R.E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study

  13. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

    2004-01-01

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  14. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    Hoffman, B.; Tsuzuki, S.

    2002-01-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  15. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  16. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  17. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  18. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

  19. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  20. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt