WorldWideScience

Sample records for active subslab depressurization

  1. MODELING THE INFLUENCE OF ACTIVE SUBSLAB DEPRESSURIZATION (ASD) SYSTEMS ON AIRFLOWS IN SUBSLAB AGGREGATE BEDS

    Science.gov (United States)

    A simple model is presented that allows the pressure difference in a subslab aggregate layer to be estimated as a function of radial distance from the central suction point of an active subslab depressurization system by knowing the average size, thickness, porosity, and permeabi...

  2. Radon remediation of a two-storey UK dwelling by active sub-slab depressurization: observations on hourly Radon concentration variations

    International Nuclear Information System (INIS)

    Denman, A.R.

    2008-01-01

    Radon concentration levels in a two-storey detached single-family dwelling in Northamptonshire, UK, were monitored at hourly intervals throughout a 5-week period during which sub-slab depressurization remediation measures, including an active sump system, were installed. Remediation of the property was accomplished successfully, with the mean radon levels upstairs and downstairs greatly reduced and the prominent diurnal variability in radon levels present prior to remediation almost completely removed. Following remediation, upstairs and downstairs radon concentrations were 32% and 16% of their pre-remediation values respectively. The mean downstairs radon concentration was lower than that upstairs, with pre-and post-remediation values of the upstairs/downstairs concentration ratio, R U/D , of 0.93 and 1.76 respectively. Cross-correlation between upstairs and downstairs radon concentration time-series indicates a time-lag of the order of 1 hour or less, suggesting that diffusion of soil-derived radon from downstairs to upstairs either occurs within that time frame or forms a relatively insignificant contribution to the upstairs radon level. Cross-correlation between radon concentration time-series and the corresponding time-series for local atmospheric parameters demonstrated correlation between radon concentrations and internal/external pressure-difference prior to remediation. This correlation disappears following remediation, confirming the effectiveness of the remediation procedure in mitigating radon ingress from the ground via the stack-effect. Overall, these observations provide further evidence that radon emanation from building materials makes a not insignificant contribution to radon concentration levels within the building. Furthermore, since this component remains essentially unaffected by sub-slab depressurization, its proportional contribution to the total radon levels in the home increases following remediation, leading to the conclusion that where

  3. DESIGN AND TESTING OF SUB-SLAB DEPRESSURIZATION FOR RADON MITIGATION IN NORTH FLORIDA HOUSES - PART I. PERFORMANCE AND DURABILITY - VOLUME 1. TECHNICAL REPORT

    Science.gov (United States)

    The report gives results of a demonstration/research project to evaluate sub-slab depressurization (SSD) techniques for radon mitigation in North Florida where the housing stock is primarily slab-on-grade and the sub-slab medium typically consists of native soil and sand. Objecti...

  4. DESIGN AND TESTING OF SUB-SLAB DEPRESSURIZATION FOR RADON MITIGATION IN NORTH FLORIDA HOUSES - PART I. PERFORMANCE AND DURABILITY - VOLUME 2. DATA APPENDICES

    Science.gov (United States)

    The report gives results of a demonstration/research project to evaluate sub-slab depressurization (SSD) techniques for radon mitigation in North Florida where the housing stock is primarily slab-on-grade and the sub-slab medium typically consists of native soil and sand. Objecti...

  5. Energy penalties associated with the use of a sub-slab depressurization system

    International Nuclear Information System (INIS)

    Clarkin, M.; Brennan, T.; Osborne, M.C.

    1990-01-01

    One of the primary radon mitigation techniques used to reduce indoor radon concentrations in houses is a sub-slab depressurization system. In this type of system, a fan removes soil gases containing radon from beneath the floor slab and exhausts the gases to the outdoors by creating a pressure field beneath the slab that is negative relative to the basement air pressure. Because of this negative pressure, indoor conditioned air can be drawn through the floor penetrations and exhausted outdoors. In order to determine the amount of conditioned air that is being lost, a series of experiments utilizing tracer gases were performed in three houses. This paper presents the results of these experiments

  6. Cost analysis of soil-depressurization techniques for indoor radon reduction

    International Nuclear Information System (INIS)

    Henschel, D.B.

    1991-01-01

    The article discusses a parametric cost analysis to evaluate active soil depressurization (ASD) systems for indoor radon reduction in houses. The analysis determined the relative importance of 14 ASD design variables and 2 operating variables on the installation and operating costs of residential ASD systems in several types of houses. Knowledge of the most important variables would enable EPA's research and development efforts to be more effectively directed at ways to reduce ASD costs and thus to increase utilization of the technology. Parameters offering the greatest potential for reductions in installation costs included three dealing with houses with poor subslab communication: (1) reducing the number of subslab depressurization pipes; (2) eliminating excavation of large subslab pits beneath the suction pipes to improve suction field extension; and (3) improving the effectiveness of premitigation subslab communication diagnostic testing in achieving simpler, less expensive ASD system designs. In addition, determining acceptable conditions for discharging ASD exhaust at grade level would reduce installation costs. Better design guidance for crawl-space submembrane depressurization (SMD) systems could reduce installation costs, if difficult membrane sealing steps and complete coverage of the crawl-space floor by the membrane can be avoided

  7. A study of the influence of a gravel subslab layer on radon entry rate using two basement structures

    International Nuclear Information System (INIS)

    Robinson, A.L.; Sextro, R.G.; Fisk, W.J.; Garbesi, K.; Wooley, J.; Wollenberg, H.A.

    1993-01-01

    In buildings with elevated radon concentrations, the dominant transport mechanism of radon is advective flow of soil gas into the building substructure. However, the building-soil system is often complex, making detailed studies of the radon source term difficult. In order to examine radon entry into buildings, the authors have constructed two room-size, precisely-fabricated basement structures at a site with relatively homogeneous, moderately permeable soil. The basements are identical except that one lies directly on native soil whereas the other lies on a high permeability aggregate layer. The soil pressure field and radon entry rate have been measured for different basement pressures and environmental conditions. The subslab gravel layer greatly enhances the advective entry of radon into the structure; when the structures are depressurized, the radon entry rate into the structure with the subslab gravel layer is more than a factor of 3 times the radon entry rate into the other structure for the same depressurization. The gravel subslab layer also spreads the pressure field around the structure, extending the field of influence of the structure and the region from which it draws radon

  8. A study of the influence of a gravel subslab layer on radon entry rate using two basement structures

    International Nuclear Information System (INIS)

    Robinson, A.L.; Sextro, R.G.; Fisk, W.J.; Garbesi, K.; Wooley, J.; Wollenberg, H.A.

    1993-01-01

    In buildings with elevated radon concentrations, the dominant transport mechanism of radon is advective flow of soil gas into the building substructure. However, the building-soil system is often complex, making detailed studies of the radon source term difficult. In order-to examine radon entry into buildings, we have constructed two room-size, precisely-fabricated basement structures at a site with relatively homogeneous, moderately permeable soil. The basements are identical except that one lies directly on native soil whereas the other lies on a high permeability aggregate layer. The soil pressure field and radon entry rate have been measured for different basement pressures and environmental conditions. The subslab gravel layer greatly enhances the advective entry of radon into the structure; when the structures are depressurized, the radon entry rate into the structure with the subslab gravel layer is more than a factor of 3 times the radon entry rate into the other structure for the same depressurization. The gravel subslab layer also spreads the pressure field around the structure, extending the field of influence of the structure and the region from which it draws radon. (orig.). (7 refs., 3 figs.)

  9. The complex modelling of various effects of the sub-slab ventilation systems

    International Nuclear Information System (INIS)

    Svoboda, Z.

    2004-01-01

    Sub-slab ventilation systems and, in particular, sub-slab depressurization (SSD) systems are among the most efficient radon protective and remedial measures. Numerical modelling can serve as a very powerful tool in the design stage of such systems. The calculations include estimation of the pressure field in the ground under the house with an SSD system and estimation of the radon concentration field. The SSD system also affects the temperature and relative humidity distribution, and therefore those fields should be calculated as well. All the analyses can be carried out applying the simplification of a non-transient steady-state behavior. The numerical solution can be obtained by using the finite difference method or the finite element method. The results of numerical calculation comprise the air pressure field under the building with SSD system, radon concentration field, and temperature and relative humidity fields. The reliability of the numerical models has been verified on six houses with different SSD systems. The results obtained from one house are presented to demonstrate the complete process of verification. The remedial action consisted in the installation of an SSD system in combination with rebuilding of the floors. Soil air temperature, relative humidity, pressure difference and soil air radon concentration were measured continuously. All measurements were carried out for the two modes, i.e. with the SSD system operational or disabled. The first numerical analysis was the calculation of the three-dimensional air pressure field in the whole sub-slab space of the experimental house. The correlation between the calculated and observed values was very good (agreement better than 10%). The calculation of the two-dimensional steady-state temperature and relative humidity field also exhibited a good agreement with the observed values, with differences below 15%. The two-dimensional steady-state field of radon concentrations in the soil under the experimental

  10. ACTIVE SOIL DEPRESSURIZATION (ASD) DEMONSTRATION IN A LARGE BUILDING

    Science.gov (United States)

    The report gives results of an evaluation of the feasibility of implementing radon resistant construction techniques -- especially active soil depressurization (ASD) -- in new large buildings in Florida. Indoor radon concentrations and radon entry were monitored in a finished bui...

  11. SUMMARY OF EPA'S RADON REDUCTION RESEARCH IN SCHOOLS DURING 1989-90

    Science.gov (United States)

    The report details radon mitigation research in schools conducted by EPA during 1989 and part of 1990. The major objective was to evaluate the potential of active subslab depressurization (ASD) in various geologic and climatic regions. The different geographic regions also pres...

  12. Testing of indoor radon-reduction techniques in basement houses having adjoining wings. Final report, August 1988-September 1989

    International Nuclear Information System (INIS)

    Messing, M.

    1990-11-01

    The report gives results of tests of indoor radon reduction techniques in 12 existing Maryland houses, with the objective of determining when basement houses with adjoining wings require active soil depressurization (ASD) treatment of both wings, and when treatment of the basement alone is sufficient. In five basement houses with adjoining slabs on grade, ASD treatment of both wings provided an incremental additional radon reduction of 0 to 5.2 pCi/L, compared to ASD treatment of either one of the slabs alone. However, basement-only treatment reduced radon to <4 pCi/L in all five houses. In six basement houses having adjoining crawl spaces, ASD treatment of both wings (including sub-liner depressurization of the crawl space) provided little additional reduction compared to basement-only treatment, when sub-slab communication was good. When communication was not good, treatment of both wings was required to achieve <4 pCi/L. Tests of one fully slab-on-grade house showed that, when there is good aggregate under the slab, a one-pipe sub-slab depressurization system can achieve <1-2 pCi/L, even when there are forced-air supply ducts under the slab

  13. Radon Sub-slab Suctioning System Integrated in Insulating Layer

    DEFF Research Database (Denmark)

    Rasmussen, Torben Valdbjørn

    2013-01-01

    This poster presents a new radon sub-slab suctioning system. This system makes use of a grid of horizontal pressurised air ducts located within the lower part of the rigid insulation layer of the ground floor slab. For this purpose a new system of prefabricated lightweight elements is introduced...

  14. HTGR depressurization analysis

    International Nuclear Information System (INIS)

    Boccio, J.L.; Colman, J.; Skalyo, J.; Beerman, J.

    1979-01-01

    Relaxation of the prima facie assumption of complete mixing of primary and secondary containment gases during HTGR depressurization has led to a study program designed to identify and selectively quantify the relevant gas dynamic processes which prevail during the depressurization event. Uncertainty in the degree of gas mixedness naturally leads to uncertainty in containment vessel design pressure and heat loads and possible combustion hazards therein. This paper succinctly details an analytical approach and modeling methodology of the exhaust jet structure/containment vessel interaction during penetration failures. (author)

  15. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

    2015-01-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  16. Passive depressurization accident management strategy for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

    2015-04-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

  17. Sediment–well interaction during depressurization

    KAUST Repository

    Shin, Hosung

    2016-10-05

    Depressurization gives rise to complex sediment–well interactions that may cause the failure of wells. The situation is aggravated when high depressurization is imposed on sediments subjected to an initially low effective stress, such as in gas production from hydrate accumulations in marine sediments. Sediment–well interaction is examined using a nonlinear finite element simulator. The hydro-mechanically coupled model represents the sediment as a Cam-Clay material, uses a continuous function to capture compressibility from low to high effective stress, and recognizes the dependency of hydraulic conductivity on void ratio. Results highlight the critical effect of hydro-mechanical coupling as compared to constant permeability models: A compact sediment shell develops against the screen, the depressurization zone is significantly smaller than the volume anticipated assuming constant permeability, settlement decreases, and the axial load on the well decreases; in the case of hydrates, gas production will be a small fraction of the mass estimated using a constant permeability model. High compressive axial forces develop in the casing within the production horizon, and the peak force can exceed the yield capacity of the casing and cause its collapse. Also tensile axial forces may develop in the casing above the production horizon as the sediment compacts in the depressurized zone and pulls down from the well. Well engineering should consider: slip joints to accommodate extensional displacement above the production zone, soft telescopic/sliding screen design to minimize the buildup of compressive axial force within the production horizon, and enlarged gravel pack to extend the size of the depressurized zone.

  18. Ground Source Heat Pump Sub-Slab Heat Exchange Loop Performance in a Cold Climate

    Energy Technology Data Exchange (ETDEWEB)

    Mittereder, N.; Poerschke, A.

    2013-11-01

    This report presents a cold-climate project that examines an alternative approach to ground source heat pump (GSHP) ground loop design. The innovative ground loop design is an attempt to reduce the installed cost of the ground loop heat exchange portion of the system by containing the entire ground loop within the excavated location beneath the basement slab. Prior to the installation and operation of the sub-slab heat exchanger, energy modeling using TRNSYS software and concurrent design efforts were performed to determine the size and orientation of the system. One key parameter in the design is the installation of the GSHP in a low-load home, which considerably reduces the needed capacity of the ground loop heat exchanger. This report analyzes data from two cooling seasons and one heating season. Upon completion of the monitoring phase, measurements revealed that the initial TRNSYS simulated horizontal sub-slab ground loop heat exchanger fluid temperatures and heat transfer rates differed from the measured values. To determine the cause of this discrepancy, an updated model was developed utilizing a new TRNSYS subroutine for simulating sub-slab heat exchangers. Measurements of fluid temperature, soil temperature, and heat transfer were used to validate the updated model.

  19. Summary of EPA's radon-reduction research in schools during 1989-90

    International Nuclear Information System (INIS)

    Leovic, K.W.

    1990-10-01

    The report details radon mitigation research in schools conducted by EPA during 1989 and part of 1990. The major objective was to evaluate the potential of active subslab depressurization (ASD) in various geologic and climatic regions. The different geographic regions also presented a variety of construction types and heating, ventilating, and air-conditioning (HVAC) system designs that are encountered in radon mitigation of school buildings. A secondary objective was to initiate research in difficult-to-mitigate schools. The research led to the following major conclusions on radon diagnostics and mitigation in schools: (1) Schools have many physical characteristics that typically make their mitigation more complex than house mitigation, including building size and substructure, subslab barriers, HVAC systems, and locations of utility lines. (2) Important school diagnostic procedures and measurements include review of radon measurements and building plans, investigation of the building to assess potential radon entry routes and confirm information in the building plans, analysis of the HVAC system and its influence on pressure differentials and radon levels, and subslab pressure field extension measurements to determine the potential applicability of ASD. (3) ASD can be applied successfully in schools where subslab communication barriers are limited

  20. The radontechnical process of the sub-slab suction system in detached, semi-detached and terraced houses; Imupistejaerjestelmaen radontekninen toiminta eri tyyppisissae pien- ja rivitaloissa

    Energy Technology Data Exchange (ETDEWEB)

    Kettunen, A.V.; Rissanen, R.; Viljanen, M. [Helsinki Univ. of Technology, Otaniemi (Finland). Lab. of Structural Engineering and Building Physics; Arvela, H. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland)

    1997-12-31

    The sub-slab suction system is a repair method for decreasing radon transfer from soil to indoor air in houses. The sub-slab suction system changes the flow field with the aim to reduce the radon content under the ground slab or/and changes the flow direction from indoor air to foundation soil. The study is divided into the theoretical and experimental part. The theoretical part concentrates on solving the factors affecting the sub-slab suction system taking into consideration features of old houses. The experimental part uses the results of the theoretical part when the level and the type of the suction point had been chosen. The effect of the sub-slab suction system was tested with pressure difference experiments and the radon content measurements in 26 different, typical Finnish detached, semi-detached and terrace houses.

  1. Risk assessment for the intentional depressurization strategy in PWRs

    International Nuclear Information System (INIS)

    Dingman, S.E.

    1994-03-01

    An accident management strategy has been proposed in which the reactor coolant system is intentionally depressurized during an accident. The aim is to reduce the containment pressurization that would result from high pressure ejection of molten debris at vessel breach. Probabilistic risk assessment (PRA) methods were used to evaluate this strategy for the Surry nuclear power plant. Sensitivity studies were conducted using event trees that were developed for the NUREG-1150 study. It was found that depressurization (intentional or unintentional) had minimal impact on the containment failure probability at vessel breach for Surry because the containment loads assessed for NUREG-1150 were not a great threat to the containment survivability. An updated evaluation of the impact of intentional depressurization on the probability of having a high pressure melt ejection was then made that reflected analyses that have been performed since NUREG-1150 was completed. The updated evaluation confirmed the sensitivity study conclusions that intentional depressurization has minimal impact on the probability of a high pressure melt ejection. The updated evaluation did show a slight benefit from depressurization because depressurization delayed core melting, which led to a higher probability of recovering emergency core coolant injection, thereby arresting the core damage

  2. Study on primary coolant system depressurization effect factor in pressurized water reactor

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

    The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

  3. The Electret Radon Sniffer as a diagnostic tool for optimizing the efficiency of sub-slab ventilation systems and/or sealing

    International Nuclear Information System (INIS)

    Yates, J.H.; Kotrappa, P.

    1990-01-01

    In this paper a method for the measurement and interpretation of radon levels beneath a poured concrete slab floor is presented, along with application strategies intended to guide the mitigator in his efforts to effect an optimally efficient sealing or sub-slab mitigation system installation in the structure so measured. Case studies are presented for illustrative purposes, and cautions and uncertainties in this method are detailed, along with related areas of inquiry. The presumed greatest utility of this method is in attacking radon in structures evidencing poor sub-slab communication

  4. Basement depressurization using dwelling mechanical exhaust ventilation system

    International Nuclear Information System (INIS)

    Collignan, B.; O'Kelly, P.; Pilch, E.

    2004-01-01

    The mechanical ventilation exhaust system is commonly used in France to generate air renewal into building and especially into dwelling. It consists of a permanent mechanical air extraction from technical rooms (kitchen, bathrooms and toilets) using a unique fan connected to exhaust ducts. Natural air inlets in living room and bed rooms ensure an air flow from living spaces towards technical rooms. To fight against radon into building, the most recognised efficient technique is the Soil Depressurization System (S.D.S.) consisting in depressurizing the house basement. The aim of this study is to test the ability of the dwelling mechanical ventilation system to depressurize the basement in conjunction with air renewal of a house. For that purpose, a S.D.S. has been installed in an experimental house at CSTB during its construction. At first, tests undertaken with a variable velocity fan connected to the S.D.S. have characterised the permeability of the basement. It is shown that basement can be depressurized adequately with a relatively low air flow rate. At a second stage, S.D.S. has been connected to the exhaust ventilation fan used for the mechanical ventilation of the house. Results obtained show the ability of such ventilation system to generate sufficient depressurization in the basement and to ensure simultaneously adequate air change rate in the dwelling. (author)

  5. Depressurization test on hot gas duct

    International Nuclear Information System (INIS)

    Tanihira, Masanori; Kunitomi; Kazuhiko; Inagaki, Yoshiyuki; Miyamoto, Yoshiaki; Sato, Yutaka.

    1989-05-01

    To study the integrity of internal structures and the characteristics in a hot gas duct under the rapid depressurization accident, depressurization tests have been carried out using a test apparatus installed the hot gas duct with the same size and the same structures as that of the High Temperature Engineering Test Reactor (HTTR). The tests have been performed with three parameters: depressurization rate (0.14-3.08 MPa/s) determined by orifice diameter, area of the open space at the slide joint (11.9-2036 mm 2 ), and initial pressure (1.0-4.0 MPa) filled up in a pressure vessel, by using nitrogen gas and helium gas. The maximum pressure difference applied on the internal structures of the hot gas duct was 2.69 MPa on the liner tube and 0.45 MPa on the separating plate. After all tests were completed, the hot gas duct which was used in the tests was disassembled. Inspection revealed that there were no failure and no deformation on the internal structures such as separating plates, insulation layers, a liner tube and a pressure tube. (author)

  6. Effect of natural circulation on RCS depressurization strategy in PWR NPP

    International Nuclear Information System (INIS)

    Zhang Kun; Tong Lili; Cao Xuewu

    2009-01-01

    The natural circulation model of Chinese Qinshan Nuclear Power Plant (NPP) Unit 2 is built using SCDAP/RELAP5 code. Selecting TMLB' accident as the base sequence, this paper analyzes the natural circulation phenomena in high-pressure core melt severe accident. In order to study the effect of natural circulation on RCS depressurization strategy, the accident progressions of RCS depressurization with and without natural circulation are simulated, respectively. According to the results, the natural circulation can delay the initiation of RCS depressurization and the whole accident progression, but it does not evidently influence the results of RCS depressurization. (authors)

  7. Reducing indoor radon concentrations by passive subslab ventilation

    International Nuclear Information System (INIS)

    Jiranek, M.

    2005-01-01

    The primary objective of our study was to establish whether passive soil ventilation systems installed under existing houses have an effect on indoor radon concentrations. Experiments were conducted in two single-family houses. The soil ventilation under each house consists of the network of flexible perforated pipes laid into the layer of coarse gravel of the minimal thickness 150 mm. Soil air from the perforated pipes is ventilated by means of the vertical exhaust pipe that runs through the heated part of the house and ends above the roof of the house. At the top of the vertical exhaust a wind turbine is mounted in order to improve the stack effect during the windy weather .In addition to the soil ventilation both houses were provided with new floors composed of concrete slab and radon proof insulation made of LDPE membrane. The efficiency of passive soil ventilation systems varies within the year in dependence on the temperature gradient and wind speed. Preliminary results indicate that temperature gradient predominates. However the maximum under-pressure at the base of the vertical exhaust pipe caused by temperature differences is not so high. During one-year observation period the maximum temperature related under-pressure was only -8 Pa. The wind effect starts to be noticeable for speeds higher than 5 m/s and more apparent becomes for speeds above 10 m/s. The maximum values of under-pressure due to wind forces were measured within the range - 20 Pa and -30 Pa for wind speeds from 20 m/s to 25 m/s. Quite significant variations of the subslab under-pressure within one day were observed. The maximum under pressure was measured at late night or early morning when the outdoor temperature was the lowest. Annual variations were also confirmed. During the winter the temperature gradient is higher than in the summer time and thus the subslab under-pressure is consistently higher in the winter. Preliminary results indicate that passive soil ventilation systems with

  8. Radon mitigation in schools

    International Nuclear Information System (INIS)

    Leovic, K.W.; Craig, A.B.; Saum, D.W.

    1990-01-01

    This article reports on radon mitigation in school buildings. Subslab depressurization (SSD) has been the most successful and widely used radon reduction method in houses. Thus far, it has also substantially reduced radon levels in a number of schools. Schools often have interior footings or thickened slabs that may create barriers for subslab air flow if a SSD system is the mitigation option. Review of foundation plans and subslab air flow testing will help to determine the presence and effect of such barriers. HVAC systems in schools vary considerable and tend to have a greater influence on pressure differentials (and consequently radon levels) than do heating and air-conditioning systems encountered in the radon mitigation of houses. As part of any radon mitigation method, ASHRAE Standard 62-1989 should be consulted to determine if the installed HVAC system is designed and operated to achieve minimum ventilation standards for indoor air quality

  9. Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

    International Nuclear Information System (INIS)

    Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

    1997-01-01

    The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

  10. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    International Nuclear Information System (INIS)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A.

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of ∼922 K (1200 degree F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs

  11. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

  12. Radon mitigation experience in difficult-to-mitigate schools

    International Nuclear Information System (INIS)

    Leovic, K.W.; Craig, A.B.

    1990-01-01

    Initial radon mitigation experience in schools has shown sub-slab depressurization (SSD) to be generally effective in reducing elevated levels of radon in schools that have a continuous layer of clean, coarse aggregate underneath the slab. However, mitigation experience is limited in schools without sub-slab aggregate and in schools with characteristics such as return-air ductwork underneath the slab or unducted return-air plenums in the drop ceiling that are open to the sub-slab area (via open tops of block walls). Mitigation of schools with utility tunnels and of schools constructed over crawl spaces is also limited. Three Maryland schools exhibiting some of the above characteristics are being researched to help understand the mechanisms that control radon entry and mitigation in schools where standard SSD systems are not effective. This paper discusses specific characteristics of potentially difficult-to-mitigate schools and, where applicable, details examples from the three Maryland schools

  13. Characteristics and design improvement of AP1000 automatic depressurization system

    International Nuclear Information System (INIS)

    Jin Fei

    2012-01-01

    Automatic depressurization system, as a specialty of AP1000 Design, enhances capability of mitigating design basis accidents for plant. Advancement of the system is discussed by comparing with traditional PWR design and analyzing system functions, such as depressurizing and venting. System design improvement during China Project performance is also described. At the end, suggestions for the system in China Project are listed. (author)

  14. Depressurization study of CAREM 25 reactor considering the structures heat transfer

    International Nuclear Information System (INIS)

    Doval, A.

    1990-01-01

    This work presents the CAREM 25 reactor depressurization analysis results as an alternative form of accidents mitigation. Such results will help to determine design pressure valves for the emergency injection system as well as the depressurization valve diameter. Calculations were made with BLOW.MOD2 program. (Author) [es

  15. Study on depressurization measurements and effect in PWR

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

    Implementation of new regulations on nuclear powered plant design and operation raise new design and management requirement for plants, and the operational plants also need accident management to enhance the reactor operation safety. Thus, for sake of reducing risk of high-pressure and mitigating the consequence, depressurization is a measure carried out to reduce primary pressure. With SCDAP/RELAP5 this paper studies the depressurization measurements and effect factors in pressurized water reactor under the important severe accident sequences induced by very small break lost of coolant accident (VSBLOCA), anticipated transient without scram (ATWS) and station blackout (SBO) plus auxiliary feedwater failure. (author)

  16. Control device for start-up of reactor depressurization system

    International Nuclear Information System (INIS)

    Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

    1998-01-01

    The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

  17. Experimental Investigation on the Behavior of Supercritical CO2 during Reservoir Depressurization.

    Science.gov (United States)

    Li, Rong; Jiang, Peixue; He, Di; Chen, Xue; Xu, Ruina

    2017-08-01

    CO 2 sequestration in saline aquifers is a promising way to address climate change. However, the pressure of the sequestration reservoir may decrease in practice, which induces CO 2 exsolution and expansion in the reservoir. In this study, we conducted a core-scale experimental investigation on the depressurization of CO 2 -containing sandstone using NMR equipment. Three different series of experiments were designed to investigate the influence of the depressurization rate and the initial CO2 states on the dynamics of different trapping mechanisms. The pressure range of the depressurization was from 10.5 to 4.0 MPa, which covered the supercritical and gaseous states of the CO 2 (named as CO 2 (sc) and CO 2 (g), respectively). It was found that when the aqueous phase saturated initially, the exsolution behavior strongly depended on the depressurization rate. When the CO 2 and aqueous phase coexisting initially, the expansion of the CO 2 (sc/g) contributed to the incremental CO 2 saturation in the core only when the CO 2 occurred as residually trapped. It indicates that the reservoir depressurization has the possibility to convert the solubility trapping to the residual trapping phase, and/or convert the residual trapping to mobile CO 2 .

  18. Costs of radon diagnostics and mitigation in school building

    International Nuclear Information System (INIS)

    Leovic, K.W.; Rector, H.; Nagda, N.

    1992-01-01

    To determine the costs of radon diagnostics and active soil depressurization (ASD) system installation in schools, seven radon mitigators with extensive experience in school buildings were surveyed. The cost data were determined by providing the mitigators with two scenarios of open-quotes typicalclose quotes school buildings with elevated radon levels. The questionnaire in schools: (1) Review Construction Plans, (2) Conduct Diagnostic Measurements, (3) Design Mitigation System, (4) Purchase ASD Material, and (5) Install and Checkout ASD System. Based on the results of this survey, it is estimated that the average cost of ASD diagnostics and mitigation in a typical school would be roughly $0.50 per ft 2 . However, these costs would be higher in schools with extensive subslab walls, very poor PFE, and lower in simple schools with very good PFE and no subslab barriers to communication. The variations in costs provided by the mitigators are primarily due to the influences of (1) experience and practices of the mitigation companies, (2) ASD system requirements as perceived by the respondents, and (3) the degree of involvement by the school system in the process

  19. Cost and effectiveness of radon-resistant features in new school buildings

    International Nuclear Information System (INIS)

    Craig, A.B.; Leovie, K.W.

    1991-01-01

    Recent concerns over elevated levels of radon in existing buildings have prompted the design and construction of a number of school buildings that either are radon resistant or incorporate features that facilitate post-construction mitigation if needed. This paper described initial results of a study of several schools with radon-resistant features that were recently constructed in the northeastern U.S. These designs are generally based on experience with radon mitigation in existing houses and schools and radon-resistant new house construction. The study was limited to slab-on-grade schools, where the most common radon-resistant school design is active subslab depressurization (ASD). The additional construction costs for eight schools built with ASD ranged from $3 to $11 per square meter of slab area. The radon contractors who designed these systems have tended to overdesign the radon-reduction systems in the absence of specific written guidance to follow to lessen potential liability in the event of system failure. Design features include detailed sealing of all slab cracks, multiple exhaust stacks, and extensive subslab piping

  20. The effect of the rate of hydrostatic pressure depressurization on cells in culture.

    Science.gov (United States)

    Tworkoski, Ellen; Glucksberg, Matthew R; Johnson, Mark

    2018-01-01

    Changes in hydrostatic pressure, at levels as low as 10 mm Hg, have been reported in some studies to alter cell function in vitro; however, other studies have found no detectable changes using similar methodologies. We here investigate the hypothesis that the rate of depressurization, rather than elevated hydrostatic pressure itself, may be responsible for these reported changes. Hydrostatic pressure (100 mm Hg above atmospheric pressure) was applied to bovine aortic endothelial cells (BAECs) and PC12 neuronal cells using pressurized gas for periods ranging from 3 hours to 9 days, and then the system was either slowly (~30 minutes) or rapidly (~5 seconds) depressurized. Cell viability, apoptosis, proliferation, and F-actin distribution were then assayed. Our results did not show significant differences between rapidly and slowly depressurized cells that would explain differences previously reported in the literature. Moreover, we found no detectable effect of elevated hydrostatic pressure (with slow depressurization) on any measured variables. Our results do not confirm the findings of other groups that modest increases in hydrostatic pressure affect cell function, but we are not able to explain their findings.

  1. Mitigation diagnostics

    International Nuclear Information System (INIS)

    Hall, S.T.

    1990-01-01

    This paper reports that experience in the remediation of schools and other large buildings has shown the importance of the effects of both the location of geologic sources and HVAC-induced distribution of indoor radon. In general, elevated radon in areas of schools with evenly distributed HVAC pressures are correlated with maximum soil radon emanations. However, strong or unequal HVAC effects can redistribute indoor radon to areas away from the direct source. Effective remediation required a complete understanding of both contributions. In some schools, highest indoor radon levels were located near large return ducts and were attributed to proximity to negative HVAC pressure. Successful sub-slab depressurization systems were installed, however, in rooms with lower indoor but greatest sub-slab radon levels, closest to the source. This shows the inadequacy of using indoor radon levels alone as a basis for remediation. Wings of two other schools with radon problems have equivalent window fan coil units in rooms of equal size and no central HVAC system. Highest indoor radon levels correlated well with highest sub-slab radon levels due to the equivalent effects of the window units. Diagnostic tests in other schools have revealed: blockwall radon transport to upper floors; high blockwall radon adjacent to sub-slab sources; and elevated indoor radon over crawlspace being drawn upward by HVAC-induced negative pressure, determined from indoor to outdoor micromanometer measurements

  2. Rapid depressurization of a compressible fluid

    International Nuclear Information System (INIS)

    Dang, M.; Dupont, J.F.; Weber, H.

    1978-08-01

    The rapid depressurization of a plenum is a situation frequently encountered in the dynamical analysis of nuclear gas cycles of the HHT type. Various methods of numerical analyses for a 1-dimensional flow model are examined: finite difference method; control volume method; method of characteristics. Based on the shallow water analogy to compressible flow, the numerical results are compared with those from a water table set up to simulate a standard problem. (Auth.)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  4. Maximum Recoverable Gas from Hydrate Bearing Sediments by Depressurization

    KAUST Repository

    Terzariol, Marco

    2017-11-13

    The estimation of gas production rates from hydrate bearing sediments requires complex numerical simulations. This manuscript presents a set of simple and robust analytical solutions to estimate the maximum depressurization-driven recoverable gas. These limiting-equilibrium solutions are established when the dissociation front reaches steady state conditions and ceases to expand further. Analytical solutions show the relevance of (1) relative permeabilities between the hydrate free sediment, the hydrate bearing sediment, and the aquitard layers, and (2) the extent of depressurization in terms of the fluid pressures at the well, at the phase boundary, and in the far field. Close form solutions for the size of the produced zone allow for expeditious financial analyses; results highlight the need for innovative production strategies in order to make hydrate accumulations an economically-viable energy resource. Horizontal directional drilling and multi-wellpoint seafloor dewatering installations may lead to advantageous production strategies in shallow seafloor reservoirs.

  5. Depressurization accident analyses for the Fort St. Vrain Reactor

    International Nuclear Information System (INIS)

    Paul, D.D.

    1976-01-01

    Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cooling system. Results are presented for the Pelton circulators operating at 10,550, 8800, and 7000 rpm. Maximum temperatures of selected components are plotted as a function of time during the transient. None of the components studied exceeded the temperature at which failure or damage may occur. However, there must be sufficient mixing of the outlet gas in the lower plenum to insure the integrity of the steel liners of the steam generator inlet ducts

  6. Vacuum horizontal drainage for depressurization of uranium tailings

    International Nuclear Information System (INIS)

    Pakalnis, R.; Chedsey, G.; Robertson, A.M.; Follin, S.

    1985-01-01

    A recent advance in tailings slope depressurization is the application of vacuum assist horizontal drainage. Horizontal drains have been used for several decades to reduce water pressures in slopes in order to improve stability. The benefit from vacuum assist arises from an increased hydraulic gradient caused by induced negative atmospheric pressures. The vacuum assist system has, since its inception in 1982, been successfully employed at two soil and four rock slope projects located in Western Canada. This paper describes the first application of this system in the United States. The technical feasibility of employing vacuum assisted horizontal drains to depressurize a uranium tailings dam near Riverton, Wyoming has been evaluated. Two horizontal drains (300 ft.) were installed and their effect monitored by nine piezometers. The study was conducted over a three-week internal with vacuum being applied for three and four day periods. The drawdowns achieved through vacuum drainage was found to be approximately double that obtained by gravity alone. The volume of water exhausted under vacuum during the seven day interval was approximately double that obtained by gravity alone

  7. DSMC Simulations of Disturbance Torque to ISS During Airlock Depressurization

    Science.gov (United States)

    Lumpkin, F. E., III; Stewart, B. S.

    2015-01-01

    The primary attitude control system on the International Space Station (ISS) is part of the United States On-orbit Segment (USOS) and uses Control Moment Gyroscopes (CMG). The secondary system is part of the Russian On orbit Segment (RSOS) and uses a combination of gyroscopes and thrusters. Historically, events with significant disturbances such as the airlock depressurizations associated with extra-vehicular activity (EVA) have been performed using the RSOS attitude control system. This avoids excessive propulsive "de-saturations" of the CMGs. However, transfer of attitude control is labor intensive and requires significant propellant. Predictions employing NASA's DSMC Analysis Code (DAC) of the disturbance torque to the ISS for depressurization of the Pirs airlock on the RSOS will be presented [1]. These predictions were performed to assess the feasibility of using USOS control during these events. The ISS Pirs airlock is vented using a device known as a "T-vent" as shown in the inset in figure 1. By orienting two equal streams of gas in opposite directions, this device is intended to have no propulsive effect. However, disturbance force and torque to the ISS do occur due to plume impingement. The disturbance torque resulting from the Pirs depressurization during EVAs is estimated by using a loosely coupled CFD/DSMC technique [2]. CFD is used to simulate the flow field in the nozzle and the near field plume. DSMC is used to simulate the remaining flow field using the CFD results to create an in flow boundary to the DSMC simulation. Due to the highly continuum nature of flow field near the T-vent, two loosely coupled DSMC domains are employed. An 88.2 cubic meter inner domain contains the Pirs airlock and the T-vent. Inner domain results are used to create an in flow boundary for an outer domain containing the remaining portions of the ISS. Several orientations of the ISS solar arrays and radiators have been investigated to find cases that result in minimal

  8. Temporary vs. Permanent Sub-slab Ports: A Comparative ...

    Science.gov (United States)

    Vapor intrusion (VI) is the migration of subsurface vapors, including radon and volatile organic compounds (VOCs), from the subsurface to indoor air. The VI exposure pathway extends from the contaminant source, which can be impacted soil, non-aqueous phase liquid, or contaminated groundwater, to indoor air-exposure points. Therefore, contaminated matrices may include groundwater, soil, soil gas, and indoor air. VOC contaminants of concern typically include halogenated solvents such as trichloroethene, tetrachloroethene, and chloroform, as well as petroleum hydrocarbons, such as the aromatic VOCs benzene, toluene, and xylenes. Radon is a colorless radioactive gas that is released by radioactive decay of radionuclides in rock and soil that migrate into homes through VI in a similar fashion to VOCs. This project focused on the performance of permanent versus temporary sub-slab sampling ports for the determination of VI of halogenated VOCs and radon into an unoccupied house. VOC and radon concentrations measured simultaneously in soil gas using collocated temporary and permanent ports appeared to be independent of the type of port. The variability between collocated temporary and permanent ports was much less than the spatial variability between different locations within a single residential duplex. The agreement of the majority of VOC and radon concentrations, 0–36% relative percent difference, and 2–19% relative standard deviation respectively, of each sub-sl

  9. Perspectives on Severe Accident Management by Depressurization and External Water Injection under Extended SBO Conditions

    International Nuclear Information System (INIS)

    Seol, Wookcheol; Park, Jongwoon

    2014-01-01

    Three major issues of severe accident management guideline (SAMG) after this sort of extended SBO would be depressurization of the primary system, external water injection and hydrogen management inside a containment. Under this situation, typical SAM actions would be depressurization and external water delivery into the core. However, limited amount of external water would necessitate optimization between core cooling, containment integrity and fission product removal. In this paper, effects of SAM actions such as depressurization and external water injection on the reactor and containment conditions after extended SBO are analyzed using MAAP4 code. Positive and negative aspects are discussed with respect to core cooling and fission product retention inside a primary system. Conclusions are made as following: Firstly, early depressurization action itself has two-faces: positive with respect to delay of the reactor vessel failure but negative with respect to the containment failure and fission product retention inside the primary system. Secondly, in order to prevent containment overpressure failure after external water injection, re-closing of PORV later should be considered in SAM, which has never been considered in the previous SAMG. Finally, in case of external water injection, the flow rate should be optimized considering not only the cooling effect but also the long term fission product retention inside the primary system

  10. Analysis of design and operational effects of filtered containment venting on depressurization and fission product release

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon; Seol, Wook-Cheol; Kim, Jisu [Dongguk Univ., Gyeongbuk (Korea, Republic of)

    2017-03-15

    Effects of design and operational parameters of filtered containment venting system during a specified containment depressurization and relative aero sol release amount are analyzed. The analyses is performed by using the MAAP4 code for the APR1400 reactor. Major results uniquely identified from the analyses can be noted as following: Even though containment depressurization is accelerated as the pipe size increases, the venting system solution is also depleted earlier. Elapsed times to reach lower end pressure of 2 bar are nearly identical regardless of the vent initiation pressure and thus early venting is not much beneficial than late venting. Stroke time of the isolation valves has no effect on the depressurization performance and thus slow opening is beneficial for load reduction from the vent effluent.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  13. Automatic depressurization system of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Fujii, Masahiko.

    1993-01-01

    In the present invention, depressurization is conducted while keeping versatility and retardancy of a water injection system so that safety is improved. That is, a means that judges whether a turbine driving water injection system is operated or not by the following conditions. (1) a discharging pressure of the turbine driving pump is greater than a set value, (2) a flow rate of the turbine driving water injection system is greater than a set value, (3) an injection valve of the turbine driving water injection system into a reactor is opened, or combination of (1) to (3). With such procedures, when an automatic depressurization system is necessary during operation of the turbine driving water injection system, reactor pressure is decreased till a low pressure water injection system is operated, but pressure is not decreased to such a level that the turbine driving water injection system is isolated. Therefore, versatility and retardancy of the water injection system are ensured. As a result, reliability of a reactor cooling means is improved. (I.S.)

  14. Ground Source Heat Pump Sub-Slab Heat Exchange Loop Performance in a Cold Climate

    Energy Technology Data Exchange (ETDEWEB)

    Mittereder, Nick [IBACOS, Inc., Pittsburgh, PA (United States); Poerschke, Andrew [IBACOS, Inc., Pittsburgh, PA (United States)

    2013-11-01

    This report presents a cold-climate project that examines an alternative approach to ground source heat pump (GSHP) ground loop design. The innovative ground loop design is an attempt to reduce the installed cost of the ground loop heat exchange portion of the system by containing the entire ground loop within the excavated location beneath the basement slab. Prior to the installation and operation of the sub-slab heat exchanger, energy modeling using TRNSYS software and concurrent design efforts were performed to determine the size and orientation of the system. One key parameter in the design is the installation of the GSHP in a low-load home, which considerably reduces the needed capacity of the ground loop heat exchanger. This report analyzes data from two cooling seasons and one heating season.

  15. A study on design enhancement of automatic depressurization system in a passive PWR

    International Nuclear Information System (INIS)

    Yu, Sung Sik

    1993-02-01

    In a Passive PWR, the successful actuation of the Automatic Depressurization System is essentially required so that no core damage is occurred following small LOCA. But it has been shown in the previous studies that Core Damage Frequency form small LOCA is significantly caused by unavailability of the ADS. In this study, the design vulnerabilities impacting the ADS unavailability are identified through the reliability assessment using the fault tree methodology and then the design enhancements towards improving the system reliability are developed. A series of small LOCA analyses using RELAP5 code are performed to validate the system requirements for the successful depressurization and to study the thermal-hydraulic feasibility of the proposed design enhancements. The impact on CDF according to the change of system unavailability is also analyzed. In addition, aqualitative analysis is performed to reduce the inadvertent opening of the ADS valves. From the results of the analyses, the ADS is understood to have less incentive on the reliability improvement through system simplification. It is found that based on system characteristics, the major contributor to the system unavailability is the first stage. A series-parallel configuration with two trains of eight valves, which shows a higher reliability compared to the base ADS design, is recommended as an alternative first stage of the ADS. In addition, establishment of the appropriate ADS operation strategy is proposed such as allowing manual operation of the first stage and allowing the forced depressurization using the normal residual heat removal system connected to the RCS following the successful depressurization up to the 3rd stage and the failure of the 4th stage

  16. A depressurization assistance control based on the posture of a seated patient on a wheelchair.

    Science.gov (United States)

    Chugo, Daisuke; Fujita, Kazuya; Sakaida, Yuki; Yokota, Sho; Takase, Kunikatsu

    2011-01-01

    For reducing the risk of pressure sore caused by long period sitting on a wheelchair, we develop a depressurization motion assistance system which is low cost and suitable for practical use. Our developing system consists of a seating cushion which the patient sits on and four air cells which can lift or incline the seating cushion. Each air cell is actuated by small air compressor, which can drive using batteries on the wheelchair respectively, and each compressor has a pressure sensor on its body. In this paper, our key ideas are two topics. One topic is mechanical design for practical use. We realize thin mechanism which enables easy implementation to the general wheelchair. For realizing this thinly design, we develop the tilt mechanism using elasticity of acrylic resin and the controller which uses only pressure sensors for estimating its lifting height and inclination. The other topic is assistance control scheme based on the patient's depressurization operation for increasing a rehabilitation performance. For realizing the proposed control scheme, we analyze the hip depressurization operation by the nursing specialists and use its results for estimating the patient's condition. Using our system, the patient can depressurize by his own will on the general wheelchair easily. The performance of our system is verified by experiments using our prototype. © 2011 IEEE

  17. Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients.

    Science.gov (United States)

    Jun, Jonathan C; Unnikrishnan, Dileep; Schneider, Hartmut; Kirkness, Jason; Schwartz, Alan R; Smith, Philip L; Polotsky, Vsevolod Y

    2016-01-01

    Obstructive Sleep Apnea (OSA) describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP) is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep. Nine obese (BMI = 40.4 ± 3.5) subjects with severe OSA (AHI = 88.9 ± 6.8) adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20) or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods. 8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (p<0.05). The average 3% ODI during CPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (p<0.05). CPAP depressurization also induced more awake (p<0.05) and stage N1 (p<0.01) sleep, and less stage REM (p<0.05) with a trend towards decreased stage N3 (p = 0.064). Acute intermittent depressurization of CPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP.

  18. Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients.

    Directory of Open Access Journals (Sweden)

    Jonathan C Jun

    Full Text Available Obstructive Sleep Apnea (OSA describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep.Nine obese (BMI = 40.4 ± 3.5 subjects with severe OSA (AHI = 88.9 ± 6.8 adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20 or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods.8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (p<0.05. The average 3% ODI during CPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (p<0.05. CPAP depressurization also induced more awake (p<0.05 and stage N1 (p<0.01 sleep, and less stage REM (p<0.05 with a trend towards decreased stage N3 (p = 0.064.Acute intermittent depressurization of CPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP.

  19. Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

    International Nuclear Information System (INIS)

    Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

    1999-12-01

    In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

  20. Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

    2007-01-01

    As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

  1. Depressurization accidents in a medium-sized high-temperature gas reactor

    International Nuclear Information System (INIS)

    Ron, S.; Tzoref, J.; Gal, D.

    1992-01-01

    The amount of fission product release during a core heatup accident in a medium-sized high-temperature gas reactor depends on the size of the inadvertent opening in the primary circuit; this dependence is assessed. The opening triggers a depressurization event that is assumed to be coupled with the failure of the forced circulation in both decay-heat removal systems. The scenario investigated is a beyond-design-base accident. The DSNP modular simulation code is used. This paper reports that a two-dimensional model is developed to simulate the HTR-500 design. The study shows that the depressurization process does not contribute significantly to the sweeping out (from the primary circuit) of fission products released from the fuel during the core heatup. There is also no significant variation in the results when the opening size is >33 cm 2 , and only a slight sensitivity is found when the rupture size is between 3.3 and 33 cm 2 . The fission product release decreases considerably in the range from 1 to 3.3 cm 2 . The small-sized rupture is of major significance, as the failure of the relief valves to reclose increases the frequency of the event

  2. Nuclear reactor steam depressurization valve

    International Nuclear Information System (INIS)

    Moore, G.L.

    1991-01-01

    This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

  3. Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

    International Nuclear Information System (INIS)

    Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

    1982-01-01

    A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

  4. Geomechanical response of permafrost-associated hydrate deposits to depressurization-induced gas production

    Science.gov (United States)

    Rutqvist, J.; Moridis, G.J.; Grover, T.; Collett, T.

    2009-01-01

    In this simulation study, we analyzed the geomechanical response during depressurization production from two known hydrate-bearing permafrost deposits: the Mallik (Northwest Territories, Canada) deposit and Mount Elbert (Alaska, USA) deposit. Gas was produced from these deposits at constant pressure using horizontal wells placed at the top of a hydrate layer (HL), located at a depth of about 900??m at the Mallik site and 600??m at the Mount Elbert site. The simulation results show that general thermodynamic and geomechanical responses are similar for the two sites, but with substantially higher production and more intensive geomechanical responses at the deeper Mallik deposit. The depressurization-induced dissociation begins at the well bore and then spreads laterally, mainly along the top of the HL. The depressurization results in an increased shear stress within the body of the receding hydrate and causes a vertical compaction of the reservoir. However, its effects are partially mitigated by the relatively stiff permafrost overburden, and compaction of the HL is limited to less than 0.4%. The increased shear stress may lead to shear failure in the hydrate-free zone bounded by the HL overburden and the downward-receding upper dissociation interface. This zone undergoes complete hydrate dissociation, and the cohesive strength of the sediment is low. We determined that the likelihood of shear failure depends on the initial stress state as well as on the geomechanical properties of the reservoir. The Poisson's ratio of the hydrate-bearing formation is a particularly important parameter that determines whether the evolution of the reservoir stresses will increase or decrease the likelihood of shear failure.

  5. Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

    Energy Technology Data Exchange (ETDEWEB)

    Mueftueoglu, A.K.; Feltus, M.A. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

    Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and other signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.

  6. Experimental Investigation Evaporation of Liquid Mixture Droplets during Depressurization into Air Stream

    Science.gov (United States)

    Liu, L.; Bi, Q. C.; Terekhov, Victor I.; Shishkin, Nikolay E.

    2010-03-01

    The objective of this study is to develop experimental method to study the evaporation process of liquid mixture droplets during depressurization and into air stream. During the experiment, a droplet was suspended on a thermocouple; an infrared thermal imager was used to measure the droplet surface temperature transition. Saltwater droplets were used to investigate the evaporation process during depressurization, and volatile liquid mixtures of ethanol, methanol and acetone in water were applied to experimentally research the evaporation into air stream. According to the results, the composition and concentration has a complex influence on the evaporation rate and the temperature transition. With an increase in the share of more volatile component, the evaporation rate increases. While, a higher salt concentration in water results in a lower evaporation rate. The shape variation of saltwater droplet also depends on the mass concentration in solution, whether it is higher or lower than the eutectic point (22.4%). The results provide important insight into the complex heat and mass transfer of liquid mixture during evaporation.

  7. Depressurization as a means of leak checking large vacuum vessels

    International Nuclear Information System (INIS)

    Callis, R.W.; Langhorn, A.; Petersen, P.I.; Ward, C.; Wesley, J.

    1985-01-01

    A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, the vessel need only be depressurized 2 psi, thus allowing personnel inside to perform the sniffing operation. Second, the sniffing probe used (Leybold--Heraus ''Quick Test'') could detect a change in helium concentration as small as 100 ppb, which allows for faster scanning of the vessel inferior. Use of this technique to find an elusive 10 -3 Torrxl/s leak in the Doublet III tokamak vacuum vessel will be presented

  8. Cost and effectiveness of radon-resistant features in new school buildings

    International Nuclear Information System (INIS)

    Craig, A.B.; Leovic, K.W.; Saum, D.W.

    1991-01-01

    Recent concerns over elevated levels of radon in existing buildings have prompted the design and construction of a number of school buildings that either are radon resistant or incorporate features that facilitate post-construction mitigation if needed. This paper describes initial results of a study of several schools with radon-resistant features that were recently constructed in the northeastern US. These designs are generally based on experience with radon mitigation in existing houses and schools and radon-resistant new house construction. The study was limited to slab-on-grade schools, where the most common radon-resistant school design is active subslab depressurization (ASD). The additional construction costs for eight schools built with ASD ranged from $3 to $11 per square meter of slab area. The radon contractors who designed these systems have tended to overdesign the radon-reduction systems in the absence of specific written guidance to follow to lessen potential liability in the event of system failure. Design features include detailed sealing of all stab cracks, multiple exhaust stacks, and extensive subslab piping. Recent Environmental Protection Agency (EPA) research on radon mitigation suggests that simpler ASD systems may provide sufficient radon resistance in new large buildings at lower costs. Components of a specification for radon-resistant school construction are discussed, based on comments from radon system designers. Another school being studied was built with a heating, ventilating, and air-conditioning (HVAC) pressurization radon control system, and considerations for this type of system are examined

  9. Dynamic solution of vessel depressuring; Simulacao dinamica de despressurizacao em vasos

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Marco Tulio; Silva Netto, Rafael [Chemtech Servicos de Engenharia e Software Ltda., Rio de Janeiro, RJ (Brazil); Aires, Joyce Stone S. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil). Centro de Pesquisas (CENPES)

    2004-07-01

    Vessel depressuring is an important phenomenon on chemical and petrochemical processes, specially those related to oil and gas exploration, production and processing. The correct modeling of this phenomenon and prediction of the temperatures, mass and thermal rates involved is essential to the adequate design of the equipment, materials specification and safety standards assurance. To fulfill these requirements, we developed a method to simulate the phenomenon of depressuring. In this approach, the dynamic process is discretized along time, being calculated at each interval the conditions inside the vessel, material flow and heat exchanges with the environment, through mass and energy balances and thermodynamic equilibrium calculation. The method can include different models to calculate heat exchanges and flow through the relief valve, and new methods can be incorporated if necessary. The efficiency of the method was verified by comparing its results with the ones obtained by market-leaders process simulators, when it was proved the robustness of the method and precision of the results. To increase its flexibility of use, the method was incorporated to the PETROBRAS Process Simulator - PETROX (PETROBRAS, 2004), in a development made by Chemtech under a contract with PETROBRAS and already used in large scale by PETROBRAS to simulate its process units. (author)

  10. Development and evaluation of a new depressurization spillage test for residential gas-fired combustion appliances : final report

    International Nuclear Information System (INIS)

    Edwards, P.

    2005-07-01

    This paper presented a newly developed combustion depressurization spillage test for residential combustion appliances. The test uses carbon dioxide (CO 2 ) that is produced in the fuel combustion process as a tracer gas. The test accurately measures the amount of combustion spillage from residential combustion appliances and their venting systems when they operate at certain levels of depressurization. Seven commonly used gas-fired appliances were used to evaluate the new test as well as the appliances. These included 2 power-vented storage-tank water heaters, 1 mid-efficiency furnace, 2 high-efficiency condensing furnaces, and 2 direct-vent gas fireplaces. Tests were performed for each unit with the test room initially depressurized by 50 Pa compared with the pressure outside the room. If the combustion spillage exceeded 2 per cent, the test was repeated with the room depressurized by 20 Pa, and then by 5 Pa. Each appliance was operated for 5 minutes of burner operation during which time the burner fuel consumption, the concentration of CO 2 and the exhaust fan flow rate were monitored. Measurements were taken for 2 minutes following burner shut off. The amount of CO 2 that was released into the test room from the appliance and its venting system was determined from the measurements and then compared with the amount of CO 2 that would be produced by combustion of the fuel that was consumed during the test. The ratio of the 2 provided a direct measure of the combustion spillage of the appliance and its venting system. The study revealed that 3 products had undetectable levels of combustion spillage, 3 products had low, but measurable combustion spillage, and 1 product had significant combustion spillage. refs., tabs., figs

  11. Application case study of AP1000 automatic depressurization system (ADS) for reliability evaluation by GO-FLOW methodology

    Energy Technology Data Exchange (ETDEWEB)

    Hashim, Muhammad, E-mail: hashimsajid@yahoo.com; Hidekazu, Yoshikawa, E-mail: yosikawa@kib.biglobe.ne.jp; Takeshi, Matsuoka, E-mail: mats@cc.utsunomiya-u.ac.jp; Ming, Yang, E-mail: myang.heu@gmail.com

    2014-10-15

    Highlights: • Discussion on reasons why AP1000 equipped with ADS system comparatively to PWR. • Clarification of full and partial depressurization of reactor coolant system by ADS system. • Application case study of four stages ADS system for reliability evaluation in LBLOCA. • GO-FLOW tool is capable to evaluate dynamic reliability of passive safety systems. • Calculated ADS reliability result significantly increased dynamic reliability of PXS. - Abstract: AP1000 nuclear power plant (NPP) utilized passive means for the safety systems to ensure its safety in events of transient or severe accidents. One of the unique safety systems of AP1000 to be compared with conventional PWR is the “four stages Automatic Depressurization System (ADS)”, and ADS system originally works as an active safety system. In the present study, authors first discussed the reasons of why four stages ADS system is added in AP1000 plant to be compared with conventional PWR in the aspect of reliability. And then explained the full and partial depressurization of RCS system by four stages ADS in events of transient and loss of coolant accidents (LOCAs). Lastly, the application case study of four stages ADS system of AP1000 has been conducted in the aspect of reliability evaluation of ADS system under postulated conditions of full RCS depressurization during large break loss of a coolant accident (LBLOCA) in one of the RCS cold legs. In this case study, the reliability evaluation is made by GO-FLOW methodology to determinate the influence of ADS system in dynamic reliability of passive core cooling system (PXS) of AP1000, i.e. what will happen if ADS system fails or successfully actuate. The GO-FLOW is success-oriented reliability analysis tool and is capable to evaluating the systems reliability/unavailability alternatively to Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) tools. Under these specific conditions of LBLOCA, the GO-FLOW calculated reliability results indicated

  12. Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients

    Science.gov (United States)

    Jun, Jonathan C.; Unnikrishnan, Dileep; Schneider, Hartmut; Kirkness, Jason; Schwartz, Alan R.; Smith, Philip L.; Polotsky, Vsevolod Y.

    2016-01-01

    Background Obstructive Sleep Apnea (OSA) describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP) is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep. Methods Nine obese (BMI = 40.4 ± 3.5) subjects with severe OSA (AHI = 88.9 ± 6.8) adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20) or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods. Results 8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (pCPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (pCPAP depressurization also induced more awake (pCPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP. PMID:26731735

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  14. BLOW.MOD2: program for a vessel depressurization calculation with the contribution of structures

    International Nuclear Information System (INIS)

    Doval, A.

    1990-01-01

    The BLOW.MOD2 program developed to calculate pressure vessels' depressurization is presented, considering heat contribution of the structures. The results are opposite to those obtained from other more complex numerical models, being the comparison extremely satisfactory. BLOW.MOD2 is a software of the 'Systems Sub-Branch', INVAP S.E. (Author) [es

  15. Testing of indoor radon reduction techniques in central Ohio houses: Phase 2 (Winter 1988-1989). Final report, September 1988-May 1989

    International Nuclear Information System (INIS)

    Findlay, W.O.; Robertson, A.; Scott, A.G.

    1990-05-01

    The report gives results of tests of developmental indoor radon reduction techniques in nine slab-on-grade and four crawl-space houses near Dayton, Ohio. The slab-on-grade tests indicated that, when there is a good layer of aggregate under the slab, the sub-slab ventilation (SSV) mitigation technique, with only one or two suction pipes, can generally reduce indoor concentrations below 2 pCi/L (86 to 99% reduction). These reductions can be achieved even when: there are forced-air supply ducts under the slab; the slab is large (up to 2600 sq ft); and the foundation walls are hollow block. Operating the SSV system in suction always gave greater reductions than did operating in pressure. The crawl-space tests demonstrated that depressurizing under a plastic liner over the crawl-space floor was able to reduce living-area radon concentrations below 2 pCi/L (81 to 96% reduction). The performance of such sub-liner depressurization gave better reductions than did crawl-space ventilation (blowing air into, or out of, the crawl space). Completely covering the crawl-space floor with plastic sheeting was not always necessary to get adequate performance

  16. Follow-up durability measurements and mitigation-performance improvement tests in 38 Eastern Pennsylvania houses having indoor radon-reduction systems. Final report, Oct 89-Feb 90

    International Nuclear Information System (INIS)

    Findlay, W.O.; Robertson, A.; Scott, A.G.

    1991-03-01

    The report gives results of follow-up tests in 38 difficult-to-mitigate Pennsylvania houses where indoor radon reduction systems had been installed 2 to 4 years earlier. Objectives were to assess system durability, methods for improving performance, and methods for reducing installation and operating costs. The durability tests indicated that the 38 systems have not experienced any significant degradation in indoor radon levels or in system flows/suctions, except in 6 houses where system fans failed, and in houses where homeowners turned off the systems. Tests to improve performance indicated that nearly all of the elevated residual radon levels are due to re-entrainment back into the house of very-high-radon exhaust gas from the soil depressurization systems, and to radon release from well water. Tests to reduce system costs showed that premitigation sub-slab suction field measurements can help prevent installation of too many suction pipes when communication is good, but suggest a need for too many pipes when communication is poor. Soil depressurization fans could not be turned down to the extent expected in some systems that were over-designed. Between 6 and 42% of the exhausted air was withdrawn from the house

  17. Mudstone depressurization behaviour in an open pit coal mine, Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Marchand, G.; Waterhouse, J. [Golder Associates, West Perth, WA (Australia); Crisostomo, J. [PT Adaro Indonesia, Jakarta (Indonesia)

    2010-07-01

    Mining activities in the Tutupan mine in Indonesia began in the mid-1990s. The open pit mine's coal seams are interbedded with fine-grained sandstones, mudstones, and carbonaceous mudstones. Slope stability analyses at the pit have integrated hydrogeology with geotechnical engineering analyses to optimize slope designs and reduce the risk of slope failure. This paper discussed the impact of mining and dewatering on mudstone depressurization. Sensors were placed at key points in the mine to obtain data related to the mudstone units. Reductions in pore pressure occurred as a result of groundwater flow away from the observed zones, increases in porosity, and increases in total porosity caused by an expansion of the rock mass as a result of drainage and hydrostatic unloading. Mudstone pore pressure trends with time were interpreted by determining the thickness of the mudstone unit, the presence or absence of known thin sandstone beds, unloading from overhead mining activities, and the position of the mudstone within the sedimentary sequence. The study showed that unloading activities have a significant impact on pore pressure in thick mudstone units, regardless of the depth, thickness, or properties of the unit. Pore pressure within high wall mudstone units typically decreased to values equivalent to the elevation of the unit where it was exposed to dips in a high wall. The dewatering of sandstone units in low walls caused a decline in pore pressure within the thick mudstone units located beneath the sandstones. Differences in primary permeabilities were attributed to greater fracturing in deeper and stronger rock units. 3 refs., 4 figs.

  18. Mathematical simulation of the drying of suspensions and colloidal solutions by their depressurization

    Science.gov (United States)

    Lashkov, V. A.; Levashko, E. I.; Safin, R. G.

    2006-05-01

    The heat and mass transfer in the process of drying of high-humidity materials by their depressurization has been investigated. The results of experimental investigation and mathematical simulation of the indicated process are presented. They allow one to determine the regularities of this process and predict the quality of the finished product. A technological scheme and an engineering procedure for calculating the drying of the liquid base of a soap are presented.

  19. Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

    Science.gov (United States)

    Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

    2017-10-01

    The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

  20. EPA program to demonstrate mitigation measures for indoor radon: initial results

    International Nuclear Information System (INIS)

    Henschel, D.B.; Scott, A.G.

    1986-01-01

    EPA has installed radon mitigation techniques in 18 concrete block basement homes in the Reading Prong region of eastern Pennsylvania. Three alternative active soil ventilation approaches were tested: suction on the void network within the concrete block basement walls; suction on the footing drain tile system; and suction on the aggregate underneath the concrete slab. The initial 18 mitigation installations were designed to demonstrate techniques which would have low to moderate installation and operating costs. Where effective closure of major openings in the block walls is possible, suction on the wall voids has proved to be extremely effective, able to reduce homes having very high radon Working Levels (up to 7 WL) to 0.02 WL and less. However, where inaccessible major openings are concealed within the wall, it is more difficult and/or more expensive to develop adequate suction on the void network, and performance is reduced. Testing is continuing to demonstrate the steps required to achieve high performance with wall suction in homes with such difficult-to close walls. Drain tile suction can be very effective where the drain tiles completely surround the home; drain tile suction is the least expensive and most aesthetic of the active soil ventilation approaches, but appears susceptible to spikes in radon levels when the basement is depressurized. Sub-slab suction as tested in this study - with one or two individual suction points in the slab - does not appear adequate to ensure sustained high levels of reduction on block wall basement homes; it appears to effectively treat slab-related soil gas entry routes so long as a uniform layer of aggregate is present, but it does not appear to effectively treat the wall-related entry routes. Closure of major openings might have improved sub-slab suction performance. 5 figures, 3 tables

  1. Development of a hybrid safety system: Actuation of the secondary automatic depressurization system at an early stage

    International Nuclear Information System (INIS)

    Nishimoto, Masae; Umezawa, Shigemitsu; Okabe, Kazuharu; Matsuoka, Tsuyoshi

    1996-01-01

    A Hybrid Safety System, which is an optimum combination of active and passive safety systems, has been developed in order to improve the safety, reliability and economic features of the next generation of PWRs. The passive safety systems include Automatic primary Depressurization System (ADS), Secondary Automatic Depressurization System (SADS), advanced accumulators, gravity injection system and so on. In this study the authors have improved the actuation logic of the passive safety systems. The original logic in the previous study actuates ADS at an early stage of an event such as a Loss-of-Coolant Accident (LOCA), and this is followed by the actuation of SADS. In this study they divide SADS into two systems. The first, small SADS, uses small valves corresponding to the relief valves of the conventional PWR plants. The second, large SADS, corresponds to the original SADS using multiple valves of large capacity. With the new logic, the passive systems are actuated during a typical small LOCA. Small LOCA analyses using several break areas were performed for a 1,400 MWe PWR plant with a Hybrid Safety System. The results predict that core uncovery does not occur in the case of a relatively small break area and that core heat removal during a small LOCA is improved in comparison with the analyses for conventional PWR plants, where the secondary pressure remains higher during the event. The results also predict that this new logic make it possible to reduce the ADS valve size and the actuation pressure setpoint of the passive safety systems

  2. Analytical method and result of radiation exposure for depressurization accident of HTTR

    International Nuclear Information System (INIS)

    Sawa, K.; Shiozawa, S.; Mikami, H.

    1990-01-01

    The Japan Atomic Energy Research Institute (JAERI) is now proceeding with the construction design of the High Temperature Engineering Test Reactor (HTTR). Since the HTTR has some characteristics different from LWRs, analytical method of radiation exposure in accidents provided for LWRs can not be applied directly. This paper describes the analytical method of radiation exposure developed by JAERI for the depressurization accident, which is the severest accident in respect to radiation exposure among the design basis accidents of the HTTR. The result is also described in this paper

  3. Feasibility study for the adoption of POSRV for KNGR safety depressurization system

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lim, Hong Sik; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

    1999-03-01

    The Korean Next Generation Reactor (KNGR) adopted an advanced design feature of safety depressurization system(SDS) to rapidly de pressurize the reactor coolant system(RCS) in case of beyond design basis events of severe accidents, or a highly unlikely event of a total loss of feedwater (TLOFW) to both steam generators. Two design approaches were considered for the KNGR SDS design. The use of bleed valves similar to those of ABB-CE's system 80+ is design option 1, while in design option 2, the Power Operated Safety Relief valve (POSRV) is considered to provide the combined function of overpressure protection and rapid depressurization. The purpose of this report is to investigate the feasibility of adoption of French SebimPOSRVs for KNGR SDS (design option 2). This report provides the methodology to analyze the TLOFW event with Sebim valves and presents the results of thermal hydraulic analyses using a best-estimate version CEFLASH-4AS/REM for the TLOFW event with feed and bleed. The analyses were performed using a preliminary KNGR design data. For design option 2, if the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible to mitigate the consequences of the TLOFW event with a sufficient margin. (Author). 22 refs., 3 tabs., 19 figs

  4. Pore network modelling of heavy oil depressurization : a parametric study of factors affecting critical gas saturation and three-phase relative permeabilities

    Energy Technology Data Exchange (ETDEWEB)

    Bondino, I.; McDougall, S.D. [Heriot-Watt Univ., Edinburgh, Scotland (United Kingdom); Hamon, G. [TotalFina Elf Exploration and Production (France)

    2002-07-01

    A review of how the bubble nucleation process affects the efficiency of heavy oil recovery was presented along with a discussion regarding a pore-scale simulator technique to depressurize heavy oil systems. A light oil depressurization simulation is also presented in which a straightforward instantaneous nucleation (IN) model and a more intricate progressive nucleation (PN) model have been used. Simulation results are compared to those derived from the heavy oil systems. The nucleation of bubbles, their growth by solute diffusion and expansion, plus the final stages of coalescence migration and production are the main steps in the depressurization process which were accounted for in a 3-phase simulator. The model can also determine the impact of bubble density and gas-oil diffusion coefficient on critical gas saturation and 3-phase relative permeability. The difference in results for light and heavy oils was also highlighted. In the first scenario, the evolution of gas was characterized by embryonic bubbles that are quickly and randomly nucleated once bubble-point pressure is reached. A stochastic algorithm was developed for PN from experimental observations. IN and PN observations were not necessarily contradictory. It was determined that the high interfacial tension of heavy oils leads to a more compact, capillary-dominated pattern of gas evolution compared to light oils, resulting in improved recoveries for heavy oil systems. 23 refs., 6 tabs., 23 figs.

  5. Mechanical Properties of Porous Titanium Structure Fabricated by Investment Casting with Pressurization/Depressurization System

    International Nuclear Information System (INIS)

    Kang, San; Lee, Ji-Woon; Hyun, Soong-Keun; Lee, Byong-Pil; Kim, Myoung-Gyun; Kim, Young-Jig

    2014-01-01

    A porous titanium structure was fabricated by investment casting with a pressurization/depressurization system, and its mechanical properties were studied. A Micro-Vickers hardness profile revealed that hardness gradually increased from the matrix to the metal/mold interface. A compression test was conducted on a single cell of the porous Ti structure. The theoretical and experimental values of yield strength were in good agreement. Such agreement suggested that the reaction layer did not affect the macro-mechanical properties of the porous Ti structure.

  6. Analysis of Depressurization Performance in Containment of Wolsong NPP Unit 1 through Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Lee, Sunghan; Kim, Jinhyuck; Suh, Nam Duk; Cho, Songwon

    2014-01-01

    Containment filtered venting system (CFVS) is designed to open and to close isolation valves passively by an operator. CFVS is operated when the containment pressure exceeds the design pressure (225 kPa(a)) and is closed when the containment pressure decreases below 151 kPa(a). The aim of this study is to analyze the depressurization performance of Wolsong unit 1 through CFVS during SBO. The thermal-hydraulic behavior in containment of Wolsong unit 1 was evaluated using the MELCOR 1.8.6 code developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC). In addition, in order to evaluate the effects of the CFVS according to the venting area, a sensitivity study depending on different venting area of the CFVS was conducted. Finally, an analysis of the effects of filtering and scrubbing of radioactive material for CFVS is important but not treated in this paper. The SBO accident is chosen to analyze the thermal-hydraulic behavior of Wolsong unit 1. During SBO, the analysis of CFVS affecting on the depressurization of the containment was conducted using MELCOR 1.8.6 code. Also, a sensitivity study was carried out to evaluate the depressurization performance according to the venting area of CFVS. The results show that the containment pressure is considerably decreased and the integrity of the containment could be maintained in case of CFVS operating. Therefore, CFVS has the capacity to keep the containment pressure below the design pressure during SBO. In addition, there are large differences in the containment pressure depending on venting area. We found that the decreasing rate of the pressure in the containment and water level in CFVS depends on the venting area. In the future, a proper requirement for CFVS sizing criteria according to accident scenarios such as LBLOCA, SBLOCA and SGTR, etc. should be evaluated in order to review the licensing for CFVS. Finally, analyses of aerosols, fission product, and radioactive material

  7. From insulation contracting to radon mitigation

    International Nuclear Information System (INIS)

    West, D.R.

    1990-01-01

    As the definition of house doctor has evolved over the past ten years and the field of energy services has grown more sophisticated, many contractors have expanded the services they offer their clients. This paper presents the story of one insulation contractor who has found a niche in radon testing and mitigation. The EPA now has a national program for the radon mitigator called the Radon Contractor Proficiency Program. The requirements include attending the Radon Technology for Mitigators course, passing an exam, and taking continuing education. In the Midwest, the most popular mitigation technique is the subslab depressurization system. To draw suction from under the slab, the system can take advantage of an existing sump crock or can penetrate the slab. Interior drain tiles collect water to empty into the crock, providing an excellent pathway to draw from. This mitigation process is explained

  8. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Ramirez G, C.; Chavez M, C.

    2012-10-01

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  9. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  10. Droplet Impact on a Heated Surface under a Depressurized Environment

    Science.gov (United States)

    Hatakenaka, Ryuta; Tagawa, Yoshiyuki

    2016-11-01

    Behavior of a water droplet of the diameter 1-3mm impacting on a heated surface under depressurized environment (100kPa -1kPa) has been studied. A syringe pump for droplet generation and a heated plate are set into a transparent acrylic vacuum chamber. The internal pressure of the chamber is automatically controlled at a target pressure with a rotary pump, a pressure transducer, and an electrical valve. A silicon wafer of the thickness 0.28 mm is mounted on the heater plate, whose temperature is directly measured by attaching a thermocouple on the backside. The droplet behavior is captured using a high-speed camera in a direction perpendicular to droplet velocity. Some unique behaviors of droplet are observed by decreasing the environmental pressure, which are considered to be due to two basic elements: Enhancement of evaporation due to the lowered saturation temperature, and shortage of pneumatic spring effect between the droplet and heated wall due to the lowered pressure of the air.

  11. Source Test Report for the 205 Delayed Coking Unit Drum 205-1201 and Drum 205-1202 Depressurization Vents (Marathon Petroleum Company LLC)

    Science.gov (United States)

    The 2010 Source Test was performed during the atmospheric depressurization step of the delayed coking process prior to the removal of petroleum coke from the coke drum. The 205 DCU was operated under a variety of conditions during the 2010 Source Test.

  12. An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-03-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

  13. Data report of ROSA/LSTF experiment SB-CL-32. 1% cold leg break LOCA with SG depressurization and no gas inflow

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2014-11-01

    An experiment SB-CL-32 was conducted on May 28, 1996 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-CL-32 simulated a 1% cold leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and no inflow of non-condensable gas from accumulator (ACC) tanks of emergency core cooling system. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after the break. After the initiation of AM action, auxiliary feedwater injection into the SG secondary-side was started with some delay. After the onset of AM action, the primary pressure decreased following the SG secondary-side pressure. Core uncovery by core boil-off started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first loop seal clearing (LSC). The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery by core boil-off took place before second LSC induced by steam condensation on ACC coolant injected into cold legs following the primary depressurization. The core liquid level recovered rapidly after the second LSC. The observed maximum fuel rod surface temperature was 772 K. The experiment was terminated when the continuous core cooling was confirmed because of the coolant injection by low pressure injection system after the isolation of ACC system. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal-hydraulic phenomena. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment SB-CL-32. (author)

  14. Assessing the effectiveness of slab flooring as a barrier to soil gas and radon infiltration

    International Nuclear Information System (INIS)

    Williamson, A.D.; Fowler, C.S.; McDonough, S.E.

    1995-01-01

    Experimental studies on the entry of soil gas and radon into slab-on-grade buildings have been carried out in instrumented, single-zone test structures. This work, as part of the Florida Radon Research Program, focused on the effectiveness of slab flooring variants as barriers to soil gas/radon entry. A second objective was the study of the role of subslab fill soil as both a potential source of and barrier to radon entry. Studies were made in well-sealed (∼ 600 mm 2 ELA) unoccupied test buildings placed on well-characterized, radium-bearing sandy fill soil. The buildings were instrumented with data acquisition systems to continuously monitor indoor radon concentrations, differential pressures at several subsurface locations, weather conditions, and soil moisture. The response of the structures to mechanical depressurization as well as natural driving forces was measured. Limited measurements were made regarding direct diffusive transport of radon through apparently intact concrete slabs, as well as transport through cracks in the floor structure

  15. Commercial mitigation techniques used in remediating a 2200 pCi/L public building

    International Nuclear Information System (INIS)

    Davidson, J.G.

    1990-01-01

    This paper reports on commercial mitigation techniques used in remediating a 2200 pCi/L public building. In March of 1989 EPA and Pa. DER officials were amazed to discover a school in Pennsylvania with levels in its library of 2200 pCi/L. The library was a 30 year old, three story slab-on-grade structure more like a commercial building than a typical school structure. It had three separate and complex HVAC systems. Initial diagnostics indicated radon levels under the slab at over 80,000 pCi/L. Further investigations revealed major entry routes and a HVAC system terribly out of balance. Remediation consisted of installing a complex sub-slab depressurization system with an exterior commercial fan unit, major entry route sealing, and working closely with a mechanical contractor to bring the HVAC systems back into balance. Initial post remediation testing showed a 99% drop in radon levels. Refinements to the system are still in progress

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

    International Nuclear Information System (INIS)

    Veneau, Tania

    1995-01-01

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

  17. An intermediate heat exchanging-depressurizing loop for nuclear hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Soo [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); No, Hee Cheon, E-mail: hcno@kaist.ac.k [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Yoon, Ho Joon; Lee, Jeong Ik [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2010-10-15

    Sulfur-iodine (SI) cycle should overcome many engineering challenges to commercialize and prove its feasibilities to compete other thermo-chemical cycles. Some critical issues such as structural material, harsh operating condition and high capital costs were considered obstacles to be actualized. Operating SI cycle at low-pressure is one of the solutions to actualize the cycle. The flash operation with over-azeotropic HI at low pressure does not require temperature and pressure as high as those in the existing methods as well as heating for separation. The operation in low pressure reduces corrosion problems and enables us to use flexible selection of structural material. We devised an intermediate heat exchanging-depressurizing loop to eliminate high operating pressure in the hydrogen side as well as a large pressure difference between the reactor side and the hydrogen side. Molten salts are adequate candidates as working fluids under the high-temperature condition with homogeneous phase during pressure changing process. Using molten salts, 2.20-4.65 MW of pumping work is required to change the pressure from 1 bar to 7 MPa. We selected BeF{sub 2}-containing salts as the possible candidates based on preliminary economic and thermal hydraulic consideration.

  18. Depressurization experiments on a plugged fibrous insulation in a horizontal pressure tube

    International Nuclear Information System (INIS)

    Lang, H.; Weise, H.J.; Ennen, P.

    1977-08-01

    Hot gas ducts for high-temperature reactors with a helium turbine are subject to additional operational loads not caused by the gas temperature. They include vibrations, caused by high gas velocities or by the sound fields emitted from the turbine, and stresses, originating from fast, short-time pressure changes. Such pressure changes occur as a rule if the generator coupled with the turbine has to be disconnected from the grid. In order to avoid no-load operation of the turbine a bypass between HP and LP side of the turbine is opened. As a consequence of this measure a sudden pressure drop occurs in the free flow cross-section causing differential pressures within the insulation. As the size of these differential pressures depends on the insulating material, the density of plugging, the kind of internals, and on the position and size of the depressurization borings, the pressure distributions in the insulation were measured on a test tube for the HP channel. (orig./RW) [de

  19. HANDBOOK: SUB-SLAB DEPRESSURIZATION FOR LOW PERMEABILITY FILL MATERIAL DESIGN AND INSTALLATION OF A HOME RADON REDUCTION SYSTEM

    Science.gov (United States)

    Radon, a radioactive gas, comes from the natural decay of uranium. It moves to the earth's surface through tiny openings and cracks in soil and rocks. In outdoor air, radon is diluted to such low concentrations that it is usually nothing to worry about. However, radon can accumul...

  20. Radon prevention in the design and construction of schools and other large buildings

    International Nuclear Information System (INIS)

    Leovic, K.W.; Craig, A.B.; Harris, D.B.

    1991-01-01

    The paper discusses radon prevention in the design and construction of schools and other large buildings. The U.S. EPA's Office of Research and Development (ORD) has started incorporating radon control measures into the design and construction of new schools and other large buildings. The goal of the new designs is twofold: (1) to prevent elevated radon levels in the completed building, and (2) to provide the protection at a fraction of the cost of retrofit systems. ORD's Air and Energy Engineering Research Laboratory (AEERL) has conducted the research necessary to develop viable designs. The Indoor Radon Abatement Act of 1988 set a national long-term goal for indoor radon of 0.5 picocurie per liter (pCi/L)--the same as that typically found in outdoor air. The EPA currently recommends that homeowners take action to reduce radon levels to below 4 pCi/L. To achieve the national goal set in the 1988 Act for new construction in radon-prone areas, AEERL research is using a combination of active subslab depressurization (ASD) and operation of the heating, ventilation, and air-conditioning (HVAC) system to pressurize buildings. These measures are described briefly, along with a case history of one building

  1. Update on radon-mitigation research in schools. Rept. for 1988-Aug 91

    International Nuclear Information System (INIS)

    Leovic, K.W.; Craig, A.B.; Harris, D.B.

    1991-01-01

    The paper is an overview of research by EPA's Air and Energy Engineering Research Laboratory (AEERL) on radon mitigation in 47 schools since 1988. The structural and heating, ventilating, and air-conditioning (HVAC) system characteristics of the research schools are presented, along with the mitigation techniques implemented in the schools. Research discussed includes recent and on-going projects in Colorado, Maine, Maryland, Ohio, South Dakota, Tennessee, and Virginia. Initial research focussed on the application of active subslab depressurization (ASD) to school buildings, and recent research has emphasized the ability and limitations of using HVAC systems to reduce radon levels in schools. A goal of future projects is to compare the effectiveness of the two techniques in the same building. Slab-on-grade is the most prevalent substructure in AEERL's research schools and, depending on pressure field extension, ASD systems have been recommended for radon control in many of them. In schools where they have been installed, ASD systems have performed well and are currently being evaluated for long-term performance. The distribution of HVAC system types in these schools is about a third central air handling systems, a third unit ventilators, and a third that do not supply conditioned outdoor air (i.e., fan coil units or radiant heat)

  2. Simulation of Fission Product Liftoff Behavior During Depressurization Transients

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  3. Two-phase mixture level swell and liquid entrainment/off-take in a vessel during rapid depressurization

    International Nuclear Information System (INIS)

    Kim, Chang Hyun

    2004-02-01

    An experimental study has been performed to analyze the two-phase mixture level swell and the liquid entrainment/off-take through the break in a vessel, which are important phenomena to determine the bleed capacity of the Safety Depressurization System (SDS) of Korea Advanced Power Reactor 1400 (APR1400). Three separate experiments are performed in this study: (a) the depressurization and two-phase mixture level swell experiment: (b) the two-phase mixture level measurement experiment: (c) the liquid entrainment and off-take experiment. A series of experiments has been performed using a scaled pressurized vessel in various depressurization conditions to analyze the two-phase mixture level swell and the liquid entrainment/off-take phenomena from the two-phase mixture surface in the first experiment. The test parameters are the initial pressure (10 - 38.75bars), the initial water level (43.7% - 80.0% of full height), the orifice inner diameter (10mm, 17.5mm, and 20mm). The liquid off-take takes place in certain experimental conditions. The measured parameters in the present experiments are axial void fraction distributions, pressures, temperatures in the test vessel, and the mixture density and mass flowrate through the discharge pipe. An assessment of RELAP5/MOD3 code with the present experimental data has been performed. With appropriate nodalization and time step, RELAP5/MOD3 showed reasonable agreement with the present experimental data for the gradual depressurization without liquid off-take. In the case that the off-take takes place, however, RELAP5/MOD3 under-predicts the amount of liquid entrainment/off-take during depressurization. In the second experiment, an assessment of an ultrasonic sensor and a two-wire type capacitance probe for the two-phase mixture level measurement has been performed under the same experimental conditions to adopt an appropriate measurement method for the two-phase mixture level swell and to investigate pool void fraction by the

  4. The Temperature of the Dimethylhydrazine Drops Moving in the Atmosphere after Depressurization of the Fuel Tank Rockets

    Directory of Open Access Journals (Sweden)

    Bulba Elena

    2016-01-01

    Full Text Available This work includes the results of the numerical modeling of temperature changes process of the dimethylhydrazine (DMH drops, taking into account the radial temperature gradient in the air after the depressurization of the fuel compartments rockets at high altitude. There is formulated a mathematical model describing the process of DMH drops thermal state modifying when it's moving to the Earth's surface. There is the evaluation of the influence of the characteristic size of heptyl drops on the temperature distribution. It's established that the temperatures of the small size droplets practically completely coincide with the distribution of temperature in the atmosphere at altitudes of up to 40 kilometers.

  5. Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

    International Nuclear Information System (INIS)

    Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

    1982-01-01

    The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

  6. Physical modelling of LNG rollover in a depressurized container filled with water

    Science.gov (United States)

    Maksim, Dadonau; Denissenko, Petr; Hubert, Antoine; Dembele, Siaka; Wen, Jennifer

    2015-11-01

    Stable density stratification of multi-component Liquefied Natural Gas causes it to form distinct layers, with upper layer having a higher fraction of the lighter components. Heat flux through the walls and base of the container results in buoyancy-driven convection accompanied by heat and mass transfer between the layers. The equilibration of densities of the top and bottom layers, normally caused by the preferential evaporation of Nitrogen, may induce an imbalance in the system and trigger a rapid mixing process, so-called rollover. Numerical simulation of the rollover is complicated and codes require validation. Physical modelling of the phenomenon has been performed in a water-filled depressurized vessel. Reducing gas pressure in the container to levels comparable to the hydrostatic pressure in the water column allows modelling of tens of meters industrial reservoirs using a 20 cm laboratory setup. Additionally, it allows to model superheating of the base fluid layer at temperatures close the room temperature. Flow visualizations and parametric studies are presented. Results are related to outcomes of numerical modelling.

  7. Hydro-geomechanical behaviour of gas-hydrate bearing soils during gas production through depressurization and CO2 injection

    Science.gov (United States)

    Deusner, C.; Gupta, S.; Kossel, E.; Bigalke, N.; Haeckel, M.

    2015-12-01

    Results from recent field trials suggest that natural gas could be produced from marine gas hydrate reservoirs at compatible yields and rates. It appears, from a current perspective, that gas production would essentially be based on depressurization and, when facing suitable conditions, be assisted by local thermal stimulation or gas hydrate conversion after injection of CO2-rich fluids. Both field trials, onshore in the Alaska permafrost and in the Nankai Trough offshore Japan, were accompanied by different technical issues, the most striking problems resulting from un-predicted geomechanical behaviour, sediment destabilization and catastrophic sand production. So far, there is a lack of experimental data which could help to understand relevant mechanisms and triggers for potential soil failure in gas hydrate production, to guide model development for simulation of soil behaviour in large-scale production, and to identify processes which drive or, further, mitigate sand production. We use high-pressure flow-through systems in combination with different online and in situ monitoring tools (e.g. Raman microscopy, MRI) to simulate relevant gas hydrate production scenarios. Key components for soil mechanical studies are triaxial systems with ERT (Electric resistivity tomography) and high-resolution local strain analysis. Sand production control and management is studied in a novel hollow-cylinder-type triaxial setup with a miniaturized borehole which allows fluid and particle transport at different fluid injection and flow conditions. Further, the development of a large-scale high-pressure flow-through triaxial test system equipped with μ-CT is ongoing. We will present results from high-pressure flow-through experiments on gas production through depressurization and injection of CO2-rich fluids. Experimental data are used to develop and parametrize numerical models which can simulate coupled process dynamics during gas-hydrate formation and gas production.

  8. AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

    2016-10-15

    Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

  9. RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

    International Nuclear Information System (INIS)

    Gencheva, Rositsa V.; Stefanova, Antoaneta E.; Groudev, Pavlin P.

    2005-01-01

    During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety

  10. Evaluation of the need for a rapid depressurization capability for Combustion Engineering plants

    International Nuclear Information System (INIS)

    Marsh, L.; Liang, C.

    1984-12-01

    This report documents the NRC staff evaluation of the need for providing a rapid primary system depressurization capability, in particular by using a power-operated relief valve(s) (PORVs), in the current 3410-MWt and 3800-MWt classes of plants designed by Combustion Engineering (CE). The staff reviewed the responses of licensees, applicants, and vendors to staff questions, supplemented by independent analyses by the staff and its contractors. The staff review led to the conclusion that, on the basis of risk reduction and cost/benefit considerations, no overwhelming benefit would result from requiring the installation of PORVs in CE plants that currently do not have them. However, when other unquantifiable considerations regarding the potential benefits of a PORV are factored into the evaluation, it appears that more substantial benefits could be realized. Given the more comprehensive studies currently under way to resolve the generic unresolved safety issue, USI A-45, Decay Heat Removal Reliability, the staff concludes that the decision regarding PORVs for these CE plants should be deferred and incorporated into the technical resolution of USI A-45

  11. A simple method for environmental cell depressurization for use with an electron microscope.

    Science.gov (United States)

    Ogawa, Naoki; Mizokawa, Ryo; Saito, Minoru; Ishikawa, Akira

    2017-12-01

    With the aid of the environmental cell (EC) in electron microscopy, hydrated specimens have been observed at high resolutions that optical microscopy cannot attain. Due to the ultra-high vacuum conditions of the inner column of the electron microscope, the EC requires sealing films that are sufficiently thin to allow electron transmission and that are sufficiently tough to withstand the pressure difference between the inside and outside of the EC. However, most hydrated specimens can be observed at low vacuum because the saturated vapor pressure of water is known to be 0.02 atm at room temperature. These concepts have been used in the differential pumping system, but it is complicated and relatively expensive. In this work, we propose a simple method for depressurization of the EC using a 'balloon structure' and demonstrate the theoretical benefits and practical improvement for specimen observations in low-vacuum conditions. © The Author 2017. Published by Oxford University Press on behalf of The Japanese Society of Microscopy. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  12. Radon diagnostics and mitigation in two public schools in Nashville, Tennessee

    International Nuclear Information System (INIS)

    Craig, A.B.; Leovic, K.W.; Harris, D.B.

    1990-01-01

    This paper reports on diagnostic measurements and mitigation studies that were carried out in two schools in Nashville, Tennessee, as part of the Environmental Protection Agency's (EPAs) School Radon Mitigation Development/Demonstration Program. Diagnostic studies included architectural plans and building examination, sub-slab radon concentrations, sub-slab communications measurements, and detailed classroom radon measurements using 2-day charcoal canisters, electret ion chambers, and continuous monitors. Although sub-slab communications varied significantly between the two schools, both were amenable to mitigation using sub-slab suction. Average premitigation levels of 39.5 and 29.7 picocuries per liter (pCi/L) were reduced to 0.78 and 1.7 pCi/L in the two schools

  13. Reactor containment depressurization and filtration equipment for use in the case of a serious accident

    International Nuclear Information System (INIS)

    L'Homme, A.

    1987-06-01

    A study was carried out under the aegis of the OECD into filtered vented containment systems which permit depressurization of the containment and filtration of the effluents released to the environment, in the event of a major accident with a pressurized water reactor (PWR) (or BWR or CANDU type reactors) involving core meltdown, with a view to minimizing the consequences. This paper describes the various systems examined which could possibly be used for this purpose. These comprised the French robust sand filtration system, the Swedish FILTRA system, the vacuum containment and discharge and emergency filtration system used by the CANDU plants of the Ontario-Hydro electricity company in Canada and the BWR pressure-suppression pounds. The positions of the various national authorities regarding incorporation of such systems into nuclear power plants, the design and technical principles underlying the systems, the procedures and criteria for their use and their advantages and disadvantages are examined [fr

  14. Numerical simulation of Class 3 hydrate reservoirs exploiting using horizontal well by depressurization and thermal co-stimulation

    International Nuclear Information System (INIS)

    Yang, Shengwen; Lang, Xuemei; Wang, Yanhong; Wen, Yonggang; Fan, Shuanshi

    2014-01-01

    Highlights: • Depressurization and thermal co-stimulation using horizontal well were proposed. • 3D stimulation showed that gas release rate was 3 × 10 5 m 3 per day within 450 days. • 2D stimulation showed that Class 3 hydrates could be dissociated within 8500 days. • 2D Simulation showed that heat flow was 1620 W lasting 1500 days, and decreased fast. • 1.1× 10 5 kg water was collected within 2000 days and then no more water was produced. - Abstract: Class 3 hydrate reservoirs exploiting using horizontal well by depressurization and thermal co-stimulation was simulated using the HydarteResSim code. Results showed that more than 20% of hydrates in the reservoirs had been dissociated within 450 days at the well temperature of 42 °C and well pressure of 0.1P 0 , 0.2P 0 (P 0 is the initial pressure of the reservoirs, simplifying 42 °C and 0.1P 0 , 42 °C and 0.2P 0 ). While the production behavior of 42 °C and 0.5P 0 , 42 °C and 0.8P 0 were not so exciting. In order to understand the production character of the well in long term, the cross section of 1 m length reservoirs was simulated. Simulation results showed that 4.5 × 10 5 m 3 gas would be collected within 4500 days and 1.1 × 10 6 kg water could be produced within 1500 days in the well at 42 °C and 0.1P 0 . 3.5 × 10 5 m 3 gas would be collected within 8500 days and 1.1 × 10 6 kg water could be produced within 1500 days in the well at 42 °C and 0.2P 0 . The heat flow was 1620 W at the beginning and then decreased rapidly in the two cases. For reservoirs of 1495.2 m in length, about 6.7 × 10 8 m 3 and 5.3 × 10 8 m 3 gas would be collected in the well corresponding to conditions of 42 °C and 0.1P 0 , and 42 °C and 0.2P 0

  15. The mechanisms of transitions from natural convection and nucleate boiling to nucleate boiling or film boiling caused by rapid depressurization in highly subcooled water

    International Nuclear Information System (INIS)

    Sakurai, Akira; Shiotsu, Masahiro; Hata, Koichi; Fukuda, Katsuya

    1999-01-01

    The mechanisms of transient boiling process including the transitions to nucleate boiling or film boiling from initial heat fluxes, q in , in natural convection and nucleate boiling regimes caused by exponentially decreasing system pressure with various decreasing periods, τ p on a horizontal cylinder in a pool of highly subcooled water were clarified. The transient boiling processes with different characteristics were divided into three groups for low and intermediate q in in natural convection regime, and for high q in in nucleate boiling regime. The transitions at maximum heat fluxes from low q in in natural convection regime to stable nucleate boiling regime occurred independently of the τ p values. The transitions from intermediate and high q in values in natural convection and nucleate boiling to stable film boiling occurred for short τ p values, although those to stable nucleate boiling occurred for tong τ p values. The CHF and corresponding surface superheat values at which the transition to film boiling occurred were considerably lower and higher than the steady-state values at the corresponding pressure during the depressurization respectively. It was suggested that the transitions to stable film boiling at transient critical heat fluxes from intermediate q in in natural convection and from high q in in nucleate boiling for short τ p occur due to explosive-like heterogeneous spontaneous nucleation (HSN). The photographs of typical vapor behavior due to the HSN during depressurization from natural convection regime for short τ p were shown. (author)

  16. Monitoring system of depressurization valves of migrated gas in annular space of flexible risers

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Luiz A.; Santos, Joilson M.; Carvalho, Antonio L.; Loureiro, Patricia [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

    2005-07-01

    PETROBRAS Research and Development Center - CENPES developed an automatic system for monitoring pressure of annular space due to permeation of gas in flexible risers to inspect continuously integrity of such lines. To help maintaining physical integrity of flexible risers, two PSV's are installed to end fittings on top of riser, so that operation of any valve grants the maximum admissible gas pressure within the riser annular space, as overpressure might cause damages to external polymeric layer of flexible riser. Due to the fact that there is no mechanism allowing operation to verify correct PSV performance and frequency of valve's closings and openings, we felt to be necessary the development and implement an automatic instrumented system, integrated to platform's automation and control infrastructure. The objective of this instrumentation is to monitor and register pressure of annular space in flexible riser, as well as XV's depressurization frequency. Having such information registered and monitored, can infer some riser structural conditions, anticipating repairs and preventive maintenance. In this paper we present developed system details including instruments required, application, operation of associated screens that are used in the ECOS, with events, alarms and industrial automation services required (Application development and system integration). (author)

  17. Radon entry into buildings: Effects of atmospheric pressure fluctuations and building structural factors

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Allen Lantham [Univ. of California, Berkeley, CA (United States). Dept. of Mechanical Engineering

    1996-05-01

    An improved understanding of the factors that control radon entry into buildings is needed in order to reduce the public health risks caused by exposure to indoor radon. This dissertation examines three issues associated with radon entry into buildings: (1) the influence of a subslab gravel layer and the size of the openings between the soil and the building interior on radon entry; (2) the effect of atmospheric pressure fluctuations on radon entry; and (3) the development and validation of mathematical models which simulate radon and soil-gas entry into houses. Experiments were conducted using two experimental basements to examine the influence of a subslab gravel layer on advective radon entry driven by steady indoor-outdoor pressure differences. These basement structures are identical except that in one the floor slab lies directly on native soil whereas in the other the slab lies on a high-permeability gravel layer. The measurements indicate that a high permeability subslab gravel layer increases the advective radon entry rate into the structure by as much as a factor of 30. The magnitude of the enhancement caused by the subslab gravel layer depends on the area of the openings in the structure floor; the smaller the area of these openings the larger the enhancement in the radon entry rate caused by the subslab gravel layer. A three-dimensional, finite-difference model correctly predicts the effect of a subslab gravel layer and open area configuration on advective radon entry driven by steady indoor-outdoor pressure differences; however, the model underpredicts the absolute entry rate into each structure by a factor of 1.5.

  18. Radon entry into buildings: Effects of atmospheric pressure fluctuations and building structural factors

    International Nuclear Information System (INIS)

    Robinson, A.L.

    1996-05-01

    An improved understanding of the factors that control radon entry into buildings is needed in order to reduce the public health risks caused by exposure to indoor radon. This dissertation examines three issues associated with radon entry into buildings: (1) the influence of a subslab gravel layer and the size of the openings between the soil and the building interior on radon entry; (2) the effect of atmospheric pressure fluctuations on radon entry; and (3) the development and validation of mathematical models which simulate radon and soil-gas entry into houses. Experiments were conducted using two experimental basements to examine the influence of a subslab gravel layer on advective radon entry driven by steady indoor-outdoor pressure differences. These basement structures are identical except that in one the floor slab lies directly on native soil whereas in the other the slab lies on a high-permeability gravel layer. The measurements indicate that a high permeability subslab gravel layer increases the advective radon entry rate into the structure by as much as a factor of 30. The magnitude of the enhancement caused by the subslab gravel layer depends on the area of the openings in the structure floor; the smaller the area of these openings the larger the enhancement in the radon entry rate caused by the subslab gravel layer. A three-dimensional, finite-difference model correctly predicts the effect of a subslab gravel layer and open area configuration on advective radon entry driven by steady indoor-outdoor pressure differences; however, the model underpredicts the absolute entry rate into each structure by a factor of 1.5

  19. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

  20. Characterization of liquid entrainment in the AP1000 automatic depressurization system from APEX tests

    International Nuclear Information System (INIS)

    Richard F Wright; Terry L Schulz; Jose N Reyes; John Groome

    2005-01-01

    Full text of publication follows: The AP1000 is a 1000 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 relies heavily on the 600 MWe AP600 which received design certification in 1999. A critical part of the AP600 design certification process involved the testing of the passive safety systems. A one-fourth height, one-fourth pressure test facility, APEX-600, was constructed at the Oregon State University to study design basis events, and to provide a body of data to be used to validate the computer models used to analyze the AP600. This facility was extensively modified to reflect the design changes for AP1000 including higher power in the electrically heated rods representing the reactor core, and changes in the size of the pressurizer, core makeup tanks and automatic depressurization system. The APEX-1000 test facility was used to perform design basis accident simulations and separate effects tests to support the AP1000 design certification process. In the event of a LOCA, the AP1000 passive core cooling system provides sources of core makeup water along with an automatic depressurization system (ADS) consisting of several stages of valves which reduce the reactor coolant system pressure in a controlled manner. The final stage of this system, ADS-4, consists of four large valves that open off the hot legs, reducing the pressure to allow gravity injection from the in-containment refueling water storage tank (IRWST) and eventually the containment sump. The 67% increase in power from AP600 to AP1000 results in proportionally larger steam velocities exiting the core. Higher steam velocities could increases the potential for significant liquid entrainment out the ADS-4 lines, affecting the liquid inventory in the reactor. Tests were performed in APEX-1000 to characterize the two

  1. Depressurization accident analysis of MPBR by PBRSIM with chemical reaction model

    International Nuclear Information System (INIS)

    No, Hee Cheon; Kadak, A. C.

    2002-01-01

    The simple model for natural circulation is implemented into PBR S IM to provide air inlet velocity from the containment air space. For the friction and form loss only the pebble region is considered conservatively modeling laminar flow through a packed bed. For the chemical reaction model of PBR S IM the oxidation rate is determined as the minimum value of three mechanisms estimated at each time step: oxygen mass flow rate entering the bottom of the reflector, oxidation rate by kinetics, and oxygen mass flow rate arriving at the graphite surface by diffusion. Oxygen mass flux arriving at the graphite surface by diffusion is estimated based on energy-mass analogy. Two types of exothermic chemical reaction are considered: (C + zO 2 → xCO + yCO 2 ) and (2CO + O 2 2CO 2 ). The heterogeneous and homogeneous chemical reaction rates by kinetics are determined by INEEL and Bruno correlations, respectively. The instantaneous depressurization accident of MPBR is simulated using PBR S IM with chemical model. The air inlet velocity is initially rapidly dropped within 10 hr and reaches a saturation value of about 1.5cm/s. The oxidation rate by the diffusion process becomes lower than that by the chemical kinetics above 600K. The maximum pebble bed temperatures without and with chemical reaction reach the peak values of 1560 and 1617 .deg. C at 80 hr and 92 hr, respectively. As the averaged temperatures in the bottom reflector and the pebble bed regions increase with time, (C+1/2O2 ->CO) reaction becomes dominant over (C+O 2 →CO 2 ) reaction. Also, the CO generated by (C+1/2O 2 →CO) reaction will be consumed by (2CO+O 2 →2CO 2 ) reaction and the energy homogeneously generated by this CO depletion reaction becomes dominant over the heterogeneous reaction

  2. CNTB program for the analysis of partially mixed containment atmospheres during depressurization events

    International Nuclear Information System (INIS)

    Landoni, J.A.

    1979-07-01

    This program describes the analytical models for the CNTB computer program, which is permanently filed in the archive library of the General Atomic (GA) San Diego Data Center under reference number THSD-2699. Developed during the last four years, this computer program has been successfully applied in its presented form to the type of containment atmosphere transients illustrated in this report. For example, the CNTB computer program is applicable (1) to the design basis depressurization accident (DBDA) to determine the effect of the partial mixing on the containment atmospheric peak pressure (known as nonmixing penalty) and (2) for Class 9 accidents, such as the loss of forced circulation (LOFC), for the AIPA Phase I studies. The capability of the CNTB computer program has been substantially improved over its precursor, the CONTEMPT-G computer program, to predict the thermodynamic behavior of the containment atmosphere during helium releases, assuming partial mixing of the original air with the effluent and to predict the amount of the environmental leaks under closed and open containment conditions. In addition, the CNTB computer program running times are considerably below the ones required for the CONTEMPT-G computer program. Computational solution of the variable parameters in the containment atmosphere is effected by an iterative technique, while the temperatures for its boundaries are obtained by finite differences. The CNTB computer program, written in FORTRAN V, has been implemented at GA on the UNIVAC 1110 computer

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  4. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  5. Radon mitigation experience in houses with basements and adjoining crawl spaces

    International Nuclear Information System (INIS)

    Messing, M.; Henschel, D.B.

    1990-01-01

    Active soil depressurization systems were installed in four basement houses with adjoining crawl spaces in Maryland. In addition, existing soil depressurization systems were modified in two additional basement-plus-crawl-space houses. These six houses were selected to include both good and poor communication beneath the basement slab, and different degrees of importance of the crawl space as a source of the indoor radon. The radon reduction effectiveness was compared for: depressurization only under the basement slab; depressurization only under a polyethylene liner over the unpaved crawl-space floor; and simultaneous depressurization under both the basement slab and the crawl-space liner. The objective of this paper is to identify under what conditions treatment of the basement alone might provide sufficient radon reductions in houses of this substructure, and what incremental benefits might be achieved by also treating the crawl space

  6. Pipe stress intensity factors and coupled depressurization and dynamic crack propagation. 1976 Annual report

    International Nuclear Information System (INIS)

    Emery, A.F.; Kobayashi, A.S.; Love, W.J.

    1978-04-01

    This report contains the description of predictive models for the initiation and propagation of cracks in pipes and the numerical results obtained. The initiation of the crack was studied by evaluating stress intensity factors under static conditions for a series of representative flaws. Three-dimensional static stress intensity factors were determined for quarter-elliptical cracks at the corner of a hole in an infinite plate and at the corner of a bore in a rotating disk. Semi-elliptical cracks for plates in bending and in pressurized and thermally stressed hollow cylinders were also evaluated. The stress fields, in the absence of a crack, were used in the ''alternating technique'' to compute the stress intensity factors along the crack front. Parametric studies were made to assess the effects of crack thickness, the ratio of the major and minor axes of the ellipse and the thickness of the cylinders or plates. These parametric results may be used to predict critical flaw sizes for the initiation of the running crack. The initiation and propagation of axial through cracks in pressurized pipes was studied by using an elastic-plastic finite different shell code coupled with a one-dimensional thermal-hydraulic code which computed the leakage through the crack opening and the depressurization of the fluid in the pipe. The effects of large deflections and different fluid pressure profiles were investigated. The results showed that the crack opening shape is dependent upon the fracture criterion used and upon the average pressure on the crack flaps, but not upon the specific pressure profile. The consideration of large deflections changed the opening size of the crack and through the coupling with the pipe pressures, strongly affected the crack tip speed. However, for equal crack lengths, there was little difference between calculations made for large and small deflection

  7. Impact of ventilation systems and energy savings in a building on the mechanisms governing the indoor radon activity concentration.

    Science.gov (United States)

    Collignan, Bernard; Powaga, Emilie

    2017-11-23

    For a given radon potential in the ground and a given building, the parameters affecting the indoor radon activity concentration (IRnAC) are indoor depressurization of a building and its air change rate. These parameters depend mainly on the building characteristics, such as airtightness, and on the nature and performances of the ventilation system. This study involves a numerical sensitivity assessment of the indoor environmental conditions on the IRnAC in buildings. A numerical ventilation model has been adapted to take into account the effects of variations in the indoor environmental conditions (depressurization and air change rate) on the radon entry rate and on the IRnAC. In the context of the development of a policy to reduce energy consumption in a building, the results obtained showed that IRnAC could be strongly affected by variations in the air permeability of the building associated with the ventilation regime. Copyright © 2017 Elsevier Ltd. All rights reserved.

  8. Phytoforensics: Trees as bioindicators of potential indoor exposure via vapor intrusion.

    Directory of Open Access Journals (Sweden)

    Jordan L Wilson

    Full Text Available Human exposure to volatile organic compounds (VOCs via vapor intrusion (VI is an emerging public health concern with notable detrimental impacts on public health. Phytoforensics, plant sampling to semi-quantitatively delineate subsurface contamination, provides a potential non-invasive screening approach to detect VI potential, and plant sampling is effective and also time- and cost-efficient. Existing VI assessment methods are time- and resource-intensive, invasive, and require access into residential and commercial buildings to drill holes through basement slabs to install sampling ports or require substantial equipment to install groundwater or soil vapor sampling outside the home. Tree-core samples collected in 2 days at the PCE Southeast Contamination Site in York, Nebraska were analyzed for tetrachloroethene (PCE and results demonstrated positive correlations with groundwater, soil, soil-gas, sub-slab, and indoor-air samples collected over a 2-year period. Because tree-core samples were not collocated with other samples, interpolated surfaces of PCE concentrations were estimated so that comparisons could be made between pairs of data. Results indicate moderate to high correlation with average indoor-air and sub-slab PCE concentrations over long periods of time (months to years to an interpolated tree-core PCE concentration surface, with Spearman's correlation coefficients (ρ ranging from 0.31 to 0.53 that are comparable to the pairwise correlation between sub-slab and indoor-air PCE concentrations (ρ = 0.55, n = 89. Strong correlations between soil-gas, sub-slab, and indoor-air PCE concentrations and an interpolated tree-core PCE concentration surface indicate that trees are valid indicators of potential VI and human exposure to subsurface environment pollutants. The rapid and non-invasive nature of tree sampling are notable advantages: even with less than 60 trees in the vicinity of the source area, roughly 12 hours of tree

  9. Phytoforensics: Trees as bioindicators of potential indoor exposure via vapor intrusion.

    Science.gov (United States)

    Wilson, Jordan L; Samaranayake, V A; Limmer, Matt A; Burken, Joel G

    2018-01-01

    Human exposure to volatile organic compounds (VOCs) via vapor intrusion (VI) is an emerging public health concern with notable detrimental impacts on public health. Phytoforensics, plant sampling to semi-quantitatively delineate subsurface contamination, provides a potential non-invasive screening approach to detect VI potential, and plant sampling is effective and also time- and cost-efficient. Existing VI assessment methods are time- and resource-intensive, invasive, and require access into residential and commercial buildings to drill holes through basement slabs to install sampling ports or require substantial equipment to install groundwater or soil vapor sampling outside the home. Tree-core samples collected in 2 days at the PCE Southeast Contamination Site in York, Nebraska were analyzed for tetrachloroethene (PCE) and results demonstrated positive correlations with groundwater, soil, soil-gas, sub-slab, and indoor-air samples collected over a 2-year period. Because tree-core samples were not collocated with other samples, interpolated surfaces of PCE concentrations were estimated so that comparisons could be made between pairs of data. Results indicate moderate to high correlation with average indoor-air and sub-slab PCE concentrations over long periods of time (months to years) to an interpolated tree-core PCE concentration surface, with Spearman's correlation coefficients (ρ) ranging from 0.31 to 0.53 that are comparable to the pairwise correlation between sub-slab and indoor-air PCE concentrations (ρ = 0.55, n = 89). Strong correlations between soil-gas, sub-slab, and indoor-air PCE concentrations and an interpolated tree-core PCE concentration surface indicate that trees are valid indicators of potential VI and human exposure to subsurface environment pollutants. The rapid and non-invasive nature of tree sampling are notable advantages: even with less than 60 trees in the vicinity of the source area, roughly 12 hours of tree-core sampling with minimal

  10. Phytoforensics: Trees as bioindicators of potential indoor exposure via vapor intrusion

    Science.gov (United States)

    Wilson, Jordan L.; Samaranayake, V.A.; Limmer, Matthew A.; Burken, Joel G.

    2018-01-01

    Human exposure to volatile organic compounds (VOCs) via vapor intrusion (VI) is an emerging public health concern with notable detrimental impacts on public health. Phytoforensics, plant sampling to semi-quantitatively delineate subsurface contamination, provides a potential non-invasive screening approach to detect VI potential, and plant sampling is effective and also time- and cost-efficient. Existing VI assessment methods are time- and resource-intensive, invasive, and require access into residential and commercial buildings to drill holes through basement slabs to install sampling ports or require substantial equipment to install groundwater or soil vapor sampling outside the home. Tree-core samples collected in 2 days at the PCE Southeast Contamination Site in York, Nebraska were analyzed for tetrachloroethene (PCE) and results demonstrated positive correlations with groundwater, soil, soil-gas, sub-slab, and indoor-air samples collected over a 2-year period. Because tree-core samples were not collocated with other samples, interpolated surfaces of PCE concentrations were estimated so that comparisons could be made between pairs of data. Results indicate moderate to high correlation with average indoor-air and sub-slab PCE concentrations over long periods of time (months to years) to an interpolated tree-core PCE concentration surface, with Spearman’s correlation coefficients (ρ) ranging from 0.31 to 0.53 that are comparable to the pairwise correlation between sub-slab and indoor-air PCE concentrations (ρ = 0.55, n = 89). Strong correlations between soil-gas, sub-slab, and indoor-air PCE concentrations and an interpolated tree-core PCE concentration surface indicate that trees are valid indicators of potential VI and human exposure to subsurface environment pollutants. The rapid and non-invasive nature of tree sampling are notable advantages: even with less than 60 trees in the vicinity of the source area, roughly 12 hours of tree-core sampling with

  11. Nuclear reactor with makeup water assist from residual heat removal system

    Science.gov (United States)

    Corletti, Michael M.; Schulz, Terry L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  12. Nuclear reactor with makeup water assist from residual heat removal system

    International Nuclear Information System (INIS)

    Corletti, M.M.; Schulz, T.L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures

  13. Nuclear reactor with makeup water assist from residual heat removal system

    Science.gov (United States)

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  14. Derivation of criteria for primary circuit activity in an HTGR

    International Nuclear Information System (INIS)

    Su, S.D.; Barsell, A.W.

    1980-11-01

    This paper derives specific criteria for the circulating and plateout activity in the primary circuit for a 2170-MW(t) high temperature gas-cooled reactor-gas turbine (HTGR-GT) plant. Results show that for a design basis, (1) the circulating activity should be limited to 14,000 Ci Kr-88 (a principal nuclide) to meet both offsite dose and containment access constraint during normal operation and depressurization accidents, and (2) the plateout inventories for those important nuclides affecting shutdown maintenance should not exceed 10,000 Ci Ag-110m, 45,000 Ci Cs-134 and 130,000 Ci Cs-137. This paper presents bases and methodology for deriving such criteria and compares them with light water reactors. 5 tables

  15. Apparatus and method for depressurizing, degassing, and affording decay of the radioactivity of weakly radioactive condensates in nuclear power plants

    International Nuclear Information System (INIS)

    Gross, R.; Plotz, J.

    1976-01-01

    Described is an apparatus for depressurizing, degassing and affording decay of weakly radioactive condensates in nuclear power plants having a turbine and a main condenser turbine wherein exhaust steam of the turbine is condensed and forms a main condensate, and includes a collecting tank for the condensate situated below the condenser. A plurality of horizontal degassing channels, each having a lateral overflow, are disposed in the upper part of the condensate collecting tank and are filled with the main condensate up to the level of the overflow. At least one feedwater preheater which is heated by bleeder steam from the turbine provides a secondary condensate. Below the overflow height of the degassing channels extend horizontal feed pipes for the secondary condensate. The feed pipes are connected to the output of pressure relieving expanding devices and are provided on their underside with discharge openings for the bubbling of the secondary condensate into the main condensate to thereby degass the main condensate. The condensate collecting tank has mutually offset partitions therein providing an adequately long path for the decay of the main and secondary condensates. The condensate which is discharged from the condensate collecting tank is returned into the cycle as feedwater. Also disclosed is a method of operating the foregoing apparatus

  16. Nuclear reactor with makeup water assist from residual heat removal system

    International Nuclear Information System (INIS)

    Schulz, T.L.; Corletti, M.M.

    1994-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit by pumping water from an in-containment refueling water storage tank during staged depressurization of the coolant circuit, the final stage including passive emergency cooling by gravity feed from the refueling water storage tank to the coolant circuit and to flood the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and avoids the final stage of depressurization with its flooding of the containment when such action is not necessary, but does not prevent the final stage when it is necessary. A high pressure makeup water storage tank coupled to the reactor coolant circuit holds makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal system can also be coupled in a loop with the refueling water supply tanks for cooling the tank. (Author)

  17. Testing of indoor radon reduction techniques in eastern Pennsylvania: An update

    International Nuclear Information System (INIS)

    Henschel, D.B.; Scott, A.G.

    1987-01-01

    EPA has installed radon reduction measures in 38 houses in the Reading Prong region of eastern Pennsylvania. All were basement houses with hollow block or poured concrete foundation walls. The reduction approaches tested in most houses involved active soil ventilation, including: suction on the footing drain tile system; suction underneath the concrete slabs, using pipes inserted through the slabs from inside the house; and ventilation of the void network inside hollow block foundation walls. Heat recovery ventilators (HRVs) were tested in three houses. The current results confirm that, for the houses tested here, drain tile suction appears consistently able to provide high radon reductions when a complete loop of drain tile exists, often reducing high-radon houses to 4 pCi/l (148 B1/m 3 ) and less. Sub-slab suction (with pipes through the slab) can also provide high reductions if a sufficient number of suction pipes are located properly. Placement of one or more sub-slab suction pipes near each perimeter wall appears in this testing to aid in treating the major soil gas entry routes, although fewer pipes can sometimes give high reductions if conditions are favorable. For effective radon reduction using any active soil ventilation technique, it is important that major wall and slab openings be closed, and that a fan be employed that is capable of developing adequate static pressure

  18. The establishment and analysis of TRACE model for ultimate response guideline of Chinshan nuclear power plant - 15448

    International Nuclear Information System (INIS)

    Huang, J.J.; Wang, J.R.; Shih, C.; Chen, S.W.; Liao, L.Y.; Lin, H.T.

    2015-01-01

    The purpose of this research is to use TRACE code to perform a simulation that executes the procedures of URG (Ultimate Response Guidelines) to deal with Fukushima-like accidents. TRACE is an advanced thermal hydraulic code that has been developed by the United States Nuclear Regulatory Commission for NPP safety analysis. In this work TRACE has been used to analyze the thermal hydraulic model for the URG of the Chinshan nuclear power plant that is composed of 2 BWR-type reactors. URG includes 2-stage depressurization, alternative water injection and removing decay heat through the ejection from containment. The 2-stage depressurization strategy includes controlled depressurization and emergency depressurization to replace traditional one-stage depressurization. Results show that by comparing with one-stage depressurization strategy, 2-stage depressurization strategy is able to reduce peak cladding temperature (PCT) effectively and needs much less minimum flow rate of alternative water injection in the accident

  19. Emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Nobuaki.

    1993-01-01

    A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

  20. Failure of preventive measures against radon penetration from the ground in a new-built family house - a case study

    International Nuclear Information System (INIS)

    Neznal, M.; Neznal, M.; Jiranek, M.; Fronka, A.

    2006-01-01

    structure in contact with soil, i.e. even under walls. Special attention should be devoted to the sealing of joints and pipe penetrations through the insulation. The high-quality insulation may be replaced by a common damp-proofing if some special conditions are met: the house is built with a cellar under the complete house area; no residential rooms are found in the cellar; all year reliable natural ventilation of the cellar is provided; the cellar entrance from the floors above is provided with an automatic closing door system and with door sealing. The above mentioned approach is considered to be sufficient even in cases when the building site is classified close to the lower limit of the high radon index (the radon concentration in soil does not exceed twice the concentration that separates the medium and high radon index). In all other cases, the radon-proof insulation in all structures in direct contact with soil must be completed with either a a sub-slab ventilation system or an air gap ventilation under the insulation. The sub-slab ventilation system should reduce the radon concentration under the foundation plate, or create negative pressure in subsoil compared to indoor air pressure. It is a system of perforated drainage pipes that are inserted into a gravel layer under the foundation plate. To ensure effective operation of sub-slab ventilation, it is recommended to extract the soil air from the sub-slab region by the vertical exhaust pipes. The ventilation system can operate in two ways: a passive ventilation, i.e. the ventilation system is controlled by temperature and pressure indoor/sub-floor differences, or an active one, using a fan. Because the fan is usually installed on the vertical exhaust pipes, each passive system may be easily transformed into an active one. A failure of preventive protective measures in a new-built family house will be described and analysed in the paper. Detailed measurements of sub-slab parameters, continual monitoring of indoor

  1. Liquid entrainment and off-take at the top of the pressurizer in the case of the actuation of safety depressurization system of APR1400

    International Nuclear Information System (INIS)

    Kim, Chang Hyun; No, Hee Cheon

    2003-01-01

    In order to determine the bleed capacity of Safety Depressurization System (SDS) of Advanced Power Reactor 1400 (APR1400) in case of Total Loss of Feed Water (TLOFW), we performed an experimental study of liquid entrainment and liquid off-take from the swelled two-phase mixture surface in a vessel. A total of 208 experimental data on the entrainment and off-take are obtained using a test vessel with the height of 2.0m and the inner diameter of 0.3m having a top break with diameter of 0.05m. Two-phase mixture levels are measured by the ultrasonic sensor within ? .77% with respect to the visual level data. Droplet entrainments are measured and compared with the existing pool entrainment data. The empirical correlation for the onset of off-take is developed in terms of the Froude number (Fr g ) at the break and non-dimensional inception height (h b /d). This correlation shows agreement with the present experimental data within ? 5%. The present off-take quality data is in agreement with Schrock's off-take quality correlation with the r.m.s. error of 15.8%. In the present experiment, droplet entrainment E fg strongly depends upon jg * /h * and is proportional to the 7 th power of jg * /h * in the same way as the off-take data

  2. Radon reduction using sub floor fans

    International Nuclear Information System (INIS)

    Harley, N.H.; Chittaporn, P.

    1996-01-01

    The basement and second floor 222 Rn concentrations in an energy efficient home were measured hourly for 6 y using continuous monitors of our design. The home had a subslab pipe network installed during construction, and for the past 2 y a 150 cfm fan was operative venting air via ductwork inside the chimney exiting on the roof. During this measurement interval, experiments were conducted with the fan in 3 modes: (1) with the subslab fan off, (2) in the conventional direction auctioning air from beneath the slab to outside, and (3) reversed, blowing outdoor air into the network under the slab. We have a large data base to show that the indoor 222 R n concentration varies inversely with the indoor/outdoor temperature difference. In order to compare the 3 fan modes directly, we selected 50 to 90 d periods when the outdoor temperature was essentially the same. For the 3 modes, the fan off, blowing upward, and blowing downward, the basement concentration averaged 80, 38, and 34 Bq m -3 , respectively. Radon peaks or surges occur over a period of about 1 d during falling barometric pressure. With the fan blowing downward, these 222 Rn peaks tend to be smaller but only marginally so. We conclude that in this home the reduction in 222 Rn with the fan and subslab pipe network operating was essentially the same regardless of the direction of flow from the fan

  3. Laboratory Evaluation of Energy Recovery Ventilators

    Energy Technology Data Exchange (ETDEWEB)

    Kosar, D. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-11-01

    Over the years, building scientists have characterized the relationship between building airtightness, exhaust-only appliances airflows, and building depressurization. Now, as the use of deep retrofit measures and new construction practices is growing to realize lower infiltration levels in increasingly tighter envelopes, performance issues can arise with the operation of exhaust-only appliances in a depressurized home. As the depressurization levels climb in tighter homes, many of these exhaust-only appliances see their rated airflows reduced and other related performance issues arise as a result. If sufficiently depressurized, atmospherically vented combustion appliances that may be present in the home can backdraft as well. Furthermore, when exhaust-only appliances operate and the tight home becomes depressurized, water vapor intrusion from outdoors can raise additional issues of mold in the building envelope in more humid climates.

  4. Novel Radon Sub-Slab Suctioning System

    DEFF Research Database (Denmark)

    Rasmussen, Torben Valdbjørn

    2013-01-01

    A new principle for radon protection is currently presented which makes use of a system of horizontal pressurised air ducts located within the lower part of the rigid insulation layer of the ground-floor slab. The function of this system is based on the principles of pressure reduction within...... a grid of horizontal air ducts with low pressure which are able to remove air and radon from the ground. Results showed the system to be effective in preventing radon infiltrating from the ground through the ground-floor slab, avoiding high concentrations of radon being accumulated inside houses....... For the system to be effective, the pressure within the ducts must be lower than the pressure inside the house. The new principle was shown to be effective in preventing radon from polluting the indoor air by introducing low pressure in the horizontal grid of air ducts. A lower pressure than the pressure inside...

  5. The application of release models to the interpretation of rare gas coolant activities

    International Nuclear Information System (INIS)

    Wise, C.

    1985-01-01

    Much research is carried out into the release of fission products from UO 2 fuel and from failed pins. A significant application of this data is to define models of release which can be used to interpret measured coolant activities of rare gas isotopes. Such interpretation is necessary to extract operationally relevant parameters, such as the number and size of failures in the core and the 131 I that might be released during depressurization faults. The latter figure forms part of the safety case for all operating CAGRs. This paper describes and justifies the models which are used in the ANAGRAM program to interpret CAGR coolant activities, highlighting any remaining uncertainties. The various methods by which the program can extract relevant information from the measurements are outlined, and examples are given of the analysis of coolant data. These analyses point to a generally well understood picture of fission gas release from low temperature failures. Areas of higher temperature release are identified where further research would be beneficial to coolant activity analysis. (author)

  6. Experiments on pollutant transport from soil into residential basements by pressure-driven airflow

    International Nuclear Information System (INIS)

    Nazaroff, W.W.; Lewis, S.R.; Doyle, S.M.; Moed, B.A.; Nero, A.V.

    1987-01-01

    At two residences in Portland, OR, they have investigated (1) the coupling between residential basements and the air in nearby soil and (2) the influence of basement depressurization on the migration of air in soil. With the basements depressurized 25-50 Pa relative to outdoor air, underpressures as great as 20-40% of those in the basement were observed at sampling points in the soil. Sulfur hexafluoride was injected into the soil near the houses and its concentration monitored in soil air and in the house over time, both with and without basement depressurization. Depressurization was seen to have a substantial effect on the migration of the tracer within the soil. For basement depressurizations of 25-50 Pa, effective transport velocities through the soil and into the houses were observed to exceed 1 m h -1 . Airborne 222 Rn concentration was monitored in the basement of one house during the 6-day investigation and was seen to increase substantially on each of the seven occasions that the house was depressurized. The techniques employed are applicable to the study of problems of excessive radon entry into buildings and the migration of toxic vapors from waste dumps and landfills

  7. A study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention. ROSA-V/LSTF test SB-PV-05

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-11-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system. (author)

  8. Radon reduction in homes constructed on saprolite in the Central Appalachians

    International Nuclear Information System (INIS)

    Mose, D.G.; Mushsrush, G.W.; Slone, J.E.

    1997-01-01

    A 3-year study of indoor radon in more than 1000 homes in northern Virginia and southern Maryland was conducted using 3-month exposure alpha-track monitors. In a study set of 200 homes, first-floor indoor radon concentrations, which most closely approximates home exposure levels, averaged slightly more than 3 pCi/l. In a study set of 100 homes, sub-slab ventilation was used to reduce indoor radon concentrations. Interest in remediation was related to public perception of the hazardous nature of radon; people living in homes with indoor radon measurements of more than 4 pCi/l were more likely to participate in the remediation phase of the project. Sub-slab ventilation was successful in more than 90% of the homes in reducing indoor radon from concentrations as high as 30 pCi/ to less than 4 pCi/l, at least for the entire year of post-remediation radon measurements. (orig.)

  9. POSSIBLE ROLE OF INDOOR RADON REDUCTION SYSTEMS IN BACK-DRAFTING RESIDENTIAL COMBUSTION APPLIANCES

    Science.gov (United States)

    The article gives results of a computational sensitivity analysis conducted to identify conditions under which residential active soil depressurization (ASD) systems for indoor radon reduction might contribute to or create back-drafting of natural draft combustion appliances. Par...

  10. Depressurization of a spread of Brazil-Bolivia gas pipeline and the emergency repair of a weld crack in a instrument derivation at Campo Grande compression station; Despressurizacao de trecho do gasoduto Bolivia-Brazil para reparo emergencial de trinca em uma derivacao de instrumento de temperatura na Estacao de Compressao de Campo Grande - MS

    Energy Technology Data Exchange (ETDEWEB)

    Dietrich, Carlos Ribeiro; Leite Junior, Ismael Casano [TBG - Transportadora Brasileira Gasoduto Bolivia Brasil S.A., Rio de Janeiro, RJ (Brazil)

    2005-07-01

    The purpose of this paper is to report the actions taken to repair a gas leak, at an original pressure of 100 kgf/cm{sup 2}, occurred due to a 1 1/2'' branch pipe weld crack, located on the 24'' Campo Grande - Mato Grosso do Sul Compression Station discharge pipe. This branch pipe was used to a thermo well installation and was submitted to an additional strength caused by thermo well vibration. The weld repair actions required an urgent depressurization of a 33 km spread of Bolivia-Brazil Pipeline in a timely manner, to avoid any negative impact in the operational schedule. (author)

  11. Update on the SDTP Sponsored Fukushima Daiichi Related Assesment Activities

    International Nuclear Information System (INIS)

    Allison, Chris

    2014-01-01

    Following the accident at Fukushima Daiichi, Innovative Systems Software and several other members of an international software development and training group (SDTP) started an assessment of the possible core/vessel damage states of Units 1-3. This assessment, using a reference RELAP/SCDAPSIM Laguna Verde BWR model developed by CNSNS, the Mexican nuclear regulatory body, was presented initially to the IAEA emergency response team for Fukushima in March of 2011. Our assessment for the IAEA indicated that significant fuel melting, fuel slumping, and lower head failure was likely for Units 1 and 3. The results for Unit 2 were inconclusive because of the complex thermal hydraulic conditions at the time of likely fuel melting. Since that time the SDTP related assessment activities have continued on three main fronts: (a) continued analysis using our representative Laguna Verde model to determine the likely failure modes leading to an un-intentional depressurization of the vessel during a SBO in a BWR, (b) development of improved RELAP/SCDAPSIM models to treat the likely mode of lower core support structure melting and failure, and (c) design studies for proposed fuel melting and relocation experiments in Japan to support model development and cleanup related activities, The presentation gives a brief summary and discussion of these activities.

  12. Data report of ROSA/LSTF experiment SB-HL-12. 1% hot leg break LOCA with SG depressurization and gas inflow

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2016-01-01

    An experiment SB-HL-12 was conducted on February 24, 1998 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-HL-12 simulated a 1% hot leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). Steam generator (SG) secondary-side depressurization by fully opening the relief valves in both SGs as an accident management (AM) action was initiated immediately after maximum surface temperature of simulated fuel rod reached 600 K. Auxiliary feedwater injection into the secondary-side of both SGs was started immediately after the initiation of AM action. After the onset of AM action due to first core uncovery by core boil-off, the primary pressure decreased following the SG secondary-side pressure, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before loop seal clearing (LSC) induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after the LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after the ACC tanks started to discharge nitrogen gas, which resulted in no actuation of LPI system of ECCS during the experiment. Third core uncovery by core boil-off occurred during the reflux condensation in the SG U-tubes under nitrogen gas inflow. The core power was automatically decreased by the LSTF core protection system when the maximum fuel rod surface temperature exceeded 908 K. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal

  13. PHYSICAL AND NUMERICAL MODELING OF ASD EXHAUST DISPERSION AROUND HOUSES

    Science.gov (United States)

    The report discusses the use of a wind tunnel to physically model the dispersion of exhaust plumes from active soil depressurization (ASD) radon mitigation systems in houses. he testing studied the effects of exhaust location (grade level vs. above the eave), as house height, roo...

  14. Supercritical impregnation of cinnamaldehyde into polylactic acid as a route to develop antibacterial food packaging materials.

    Science.gov (United States)

    Villegas, Carolina; Torres, Alejandra; Rios, Mauricio; Rojas, Adrián; Romero, Julio; de Dicastillo, Carol López; Valenzuela, Ximena; Galotto, María José; Guarda, Abel

    2017-09-01

    Supercritical impregnation was used to incorporate a natural compound with antibacterial activity into biopolymer-based films to develop active food packaging materials. Impregnation tests were carried out under two pressure conditions (9 and 12MPa), and three depressurization rates (0.1, 1 and 10MPamin -1 ) in a high-pressure cell at a constant temperature equal to 40°C. Cinnamaldehyde (Ci), a natural compound with proven antimicrobial activity, was successfully incorporated into poly(lactic acid) films (PLA) using supercritical carbon dioxide (scCO 2 ), with impregnation yields ranging from 8 to 13% w/w. Higher pressure and slower depressurization rate seem to favor the Ci impregnation. The incorporation of Ci improved thermal, structural and mechanical properties of the PLA films. Impregnated films were more flexible, less brittle and more resistant materials than neat PLA films. The tested samples showed strong antibacterial activity against the selected microorganisms. In summary, this study provides an innovative route to the development of antibacterial biodegradable materials, which could be used in a wide range of applications of active food packaging. Copyright © 2017 Elsevier Ltd. All rights reserved.

  15. Impregnation of cinnamaldehyde into cassava starch biocomposite films using supercritical fluid technology for the development of food active packaging.

    Science.gov (United States)

    de Souza, Ana Cristina; Dias, Ana M A; Sousa, Hermínio C; Tadini, Carmen C

    2014-02-15

    In this work, supercritical solvent impregnation (SSI) has been tested for the incorporation of natural compounds into biocomposite materials for food packaging. Cinnamaldehyde, with proved antimicrobial activity against fungi commonly found in bread products, was successfully impregnated on biocomposite cassava starch based materials using supercritical carbon dioxide as solvent. Different process experimental conditions were tested (pressure, impregnation time and depressurization rate) at a fixed temperature (35 °C) in order to study their influence on the amount of impregnated cinnamaldehyde as well as on the morphology of the films. Results showed that all conditions permitted to impregnate antimicrobial active amounts superior to those previously obtained using conventional incorporation methods. Moreover, a significant decrease of the equilibrium water vapor sorption capacity and water vapor permeability of the films was observed after SSI processing which is a clear advantage of the process, considering the envisaged applications. Copyright © 2013 Elsevier Ltd. All rights reserved.

  16. Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae K., E-mail: taekyu8@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Arcilesi, David J., E-mail: arcilesi.1@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Kim, In H., E-mail: ihkim0730@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Sun, Xiaodong, E-mail: sun.200@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Christensen, Richard N., E-mail: rchristensen@uidaho.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Oh, Chang H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Kim, Eung S., E-mail: kes7741@snu.ac.kr [Idaho National Laboratory, Idaho Falls, ID 83402 (United States)

    2016-04-15

    Highlights: • Uncertainty quantification and benchmark study are performed to validate an ANSYS FLUENT computer model for a depressurization process in a high-temperature gas-cooled reactor. • An ANSYS FLUENT computer model of a 1/8th scaled-down geometry of a VHTR hot exit plenum is presented, which is similar to the experimental test facility that has been constructed at The Ohio State University. • Using the computer model of the scaled-down geometry, the effects of the depressurization process and flow oscillations on the subsequent density-driven stratified flow phenomenology are examined computationally. • The effects of the scaled-down hot exit plenum internal structure temperature on the density-driven stratified flow phenomenology are investigated numerically. - Abstract: An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment. Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study

  17. Emergency operating procedures improvement based on the lesson learned from the Fukushima Daiichi accident

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Wen-Hsiung, E-mail: whwu1127@aec.gov.tw [Atomic Energy Council, 2F., No. 80, Sec.1, Chenggong Rd., Yonghe Dist., New Taipei City 234, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Guangfu Rd., Hsinchu City 300, Taiwan (China); Liao, Lih-Yih, E-mail: lyliao@iner.gov.tw [Institute of Nuclear Energy Research, Atomic Energy Council, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 325, Taiwan (China)

    2016-12-01

    Highlights: • Discuss the problem of EOPs at the time of Fukushima accident to deal with the prolonged SBO. • Elaborate the potential risk accompanied with the emergency depressurization in the SBO. • Describe a special guideline to cope with Fukushima-like accidents and provide its technical basis. • Point out that Fukushima accident might have been prevented if improved EOPs had been used. • Propose key points and suggestions for improving the EOPs. - Abstract: One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima

  18. Emergency operating procedures improvement based on the lesson learned from the Fukushima Daiichi accident

    International Nuclear Information System (INIS)

    Wu, Wen-Hsiung; Liao, Lih-Yih

    2016-01-01

    Highlights: • Discuss the problem of EOPs at the time of Fukushima accident to deal with the prolonged SBO. • Elaborate the potential risk accompanied with the emergency depressurization in the SBO. • Describe a special guideline to cope with Fukushima-like accidents and provide its technical basis. • Point out that Fukushima accident might have been prevented if improved EOPs had been used. • Propose key points and suggestions for improving the EOPs. - Abstract: One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima

  19. Concept Study for Military Port Design Using Natural Processes.

    Science.gov (United States)

    1982-06-15

    concept, which he called pressure retarded osmosis. In his system "the volume-enhanced brine would be subsequently depressurized through a hydroturbine ...pressure gradient, i.e. the flux is "uphill". The subsequent depressurization of the permeate through a hydroturbine -generator set would produce

  20. ECCS control circuit

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1986-01-01

    Purpose: To afford a sufficient margin to pressure vibrations upon starting of an automatic depressurization system by dispersing pressure vibration in suppression water due to the opening action of an automatic releaf valve in the automatic depressurization system thereby reducing the dynamic load exerted to the surface of the suppression walls. Constitution: Upon occurrence of loss of coolant accidents, an automatic releaf valve for automatic depressurization is opened to deliver the steams in the pressure vessel into the suppression pool. Since a plurality of automatic releaf valves have usually been disposed, if they are opened simultaneously, excess dynamic loads are exerted due to the pressure vibrations to the wall surface of the suppression pool. In this invention, a control circuit is disposed such that the opening timing for each of the automatic releaf valves is deviated upon occurrence of a driving signal for the automatic depressurization system to thereby disperse the pressure vibrations in the suppression water. (Kamimura, M.)

  1. Inactivation of Staphylococcus aureus in raw salmon with supercritical CO2 using experimental design

    Directory of Open Access Journals (Sweden)

    Mônica CUPPINI

    2016-01-01

    Full Text Available Abstract Considering the microbial safety of consumption of raw foods (Asian food, this study aimed to explore the inactivation S. aureus in raw salmon by supercritical CO2 treatment (SC-CO2. For this purpose, experimental design methodology was employed as a tool to evaluate the effects of pressure (120-220 bar, the depressurization rate (10 to 100 bar.min–1 and the salmon:CO2 mass relation (1:0.2 to 1:1.0. It was observed that the pressure and the depressurization rate was statistically significant, i.e. the higher the system pressure and depressurization rate, the greater the microbial inactivation. The salmon: CO2 mass relation did not influence the S. aureus inactivation in raw salmon. There was a total reduction in S. aureus with 225 bar, a depressurizing rate of 100 bar.min–1, a salmon: CO2 mass relation of 1:0.6, for 2 hours at 33 °C.

  2. Basement radon entry and stack driven moisture infiltration reduced by active soil depressurization

    Science.gov (United States)

    C.R. Boardman; Samuel V. Glass

    2015-01-01

    This case study presents measurements of radon and moisture infiltration from soil gases into the basement of an unoccupied research house in Madison, Wisconsin, over two full years. The basement floor and exterior walls were constructed with preservative-treated lumber and plywood. In addition to continuous radon monitoring, measurements included building air...

  3. System and method for slurry handling

    Science.gov (United States)

    Steele, Raymond Douglas; Oppenheim, Judith Pauline

    2015-12-29

    A system includes a slurry depressurizing system that includes a liquid expansion system configured to continuously receive a slurry at a first pressure and continuously discharge the slurry at a second pressure. For example, the slurry depressurizing system may include an expansion turbine to expand the slurry from the first pressure to the second pressure.

  4. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  5. Mathematical Model and Simulation of Gas Hydrate Reservoir Decomposition by Depressurization Modèle mathématique et simulation de dépressurisation et de décompression d’un réservoir d’hydrates de méthane

    Directory of Open Access Journals (Sweden)

    Zhao J.

    2012-05-01

    Full Text Available The numerical model for the depressurization of methane hydrates in a confined reservoir is presented based on mass conservation in porous media, incorporating multiphase flow theory and kinetics of gas hydrate dissociation. The universal implicit difference method is adopted, and the corresponding computer program is developed. During the production of the hydrate reservoir, distribution and the physical changes are analyzed and the gas hydrate dissociation and gas production law are studied from the computation. A numerical simulation shows that the reservoir pressure is descending slowly, which benefits the stabilization of the reservoir and inevitably decreases the efficiency in the production of gas hydrates in the depressurizing process. The gas production rate is controlled by the well pressure. The results are presented to show how this model may be used to estimate a lower downhole pressure of the well for hydrate recovery and how these results depend on reservoir and hydrate properties. Le modèle numérique présenté ici simule la dépressurisation d’hydrates de méthane dans un réservoir confiné; il se base sur le principe de conservation de la masse en milieu poreux, en intégrant la théorie de l’écoulement polyphasique et la cinétique de dissociation des hydrates de méthane. La méthode implicite et universelle des différences finies est utilisée et le programme informatique qui s’y rapporte est développé. Lors de l’exploitation du réservoir d’hydrates de méthane, la répartition et les changements physiques sont analysés et les lois sur la dissociation des hydrates de méthane et la production de gaz sont étudiées à partir des calculs. Une simulation numérique montre que la pression dans le réservoir diminue lentement, ce qui permet au réservoir de se stabiliser et diminue inévitablement le rendement de l’exploitation d’hydrates de méthane lors du processus de dépressurisation. Le rythme de

  6. On the mineralization model of 'three sources--heat, water and uranium'

    International Nuclear Information System (INIS)

    Li Xueli

    1992-01-01

    In response to the relations between geological and geothermal settings, geothermal water and uranium mineralizations in the Southeastern China, the model of uranium mineralization in discharge area (depressurization area) of fossil geothermal systems in Mesozoic-Cenozoic Volcanic-magmatic active areas has been put forward and expounded in the view of mineral-formation by the 'three sources'-heat, water and uranium

  7. Soil gas measurements at high permeabilities and below foundation depth

    International Nuclear Information System (INIS)

    Johner, H.U; Surbeck, H.

    2000-01-01

    We started a project of soil gas measurements beneath houses. Since the foundations of houses often lie deeper than 0.5 to 1 m - the depth where soil gas measurements are often made - the first approach was to apply the method developed previously to deeper soil layers. The radon availability index (RAI), which was defined empirically, proved to be a reliable indicator for radon problems in nearby houses. The extreme values of permeability, non-Darcy flow and scale dependence of permeability stimulated the development of a multi-probe method. A hydrological model was applied to model the soil gas transport. The soil gas measurements below foundation depth provided a wealth of new information. A good classification of soil properties could be achieved. If soil gas measurements are to be made, the low permeability layer has to be traversed. A minimum depth of 1 .5 m is suggested, profiles to below the foundation depth are preferable. There are also implications for mitigation works. A sub-slab suction system should reach the permeable layer to function well. This also holds for radon wells. If a house is located on a slope, it is most convenient to install the sub-slab suction system on the hillside, as the foundation reaches the deepest levels there

  8. Sediment–well interaction during depressurization

    KAUST Repository

    Shin, Hosung; Santamarina, Carlos

    2016-01-01

    production from hydrate accumulations in marine sediments. Sediment–well interaction is examined using a nonlinear finite element simulator. The hydro-mechanically coupled model represents the sediment as a Cam-Clay material, uses a continuous function

  9. Radon remediation in irish schools

    International Nuclear Information System (INIS)

    Synnott, H.

    2006-01-01

    Full text: Commencing in 1998, the Radiological Protection Institute of Ireland carried out radon measurements in 3826 schools in the Republic of I reland on behalf of the Irish Department of Education and Science (D.E.S.). This represents approximately 97% of all schools in the country. Approximately 25% (984) schools had radon concentrations above the Irish national schools Reference Level for radon of 200 Bq/m 3 and required remedial work. The number of individual rooms with radon concentrations above 200 Bq/m 3 was 3020. Remedial work in schools commenced in early 2000. In general schools with maximum radon concentrations in the range 200 -400 Bq/m 3 in one or more rooms were remediated through the installation of passive systems such as an increase in permanent background ventilation mainly wall vents and trickle vents in windows. Schools with maximum radon concentrations greater than 400 Bq/m 3 were usually remediated through the provision of active systems mainly fan assisted sub -slab de pressurization or where this was not possible fan assisted under floor ventilation. The cost of the remedial programme was funded by central Government. Active systems were installed by specialized remedial contractors working to the specifications of a radon remedial expert appointed by the D.E.S. to design remedial systems for affected schools. Schools requiring increased ventilation were granted aided 190 pounds per affected room and had to organize the work themselves. In most schools radon remediation was successful in reducing existing radon concentrations to below the Reference Level. Average radon concentration reduction factors for sub-slab de pressurization systems and fan assisted fan assisted under floor ventilation ranged from 5 to 40 with greater reduction rates found at higher original radon concentrations. Increasing ventilation in locations with moderately elevated radon concentrations (200 - 400 Bq/m 3 ) while not as effective as active systems produced on

  10. Analysis of an ADS spurious opening event at a BWR/6 by means of the TRACE code

    International Nuclear Information System (INIS)

    Nikitin, Konstantin; Manera, Annalisa

    2011-01-01

    Highlights: → The spurious opening of 8 relief valves of the ADS system in a BWR/6 has been simulated. → The valves opening results in a fast depressurization and significant loads on the RPV internals. → This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the available plant data. - Abstract: The paper presents the results of a post-event analysis of a spurious opening of 8 relief valves of the automatic depressurization system (ADS) occurred in a BWR/6. The opening of the relief valves results in a fast depressurization (pressure blow down) of the primary system which might lead to significant dynamic loads on the RPV and associated internals. In addition, the RPV level swelling caused by the fast depressurization might lead to undesired water carry-over into the steam line and through the safety relief valves (SRVs). Therefore, the transient needs to be characterized in terms of evolution of pressure, temperature and fluid distribution in the system. This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the plant data.

  11. Gest-sip1 experiments and post-test calculations with the relap5 code

    International Nuclear Information System (INIS)

    Achilli, A.; Cattadori, G.; Ferri, R.; Gandolfi, S.; Bianchi, F.; Meloni, P.

    2001-01-01

    The SIP-1 apparatus (Sistema di Iniezione Passiva) was conceived, designed, numerically simulated and tested by the SIET company as an innovative depressurization and make-up device for the New Generation LWRs. In particular it is suitable to cope with those accidents where pressure in the circuit must be dumped to allow low pressure injection systems to intervene. The main peculiarity of SIP-1 is the capability of de-pressurizing a system by cold water injection, rather than by discharging mass to the outlet, as in the common depressurization systems. ENEA sponsored all the research activity, starting from the SIP-1 design, its numerical simulation with the Relap5 code, the realisation of an experimental facility up to the test execution and post-test calculations. An experimental campaign on the GEST-SIP1 facility was performed in July 2000. The facility is mainly constituted by a U-tube Steam Generator which a proper model of SIP-1 apparatus is connected to. A series of Small Break LOCAs was simulated by varying the break size and different steady conditions were investigated to verify the stability of SIP-1, the lack of unexpected interventions and the actuation modalities. This paper deals with the description of the GEST-SIP1 experimental facility, the SIP-1 operating principles, the most meaningful results of the tests and the capability of the Relap5 code in reproducing phenomena and events. (author)

  12. Underground mining of the lower 163 zone through groundwater drainage at the Eagle Point Mine

    International Nuclear Information System (INIS)

    Robson, D.M.; Bashir, R.; Thomson, J.; Klemmer, S.; Rigden, A.

    2010-01-01

    The Eagle Point Mine is part of the Cameco Rabbit Lake Operation. The mine produces uranium ore using the long-hole, vertical and horizontal retreat mining method. The majority of the mine workings are under Wollaston Lake and cementitious grouting is used as one of the water control measures. Historical groundwater table in the mining area was close to ground surface. The Lower 163 Zone encompasses an estimated 4.2 million pounds U_3O_8 geological resource that was not considered feasible to mine due to the expected groundwater flows in the area. Cross-hole testing was conducted to better understand the groundwater flow through various geologic units. A local depressurization test was conducted to assess the potential for lowering the water table. Following testing an active depressurization was conducted to lower the groundwater table below the planned mining areas. This resulted in safe and drier mining conditions and allowed for the successful extraction of the ore body. (author)

  13. Identification of advective entry of soil-gas radon into a crawl space covered with sheets of polyethylene foil

    International Nuclear Information System (INIS)

    Andersen, C.; Koopmanns, M.; Meijer, R.J. de

    1996-04-01

    To assess the effectiveness of mitigative measures against radon ( 222 Rn) entry into houses, experiments were conducted in a crawl-space house where the dirt floor of the crawl space was covered with sheets of 0.23 mm polyethylene foil fixed to the walls. The radon concentration was measured below the foil and in the crawl space together with environmental variables such as indoor-outdoor pressure differences. The experimental data was analyzed using various types of models including a simplistic mass-balance model, a regression model, and a two-dimensional numerical model based on Darcy flow or soil gas and combined diffusive and advective transport of radon. The main outcome of the work was that: (i) The soil-gas entry rate per pascal depressurization was at the order of 1 m 3 h -1 , (ii) the stack-related part of the depressurization of the crawl space (approx. 0.1 Pa deg. C -1 ) was controlled by the temperature difference between the living room of the house and the outdoors (not by the difference between the crawl space and the outdoors), (iii) that part of the wind-related depressurization that was measured by the pressure transducers seemed to force radon into the crawl space in the same proportion as the stack-related part of the depressurization, (iv) the ratio of advective and diffusive entry was approx. 0.7, when the crawl space was depressurized 1.5 Pa, (v) the effective diffusivity of the foil was found to be three orders of magnitude larger than that measured in the laboratory (the enhanced diffusivity was most likely caused by leaks in the foil and by mixing fans located in the crawl space), and (vi) there was no measurable mitigative impact of having the sheets of foil on the crawl-space floor even if the crawl space was artificially pressurized or depressurized. (au) 28 tabs., 36 ills., 61 refs

  14. Identification of advective entry of soil-gas radon into a crawl space covered with sheets of polyethylene foil

    Energy Technology Data Exchange (ETDEWEB)

    Andersen, C. [Risoe National Lab., Dept. of Nucl. Safety Res. and Nucl. Facilities, Roskilde (Denmark); Koopmanns, M.; Meijer, R.J. de [Kernfysische Versneller Inst., Environmental Radioactivity Res., Groningen (Netherlands)

    1996-04-01

    To assess the effectiveness of mitigative measures against radon ({sup 222}Rn) entry into houses, experiments were conducted in a crawl-space house where the dirt floor of the crawl space was covered with sheets of 0.23 mm polyethylene foil fixed to the walls. The radon concentration was measured below the foil and in the crawl space together with environmental variables such as indoor-outdoor pressure differences. The experimental data was analyzed using various types of models including a simplistic mass-balance model, a regression model, and a two-dimensional numerical model based on Darcy flow or soil gas and combined diffusive and advective transport of radon. The main outcome of the work was that: (i) The soil-gas entry rate per pascal depressurization was at the order of 1 m{sup 3} h{sup -1}, (ii) the stack-related part of the depressurization of the crawl space (approx. 0.1 Pa deg. C{sup -1}) was controlled by the temperature difference between the living room of the house and the outdoors (not by the difference between the crawl space and the outdoors), (iii) that part of the wind-related depressurization that was measured by the pressure transducers seemed to force radon into the crawl space in the same proportion as the stack-related part of the depressurization, (iv) the ratio of advective and diffusive entry was approx. 0.7, when the crawl space was depressurized 1.5 Pa, (v) the effective diffusivity of the foil was found to be three orders of magnitude larger than that measured in the laboratory (the enhanced diffusivity was most likely caused by leaks in the foil and by mixing fans located in the crawl space), and (vi) there was no measurable mitigative impact of having the sheets of foil on the crawl-space floor even if the crawl space was artificially pressurized or depressurized. (au) 28 tabs., 36 ills., 61 refs.

  15. Long term performance of radon mitigation systems

    International Nuclear Information System (INIS)

    Prill, R.; Fisk, W.J.

    2002-01-01

    Researchers installed radon mitigation systems in 12 houses in Spokane, Washington and Coeur d'Alene, Idaho during the heating season 1985--1986 and continued to monitor indoor radon quarterly and annually for ten years. The mitigation systems included active sub-slab ventilation, basement over-pressurization, and crawlspace isolation and ventilation. The occupants reported various operational problems with these early mitigation systems. The long-term radon measurements were essential to track the effectiveness of the mitigation systems over time. All 12 homes were visited during the second year of the study, while a second set 5 homes was visited during the fifth year to determine the cause(s) of increased radon in the homes. During these visits, the mitigation systems were inspected and measurements of system performance were made. Maintenance and modifications were performed to improve system performance in these homes

  16. Numerical Simulation of the Depressurization Process of a Natural Gas Hydrate Reservoir: An Attempt at Optimization of Field Operational Factors with Multiple Wells in a Real 3D Geological Model

    Directory of Open Access Journals (Sweden)

    Zhixue Sun

    2016-09-01

    Full Text Available Natural gas hydrates, crystalline solids whose gas molecules are so compressed that they are denser than a typical fluid hydrocarbon, have extensive applications in the areas of climate change and the energy crisis. The hydrate deposit located in the Shenhu Area on the continental slope of the South China Sea is regarded as the most promising target for gas hydrate exploration in China. Samples taken at drilling site SH2 have indicated a high abundance of methane hydrate reserves in clay sediments. In the last few decades, with its relatively low energy cost, the depressurization gas recovery method has been generally regarded as technically feasible and the most promising one. For the purpose of a better acquaintance with the feasible field operational factors and processes which control the production behavior of a real 3D geological CH4-hydrate deposit, it is urgent to figure out the effects of the parameters such as well type, well spacing, bottom hole pressure, and perforation intervals on methane recovery. One years’ numerical simulation results show that under the condition of 3000 kPa constant bottom hole pressure, 1000 m well spacing, perforation in higher intervals and with one horizontal well, the daily peak gas rate can reach 4325.02 m3 and the cumulative gas volume is 1.291 × 106 m3. What’s more, some new knowledge and its explanation of the curve tendency and evolution for the production process are provided. Technically, one factor at a time design (OFAT and an orthogonal design were used in the simulation to investigate which factors dominate the productivity ability and which is the most sensitive one. The results indicated that the order of effects of the factors on gas yield was perforation interval > bottom hole pressure > well spacing.

  17. Assessment of the potential for HPME during a station blackout in the Surry and Zion PWRS

    International Nuclear Information System (INIS)

    Knudson, D.L.; Bayless, P.D.; Dobbe, C.A.; Odar, F.

    1994-01-01

    The integrity of a PWR (pressurized water reactor) containment structure could be challenged by direct heating associated with a HPME (high pressure melt ejection) of core materials following reactor vessel lower head breach during certain severe accidents. Structural failure resulting from direct containment heating is a contributor to the risk of operating a PWR. Intentional RCS (reactor coolant system) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce those risks by mitigating the severity of the HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before failure of the lower head providing an unintentional mechanism for depressurization and HPME mitigation. This paper summarizes an assessment of RCS depressurization with respect to the potential for HPME during a station blackout in the Surry and Zion PWRs. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code and an evaluation of RCS depressurization-related probabilities primarily based on the code results

  18. Evaluation report on CCTF CORE-I REFLOOD TEST Cl-4 (Run 13) and Cl-15 (Run 24)

    International Nuclear Information System (INIS)

    Sudoh, Takashi; Murao, Yoshio.

    1983-08-01

    The tests Cl-4 and Cl-15 were performed with the Cylindrical Core Test Facility (CCTF) to investigate the effects of the depressurization process to simulate the refill phase, and the effects of the nitrogen to be injected after the end of the accumulator injection on the thermo-hydraulic behavior in the core and primary loop system during refill and reflood phases. In these tests, after the lower plenum was filled to 0.9m level with saturated water at 0.6 MPa, the accumulator water was injected into three intact cold legs in the depressurization period from 0.6 MPa to 0.2 MPa. The water in the lower plenum voided during the depressurization and the significant steam condensation occurred in or near the intact cold legs. The condensation caused high steam flow rate in the intact loops and the lower plenum flashing resulted in suppressed core water accumulation. The slightly lower core heat transfer coefficient due to the less core water caused the higher turnaround temperature and the longer quench time than those of the normal reflood test without the depressurization process. The nitrogen injection followed the accumulator injection was allowed in the test Cl-15. However, significant effects of the nitrogen injection was not observed. (author)

  19. ROSA-II test data report, 10

    International Nuclear Information System (INIS)

    1977-12-01

    Results of the ROSA-II test simulating a loss-of coolant accident (LOCA) in a light water reactor (LWR) are presented, including the test conditions and interpretation of the phenomena for test runs 415, 417, 421 and 422. Even in small break at the cold leg, the core is exposed to void and the temperature rises. In small break of the hot leg, however, core cooling keeps without temperature rise, because there still remains much residual water and upward core flow exists. Direct effect of the HPCI on the depressurization rate is small, but it increases the accumulator injection rate, leading to early core reflooding and early core cooling from upward. Effects of the secondary system depressurization are increase of depressurization and discharge rates of the primary loop, which results in early initiation of the accumulator injection and core reflooding. (auth.)

  20. Fluid structural response of axially cracked cylinders

    International Nuclear Information System (INIS)

    Garnich, M.R.; Simonen, F.A.

    1985-03-01

    The fluid structural (FS) response of a cylindrical pressure vessel to a suddenly occurring longitudinal through-wall crack is predicted. The effects of vessel internals and depressurization of the compressed water on dynamic crack opening displacements are investigated. A three dimensional (3D) structural finite element model is used as a basis for the development of a two dimensional (2D) FS model. A slice of the vessel taken at the crack midspan and normal to the cylinder axis is modeled. Crack opening displacements are compared between the 2D and 3D models, between the different assumptions about fluid depressurization, and between the static and dynamic solutions. The results show that effects of dynamic amplification associated with the sudden opening of the crack in the cylinder are largely offset by the local depressurization of the fluid adjacent to the crack

  1. Culturable prokaryotic diversity of deep, gas hydrate sediments: first use of a continuous high-pressure, anaerobic, enrichment and isolation system for subseafloor sediments (DeepIsoBUG)

    OpenAIRE

    Parkes, R John; Sellek, Gerard; Webster, Gordon; Martin, Derek; Anders, Erik; Weightman, Andrew J; Sass, Henrik

    2009-01-01

    Deep subseafloor sediments may contain depressurization-sensitive, anaerobic, piezophilic prokaryotes. To test this we developed the DeepIsoBUG system, which when coupled with the HYACINTH pressure-retaining drilling and core storage system and the PRESS core cutting and processing system, enables deep sediments to be handled without depressurization (up to 25 MPa) and anaerobic prokaryotic enrichments and isolation to be conducted up to 100 MPa. Here, we describe the system and its first use...

  2. Geodynamic Modeling of the Subduction Zone around the Japanese Islands

    Science.gov (United States)

    Honda, S.

    2017-06-01

    In this review, which focuses on our research, we describe the development of the thermomechanical modeling of subduction zones, paying special attention to those around the Japanese Islands. Without a sufficient amount of data and observations, models tended to be conceptual and general. However, the increasing power of computational tools has resulted in simple analytical and numerical models becoming more realistic, by incorporating the mantle flow around the subducting slab. The accumulation of observations and data has made it possible to construct regional models to understand the detail of the subduction processes. Recent advancements in the study of the seismic tomography and geology around the Japanese Islands has enabled new aspects of modeling the mantle processes. A good correlation between the seismic velocity anomalies and the finger-like distribution of volcanoes in northeast Japan has been recognized and small-scale convection (SSC) in the mantle wedge has been proposed to explain such a feature. The spatial and temporal evolution of the distribution of past volcanoes may reflect the characteristics of the flow in the mantle wedge, and points to the possibility of the flip-flopping of the finger-like pattern of the volcano distribution and the migration of volcanic activity from the back-arc side to the trench side. These observations are found to be qualitatively consistent with the results of the SSC model. We have also investigated the expected seismic anisotropy in the presence of SSC. The fast direction of the P-wave anisotropy generally shows the trench-normal direction with a reduced magnitude compared to the case without SSC. An analysis of full 3D seismic anisotropy is necessary to confirm the existence and nature of SSC. The 3D mantle flow around the subduction zone of plate-size scale has been modeled. It was found that the trench-parallel flow in the sub-slab mantle around the northern edge of the Pacific plate at the junction between

  3. Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Hong, Seong-Wan

    2011-01-01

    A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

  4. Simulation of a postulated 2% cold leg break in Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Palmieiri, Elcio Tadeu; Azevedo, Carlos Vicente Goulart de; Aronne, Ivan Dionysio

    2007-01-01

    This paper presents the simulation of a 2% break in the cold leg pipe of Angra 2 nuclear power plant, with the computer code RELAP5/Mod3.3. The main boundary conditions specified for this simulation were: no injection from high pressure injection system; enhanced depressurization of the primary system by opening the pressure operated relief valve (PORV) and the safety relief valve (SRV) when core temperature reaches circa 100 K above saturation; and accumulator injection starting at 2.7 MPa. The specific objectives to be addressed with this simulation are: the core boil-off and dryout at relatively high pressure in the primary system; the phenomena during enhanced primary depressurization; the effectiveness of hot leg accumulator injection into the partially uncovered rod bundle; and the core rewetting. The results obtained were compared with the Lobi A1-93 test, which was performed under the same boundary conditions. This activity was executed in the scope of IAEA research project Evaluation of Uncertainties in the Simulation of Accidents in Angra 2 using RELAP5/MOD3 Code Applying CIAU Methodology (author)

  5. BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

    1985-03-01

    This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

  6. Hydrate prevention in petroleum production sub sea system

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Paula L.F.; Rocha, Humberto A.R. [Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Rodrigues, Antonio P. [Universidade Federal do Rio Grande do Norte (UFRN), Natal, RN (Brazil)

    2012-07-01

    In spite of the merits of the several hydrate prevention techniques used nowadays, such as: chemical product injection for inhibition and use of thick thermal insulate lines; hydrates per times happen and they are responsible for considerable production losses. Depressurization techniques can be used so much for prevention as in the remediation. Some hydrate removal techniques need a rig or vessel, resources not readily available and with high cost, reason that limits such techniques just for remediation and not for prevention. In the present work it is proposed and described an innovative depressurization system, remote and resident, for hydrate prevention and removal, applicable as for individual sub sea wells as for grouped wells by manifold. Based on low cost jet pumps, without movable parts and with a high reliability, this technique allows hydrate prevention or remediation in a fast and remote way, operated from the production unit. The power fluid line and fluid return line can be integrated in the same umbilical or annulus line structure, without significant increase in the construction costs and installation. It is not necessary to wait for expensive resource mobilization, sometimes not available quickly, such as: vessels or rigs. It still reduces the chemical product consumption and permits to depressurized stopped lines. Other additional advantage, depressurization procedure can be used in the well starting, removing fluid until riser emptying. (author)

  7. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  8. Survival and ice nucleation activity of bacteria as aerosols in a cloud simulation chamber

    Science.gov (United States)

    Amato, P.; Joly, M.; Schaupp, C.; Attard, E.; Möhler, O.; Morris, C. E.; Brunet, Y.; Delort, A.-M.

    2015-06-01

    The residence time of bacterial cells in the atmosphere is predictable by numerical models. However, estimations of their aerial dispersion as living entities are limited by a lack of information concerning survival rates and behavior in relation to atmospheric water. Here we investigate the viability and ice nucleation (IN) activity of typical atmospheric ice nucleation active bacteria (Pseudomonas syringae and P. fluorescens) when airborne in a cloud simulation chamber (AIDA, Karlsruhe, Germany). Cell suspensions were sprayed into the chamber and aerosol samples were collected by impingement at designated times over a total duration of up to 18 h, and at some occasions after dissipation of a cloud formed by depressurization. Aerosol concentration was monitored simultaneously by online instruments. The cultivability of airborne cells decreased exponentially over time with a half-life time of 250 ± 30 min (about 3.5 to 4.5 h). In contrast, IN activity remained unchanged for several hours after aerosolization, demonstrating that IN activity was maintained after cell death. Interestingly, the relative abundance of IN active cells still airborne in the chamber was strongly decreased after cloud formation and dissipation. This illustrates the preferential precipitation of IN active cells by wet processes. Our results indicate that from 106 cells aerosolized from a surface, one would survive the average duration of its atmospheric journey estimated at 3.4 days. Statistically, this corresponds to the emission of 1 cell that achieves dissemination every ~ 33 min m-2 of cultivated crops fields, a strong source of airborne bacteria. Based on the observed survival rates, depending on wind speed, the trajectory endpoint could be situated several hundreds to thousands of kilometers from the emission source. These results should improve the representation of the aerial dissemination of bacteria in numeric models.

  9. Cooldown to residual heat removal entry conditions using atmospheric dump valves and auxiliary pressurizer spray following a loss-of-offsite power at Calvert Cliffs, Unit 1

    International Nuclear Information System (INIS)

    Jenks, R.P.

    1984-01-01

    An investigation of cooldown using atmospheric dump valves (ADVs) and auxiliary pressurizer spray (APS) following loss-of-offsite power at Calvert Cliffs-1 showed residual heat removal entry conditions could not be reached with the plant ADVs alone. Use of APS with the plant ADVs enhanced depressurization, but still provided insufficient cooldown. Effective cooldown and depressurization was shown to occur when rated steady state flow through the ADVs was increased by a factor of four. 6 refs., 30 figs., 2 tabs

  10. Inversion for Double-Layer Anisotropy in the Mantle Beneath the Middle America and Izu-Bonin Subduction Zones

    Science.gov (United States)

    Kuo, B. Y.

    2017-12-01

    We measured shear wave splitting for the intraslab events in the Middle America and Izu-Bonin subduction zones recorded at Pacific stations to infer the anisotropic structure in the subslab mantle. The receiver-side anisotropy is accounted for by considering both azimuthal anisotropy determined by SKS splitting and radial anisotropy given in global tomographic model, although the latter does not change the overall pattern of subslab anisotropy. By removing the anisotropy effects from both receiver and source sides, the initial polarization directions (p) of the shear waves used were recovered, most of which are in reasonable agreement with that predicted form the CMT solutions. For both subduction zones, the polarization-splitting plots strongly suggest the presence of two layers of anisotropy. To constrain the two-layer model, we perform inversions which minimize the misfit in both the splitting parameters and p. In the MASZ, the best model contains an upper layer with the fast direction in parallel with the absolute plate motion of the Cocos plate and a lower layer 40-60 degree clockwise from the APM. The delay times are 1.5 and 1.9 s respectively. The interference of the double layer produced dts in excess of 3 s at a certain range of p. The SKS splitting were also inverted for a two-layer model, yielding similar splitting characters and the clockwise rotation. We are investigating why this rotation takes place and how this observation is related to the dynamics of the asthenosphere.

  11. Active joint mechanism driven by multiple actuators made of flexible bags: a proposal of dual structural actuator.

    Science.gov (United States)

    Kimura, Hitoshi; Matsuzaki, Takuya; Kataoka, Mokutaro; Inou, Norio

    2013-01-01

    An actuator is required to change its speed and force depending on the situation. Using multiple actuators for one driving axis is one of the possible solutions; however, there is an associated problem of output power matching. This study proposes a new active joint mechanism using multiple actuators. Because the actuator is made of a flexible bag, it does not interfere with other actuators when it is depressurized. The proposed joint achieved coordinated motion of multiple actuators. This report also discusses a new actuator which has dual cylindrical structure. The cylinders are composed of flexible bags with different diameters. The joint torque is estimated based on the following factors: empirical formula for the flexible actuator torque, geometric relationship between the joint and the actuator, and the principle of virtual work. The prototype joint mechanism achieves coordinated motion of multiple actuators for one axis. With this motion, small inner actuator contributes high speed motion, whereas large outer actuator generates high torque. The performance of the prototype joint is examined by speed and torque measurements. The joint showed about 30% efficiency at 2.0 Nm load torque under 0.15 MPa air input.

  12. Risk impact of two accident management strategies

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, A.

    1992-01-01

    This report probabilistic Risk Assessment is used to evaluate two accident management strategies: intentionally depressurizing the reactor coolant system of a pressurized water reactor to prevent containment-pressurization during high pressure melt ejection, and flooding the containment of a boiling water reactor to prevent or delay vessel breach. Sensitivity studies indicated that intentional depressurization would not provide a significant risk reduction at Surry. A preliminary evaluation of the containment flooding strategy indicated that it might prove beneficial for some plants, but that further strategy development would be needed to fully evaluate the strategy-

  13. Validation of thermohydraulic codes by comparison of experimental results with computer simulations

    International Nuclear Information System (INIS)

    Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

    1989-01-01

    The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

  14. Experimental study of aerosol reentrainment from flashing pool in ALPHA program

    International Nuclear Information System (INIS)

    Kudo, T.; Yamano, N.; Moriyama, K.; Maruyama, Y.; Sugimoto, J.

    1994-01-01

    Aerosol reentrainment experiments are being performed as a part of the ALPHA (Assessment of Loads and Performance of Containment in a Hypothetical Accident) program at JAERI (Japan Atomic Energy Research Institute). The major objective of the experiments is to quantify and characterize the reentrainment of the dissolved material from a flashing pool during the rapid depressurization of a reactor containment vessel. Two experiments were performed. In the experiments a water pool dissolving sodium sulfate as FP simulant was located in the model containment vessel and the containment breach area was simulated with an orifice with 24 mm diameter. This orifice was estimated to give the same order of depressurization rate as the case of BWR Mark 1 containment failure with most likely breach size. In the first experiment ARE001, a pool water of 800 kg dissolving 50 kg of sodium sulfate was employed. The model containment was depressurized from 1.5 MPa to 0.1 MPa in approximately 45 minutes. In the second experiment ARE002, the mass of the pool water was reduced to 400 kg dissolving 25 kg of sodium sulfate. The internal pressure of the containment was decreased from 1.3 MPa to 0.1 MPa in approximately 40 minutes. At the beginning of the depressurization the pool water was heated to the saturation temperature at the internal pressure of the containment. The entrained droplets were sampled during depressurization period. Sodium sulfate deposited in all parts of the test facility was collected and weighed after the experiments. Results of the experiments showed that very small fraction of the dissolved material (less than 0.03%) was reentrained although approximately, 20% of water was evaporated from the pool water. The reentrained mass predicted with the Kataoka-Ishii model was approximately 1/110 of the mass evaluated in the experiments. This may be due to multi-dimensional features of the pool geometry. (author)

  15. Temporary Vs. Permanent Sub-Slab Ports: A Comparative Performance Study

    Data.gov (United States)

    U.S. Environmental Protection Agency — Data Tables from Published Article. This dataset is associated with the following publication: Zimmerman, J., C. Lutes, B. Cosky, B. Schumacher, D. Salkie, and R....

  16. Temporary vs. Permanent Sub-slab Ports: A Comparative Performance Study

    Science.gov (United States)

    Vapor intrusion (VI) is the migration of subsurface vapors, including radon and volatile organic compounds (VOCs), from the subsurface to indoor air. The VI exposure pathway extends from the contaminant source, which can be impacted soil, non-aqueous phase liquid, or contaminated...

  17. Ventilation systems analysis during tornado conditions. Progress report, January--June 1975

    International Nuclear Information System (INIS)

    Bennett, G.A.; Gregory, W.S.; Smith, P.R.

    1975-11-01

    The principal concern of this investigation is to develop the capability to simulate the dynamic effects of a tornado depressurization on a ventilation system. The basic formulation and solution of the two-zone series model ventilation subsystem is based on lumped parameter component response equations, the isothermal compression of air, and the conservation of mass. Solutions based on these assumptions are also presented for the two-zone series model with natural bypass, the two-zone series model with recirculation, and the natural branching model. A parameter study is presented comparing the effects of changes in system resistance, system capacitance, and variable tornado depressurization rates. The adaptability of the basic formulation to adiabatic compression of air and the addition of duct resistance is examined. A quasi-steady formulation is introduced and preliminary considerations of the importance of inertia are presented. Preliminary conclusions in this area indicate that inertial effects can be neglected. For relatively long ducts slow shock development appears possible. Work on the effect of tornado depressurization rates as related to shock development and on the importance of inertia effects is continuing

  18. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

    2001-01-01

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

  19. LOCA analysis of the IRIS reactor

    International Nuclear Information System (INIS)

    Bajs, T.; Grgic, D.; Cavlina, N.

    2003-01-01

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

  20. Microwave interaction with nonuniform hydrogen gas in carbon nanotubes

    International Nuclear Information System (INIS)

    Babaei, S.; Babaei, Sh.

    2009-01-01

    In this paper we study the reflection, absorption, and transmission of microwave from nonuniform hydrogen gas in carbon nanotubes, grown by iron-catalyzed high-pressure carbon monoxide disproportionate (HiPco) process. A discussion on the effect of various hydrogen gas parameters on the reflected power, absorbed power, and transmitted power is presented. The nonuniform hydrogen gas slab is modeled by a series of subslabs. The overall number density profile across the whole slab follows a parabolic function. The total reflected, absorbed, and transmitted powers are then deduced and their functional dependence on the number density, collision frequency, and angle of propagation is studied

  1. New school radon abatement systems

    International Nuclear Information System (INIS)

    Simon, R.F.; Maniscalco, P.J.

    1990-01-01

    This paper describes the methods used to develop a state-of-the-art Radon Abatement system: all aspects of design and implementation from proper sizing radon ventilation ductwork (RVD) in relationship to the amount of free air available in sub-slab aggregate, review of electrical systems with their monitoring devices from the very basic to the more sophisticated type of installation, review abatement designs for their durability and application as well as methods and techniques. Building codes will also be reviewed for commercial construction applications, spot-lighting the usage of specific materials and techniques and their impact on the industry

  2. Depressurized pipes decontamination by using circulation foam

    International Nuclear Information System (INIS)

    Damerval, Frederique; Belz, Jacques; Renouf, Marjorie; Janneau, Patrice

    2012-09-01

    Decontamination of pipes remains a necessity in order to reduce the radiation level during maintenance or dismantling operations but it is not so easy to do it, especially in case of a long pipe network. To achieve this operation, the use of chemistry is one of the more relevant methods; moreover, the liquid waste production still remains an issue that it can be avoided by the use of decontamination foams. (authors)

  3. Assessment of the potential for high-pressure melt ejection resulting from a Surry station blackout transient

    International Nuclear Information System (INIS)

    Knudson, D.L.; Dobbe, C.A.

    1993-11-01

    Containment integrity could be challenged by direct heating associated with a high pressure melt ejection (HPME) of core materials following reactor vessel breach during certain severe accidents. Intentional reactor coolant system (RCS) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce risks by mitigating the severity of HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before vessel breach, providing an alternate mechanism for RCS depressurization and HPME mitigation. This report contains an assessment of the potential for HPME during a Surry station blackout transient without operator action and without recovery. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code to calculate the plant response with and without hot leg countercurrent natural circulation, with and without reactor coolant pump seal leakage, and with variations on selected core damage progression parameters. RCS depressurization-related probabilities were also evaluated, primarily based on the code results

  4. Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Park, R.J.; Kim, S.B.; Kim, H.D.

    2004-01-01

    High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

  5. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  6. Test Methods and Protocols for Environmental and Safety Hazards Associated with Home Energy Retrofits

    Energy Technology Data Exchange (ETDEWEB)

    Cautley, D. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Viner, J. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Lord, M. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Pearce, M. [NorthernSTAR Building America Partnership, St. Paul, MN (United States)

    2012-12-01

    A number of health hazards and hazards to the durability of homes may be associated with energy retrofitting and home renovation projects. Among the hazards associated with energy retrofit work, exposure to radon is thought to cause more than 15,000 deaths per year in the U.S., while carbon monoxide poisoning results in about 20,000 injuries and 450 deaths per year. Testing procedures have been developed for identifying and quantifying hazards during retrofitting. These procedures commonly include a battery of tests to screen combustion appliances for safe operation, including worst case depressurization measurement, backdrafting (spillage) under depressurized or normal conditions, and carbon monoxide production.

  7. SFR Safety Consideration in Light of Fukushima Dai-ichi Accident

    International Nuclear Information System (INIS)

    Yamaguchi, Akira

    2013-01-01

    SFR Considerations: Fukushima Dai-ichi Accident: • Combined LORL and LOHS type initiated from SBO; • High pressure water-steam cooling system: – Depressurization - Not needed; – Ultimate heat sink - Robust (NC to atmosphere); – Continuous injection - Not needed (large sensible heat capacity). • Severe accident management: – RPV failure resulted in depressurization - Elevated temperature; – Heat sink to atmosphere - Freeing risk, sodium fire risk; – Mobile power supply - External resource may not be needed; – Seawater injection with fire engines - Sodium injection not needed; • Containment performance and accessibility: – Containment - Large containment volume and low pressure system; – Explosives - Sodium fire and hydrogen explosion

  8. Fuel Oxidizer Reaction Products (FORP) Contamination of Service Module (SM) and Release of N-nitrosodimethylamine(NDMA)in a Humid Environment from Crew EVA Suits Contaminated with FORP

    Science.gov (United States)

    Schmidl, William; Mikatarian, Ron; Lam, Chiu-Wing; West, Bil; Buchanan, Vanessa; Dee, Louis; Baker, David; Koontz, Steve

    2004-01-01

    The Service Module (SM) is an element of the Russian Segment of the International Space Station (ISS). One of the functions of the SM is to provide attitude control for the ISS using thrusters when the U.S. Control Moment Gyros (CMG's) must be desaturated. Prior to an Extravehicular Activity (EVA) on the Russian Segment, the Docking Compartment (DC1) is depressurized, as it is used as an airlock. When the DC1 is depressurized, the CMG's margin of momentum is insufficient and the SM attitude control thrusters need to fire to desaturate the CMG's. SM roll thruster firings induce contamination onto adjacent surfaces with Fuel Oxidizer Reaction Products (FORP). FORP is composed of both volatile and non-volatile components. One of the components of FORP is the potent carcinogen N-nitrosdimethylamine (NDMA). Since the EVA crewmembers often enter the area surrounding the thrusters for tasks on the aft end of the SM and when translating to other areas of the Russian Segment, the presence of FORP is a concern. This paper will discuss FORP contamination of the SM surfaces, the release of NDMA in a humid environment from crew EVA suits, if they happen to be contaminated with FORP, and the toxicological risk associated with the NDMA release.

  9. Cyclic Activity of Mud Volcanoes: Evidences from Trinidad (SE Caribbean)

    Science.gov (United States)

    Deville, E.

    2007-12-01

    Fluid and solid transfer in mud volcanoes show different phases of activity, including catastrophic events followed by periods of relative quiescence characterized by moderate activity. This can be notably shown by historical data onshore Trinidad. Several authors have evoked a possible link between the frequencies of eruption of some mud volcanoes and seismic activity, but in Trinidad there is no direct correlation between mud eruptions and seisms. It appears that each eruptive mud volcano has its own period of catastrophic activity, and this period is highly variable from one volcano to another. The frequency of activity of mud volcanoes seems essentially controlled by local pressure regime within the sedimentary pile. At the most, a seism can, in some cases, activate an eruption close to its term. The dynamics of expulsion of the mud volcanoes during the quiescence phases has been studied notably from temperature measurements within the mud conduits. The mud temperature is concurrently controlled by, either, the gas flux (endothermic gas depressurizing induces a cooling effect), or by the mud flux (mud is a vector for convective heat transfer). Complex temperature distribution was observed in large conduits and pools. Indeed, especially in the bigger pools, the temperature distribution characterizes convective cells with an upward displacement of mud above the deep outlet, and ring-shaped rolls associated with the burial of the mud on the flanks of the pools. In simple, tube-like shaped, narrow conduits, the temperature is more regular, but we observed different types of profiles, with either downward increasing or decreasing temperatures. If the upward flow of mud would be regular, we should expect increasing temperatures and progressively decreasing gradient with depth within the conduits. However, the variable measured profiles from one place to another, as well as time-variable measured temperatures within the conduits and especially, at the base of the

  10. Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

    2016-10-15

    A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

  11. Assessment of Literature Related to Combustion Appliance Venting Systems

    Energy Technology Data Exchange (ETDEWEB)

    Rapp, Vi H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Singer, Brett C. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Stratton, Chris [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Wray, Craig P. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2012-06-01

    In many residential building retrofit programs, air tightening to increase energy efficiency is constrained by concerns about related impacts on the safety of naturally vented combustion appliances. Tighter housing units more readily depressurize when exhaust equipment is operated, making combustion appliances more prone to backdraft or spillage. Several test methods purportedly assess the potential for depressurization-induced backdrafting and spillage, but these tests are not robustly reliable and repeatable predictors of venting performance, in part because they do not fully capture weather effects on venting performance. The purpose of this literature review is to investigate combustion safety diagnostics in existing codes, standards, and guidelines related to combustion appliances. This review summarizes existing combustion safety test methods, evaluations of these test methods, and also discusses research related to wind effects and the simulation of vent system performance. Current codes and standards related to combustion appliance installation provide little information on assessing backdrafting or spillage potential. A substantial amount of research has been conducted to assess combustion appliance backdrafting and spillage test methods, but primarily focuses on comparing short-term (stress) induced tests and monitoring results. Monitoring, typically performed over one week, indicated that combinations of environmental and house operation characteristics most conducive to combustion spillage were rare. Research, to an extent, has assessed existing combustion safety diagnostics for house depressurization, but the objectives of the diagnostics, both stress and monitoring, are not clearly defined. More research is also needed to quantify the frequency of test “failure” occurrence throughout the building stock and assess the statistical effects of weather (especially wind) on house depressurization and in turn on combustion appliance venting

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  13. Experimental studies on the thermal stratification and its influence on BLEVEs

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Wensheng; Gong, Yanwu; Gao, Ting; Gu, Anzhong; Lu, Xuesheng [Institute of Refrigeration and Cryogenics, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2010-10-15

    The thermal stratification of Liquefied Petroleum Gas (LPG) and its effect on the occurrence of the boiling liquid expanding vapor explosion (BLEVE) have been investigated experimentally. Stratifications in liquid and vapor occur when the LPG tank is heated. The degree of the liquid stratification {beta} increases with an increasing heat flux and decreasing filling ratio. The effect of stratification on the BLEVE has been examined with depressurization tests of LPG. The results show that the pressure recovery for the stratified LPG ({beta} = 1.4) upon sudden depressurization is much lower than that for the isothermal LPG ({beta} = 1). It can be concluded that the liquid stratification decreases the liquid energy and the occurrence of the BLEVE. (author)

  14. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

    2017-08-15

    An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  15. ROSA/LSTF experiment report for RUN SB-CL-24 repeated core heatup phenomena during 0.5% cold leg break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Mitsuhiro; Anoda, Yoshinari [Department of Reactor Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-03-01

    A small break loss-of-coolant accident (SBLOCA) in a Westinghouse-type four-loop PWR was simulated in an experiment (SB-CL-24) conducted at the Large-Scale Test Facility (LSTF) with an intention to study repeated core heatup during a long-term cooldown process. The experiment was conducted on February 28, 1990 with specified test conditions including failure assumptions both on the high pressure injection (HPI) and the auxiliary feedwater systems, and the intentional secondary system depressurization as an operator action. The secondary depressurization contributed to promote the primary depressurization and the actuation of accumulator injection system (AIS). A temporary core heatup was observed in each of three loopseal clearing (LSC) processes. A significant core heatup occurred in the following boil-off process after loss of the secondary coolant mass and the AIS termination due to increase of the primary pressure. By additional opening of the pressurizer relief valves and safety valves, the primary pressure rapidly decreased to result in the low pressure injection (LPI) which cooled the heated core. This report summarizes results of the experiment (SB-CL-24) in addition to typical responses of some accident indication systems including the core exit thermocouples (CETs) and the water level meters in the primary system. (author)

  16. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  17. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2017-08-01

    Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  18. The assessment of RELAP5/MOD2 based on pressurizer transient experiments

    International Nuclear Information System (INIS)

    Xue Hanjun; Tanrikut, A.; Menzel, R.

    1992-03-01

    Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

  19. Gas-cooled fast reactor safety

    International Nuclear Information System (INIS)

    Rickard, C.L.; Simon, R.H.; Buttemer, D.R.

    1977-01-01

    Initial conceptual design work on the GCFR began in the USA in the early 1960s and since the later 1960s has proceeded with considerable international cooperation. A 300 MWe GCFR demonstration plant employing three main cooling loops is currently being developed at General Atomic. A major preapplication licensing review of this demonstration plant was initiated in 1971 leading in 1974 to publication of a Safety Evaluation Report by the USAEC Directorate of Licensing. The preapplication review is continuing by addressing areas of concern identified in this report such that a major part of the work necessary to support the actual licensing of a GCFR demonstration plant has been established. The safety performance of the GCFR demonstration plant is based upon its inherent safety characteristics among which are the single phase and chemically inert coolant which is not activated and has a low reactivity worth, the negative core power and temperature reactivity coefficients and the small and negative steam reactivity worth. Recent studies of larger core designs indicate that as the reactor size increases central fuel, clad and coolant reactivity worths decrease and the Doppler coefficient becomes more negative. These inherent safety characteristics are complemented by safety design features such as enclosing the entire primary coolant system within a prestressed concrete pressure vessel (PCRV), providing two independent and diverse shutdown systems and residual heat removal (RHR) systems, limiting the worth of control rods to less than $1, employing pressure-equalized fuel rods, a core supported rigidly at its upper end and otherwise unrestrained and coolant downflow within the core to enhance debris removal should local melting occur. The structurally redundant PCRV design allows the potential depressurization leak area to be controlled and, since the PCRV is located within a containment building, coolant is present even after a depressurization accident and each RHR

  20. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

  1. System and method for continuous solids slurry depressurization

    Science.gov (United States)

    Leininger, Thomas Frederick; Steele, Raymond Douglas; Yen, Hsien-Chin William; Cordes, Stephen Michael

    2017-10-10

    A continuous slag processing system includes a rotating parallel disc pump, coupled to a motor and a brake. The rotating parallel disc pump includes opposing discs coupled to a shaft, an outlet configured to continuously receive a fluid at a first pressure, and an inlet configured to continuously discharge the fluid at a second pressure less than the first pressure. The rotating parallel disc pump is configurable in a reverse-acting pump mode and a letdown turbine mode. The motor is configured to drive the opposing discs about the shaft and against a flow of the fluid to control a difference between the first pressure and the second pressure in the reverse-acting pump mode. The brake is configured to resist rotation of the opposing discs about the shaft to control the difference between the first pressure and the second pressure in the letdown turbine mode.

  2. System and method for continuous solids slurry depressurization

    Science.gov (United States)

    Leininger, Thomas Frederick; Steele, Raymond Douglas; Cordes, Stephen Michael

    2017-07-11

    A system includes a first pump having a first outlet and a first inlet, and a controller. The first pump is configured to continuously receive a flow of a slurry into the first outlet at a first pressure and to continuously discharge the flow of the slurry from the first inlet at a second pressure less than the first pressure. The controller is configured to control a first speed of the first pump against the flow of the slurry based at least in part on the first pressure, wherein the first speed of the first pump is configured to resist a backflow of the slurry from the first outlet to the first inlet.

  3. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  4. Environmental protection and recovery of valuable feedstocks using carbon containing adsorbents. Developments and trends of Bergbau-Forschung GmbH

    Energy Technology Data Exchange (ETDEWEB)

    Knoblauch, K; Richter, E

    1986-06-01

    Activated carbons, active cokes and carbon molecular sieves are used for regenerative processes for environmental protection and for processing of valuable feedstocks. Development of adsorption processes and their lay-out base on adsorption equilibria, adsorption kinetics, kinetics of desorption by heating, depressurization or purging not only as single steps but in the same combination as in the regenerative process. For example, some adsorption processes are described which are applied in pilot scale or industrially: Nitrogen production from air by pressure swing adsorption (PSA); Hydrogen production from coke oven gas by PSA; Upgrading of methane from biogas and from fire damp; Removal of hydrogen sulfide from biogas; Removal of sulfur dioxide and nitrogen oxides from flue gases and drinking water supply and waste water treatment.

  5. Environmental protection and processing of feedstocks by adsorption on carbonaceous materials - developments at Bergbau- Forschung GmbH

    Energy Technology Data Exchange (ETDEWEB)

    Knoblauch, K; Richter, E

    1986-06-01

    Activated carbons, active cokes and carbon molecular sieves are used for regenerative processes for environmental protection and for processing of valuable feedstocks. Development of adsorption processes and their layout based on adsorption equilibria, adsorption kinetics, kinetics of desorption by heating, depressurization or purging not only as single steps but in the same combination as in the regenerative process. For example some adsorption processes are decsribed which are applied in pilot scale or industrially. These include: nitrogen production from air by pressure swing adsorption (PSA); hydrogen production from coke oven gas by PSA; upgrading of methane from biogas and from fire damp; removal of hydrogen sulfide from biogas; removal of sulfur dioxide and nitrogen oxides from flue gases and drinking water supply and waste water treatment. (71 refs.)

  6. RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

    International Nuclear Information System (INIS)

    Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

    2005-01-01

    Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test

  7. Experimental investigation of methane release from hydrate formation in sandstone through both hydrate dissociation and CO{sub 2} sequestration

    Energy Technology Data Exchange (ETDEWEB)

    Husebo, J.; Graue, A.; Kvamme, B. [Bergen Univ., Bergen (Norway). Dept. of Physics and Technology; Stevens, J.; Howard, J.J. [ConocoPhillips, Ponca City, OK (United States); Baldwin, B.A. [Green Country Petrophysics LLC, Dewey, OK (United States)

    2008-07-01

    Large amounts of natural gas trapped in hydrate reservoirs are found in Arctic regions and in deep offshore locations around the world. Natural gas production from hydrate deposits offer significant potential for future energy needs. However, research is needed in order to propose potential production schemes for natural gas hydrates. Natural gas molecules can be freed from hydrate structured cages by depressurization, by heating and by exposing the hydrate to a substance that will form a thermodynamically more stable hydrate structure. This paper provided a comparison of two approaches for releasing methane from methane hydrate in porous sandstone. The study scope covered the dissociation rate of methane hydrate in porous media through depressurization, and also referred to previous work done on producing methane from hydrates in sandstone while sequestering carbon dioxide (CO{sub 2}). The study was conducted in a laboratory setting. The paper discussed the experimental design which included the placing of a pressure- and temperature-controlled sample holder inside the bore of a magnetic resonance imager. The experimental procedures were then outlined, with reference to hydrate formation; carbon dioxide sequestration; hydrate dissociation experiments with constant volume; and hydrate dissociation experiments at constant pressure. The constant volume experiments demonstrated that in order to dissociate a large amount of hydrate, the initial depressurization had to be significantly lower than the hydrate stability pressure. 9 refs., 9 figs.

  8. Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

    Science.gov (United States)

    Clamens, Olivier; Lecerf, Johann; Hudelot, Jean-Pascal; Duc, Bertrand; Cadiou, Thierry; Blaise, Patrick; Biard, Bruno

    2018-01-01

    CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

  9. Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

    Directory of Open Access Journals (Sweden)

    Clamens Olivier

    2018-01-01

    Full Text Available CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

  10. Assessment of current techniques for reduction of indoor radon concentration in existing and new houses in European countries

    International Nuclear Information System (INIS)

    Holmgren, O.; Arvela, H.

    2012-03-01

    Radon control technologies aim at the reduction of indoor radon concentrations in existing buildings and in new construction through remedial and preventive measures. In recent years, rising ecological awareness and rising energy costs have stimulated the development of low energy and passive houses to save energy. This report contains the analysis and assessment of current techniques and technologies used to achieve the reduction of indoor radon concentrations in existing and new houses with regard to the reduction efficiency and potential impact on energy consumption (qualitative analysis). A questionnaire was prepared and sent to all RADPAR partners in 14 different countries in order to gather national information about the current remediation and prevention techniques. Responses with variable amounts of information were obtained. Based on the questionnaire responses, the status of radon remediation and prevention in each country was assessed, in addition to the reduction efficiency and potential impact on energy consumption of the current remediation and prevention techniques. The number of dwellings with an elevated indoor radon concentration typically ranges from tens of thousands to a million. The percentage of these houses already remediated varies from zero to 15%. Preventive measures in new construction have been taken from a small number of houses to over half a million houses. The research data on the current situation, the number of houses with preventive measures and the efficiency of these measures is currently still quite inadequate. Assessment of the techniques and also the surveys aiming at exploring the impact of remedial and preventive measures is greatly needed in order to promote the work at the national level. The most efficient remediation method is the active sub-slab depressurization (SSD) and the radon well, for which the reduction in the radon concentration is typically 70 - 95%. Other methods, such as sealing entry routes and improving

  11. Hydro-bio-geomechanical properties of hydrate-bearing sediments from Nankai Trough

    Science.gov (United States)

    Santamarina, J.C.; Dai, Shifeng; Terzariol, M.; Jang, Jeonghwan; Waite, William F.; Winters, William J.; Nagao, J.; Yoneda, J.; Konno, Y.; Fujii, T.; Suzuki, K.

    2015-01-01

    Natural hydrate-bearing sediments from the Nankai Trough, offshore Japan, were studied using the Pressure Core Characterization Tools (PCCTs) to obtain geomechanical, hydrological, electrical, and biological properties under in situ pressure, temperature, and restored effective stress conditions. Measurement results, combined with index-property data and analytical physics-based models, provide unique insight into hydrate-bearing sediments in situ. Tested cores contain some silty-sands, but are predominantly sandy- and clayey-silts. Hydrate saturations Sh range from 0.15 to 0.74, with significant concentrations in the silty-sands. Wave velocity and flexible-wall permeameter measurements on never-depressurized pressure-core sediments suggest hydrates in the coarser-grained zones, the silty-sands where Sh exceeds 0.4, contribute to soil-skeletal stability and are load-bearing. In the sandy- and clayey-silts, where Sh < 0.4, the state of effective stress and stress history are significant factors determining sediment stiffness. Controlled depressurization tests show that hydrate dissociation occurs too quickly to maintain thermodynamic equilibrium, and pressure–temperature conditions track the hydrate stability boundary in pure-water, rather than that in seawater, in spite of both the in situ pore water and the water used to maintain specimen pore pressure prior to dissociation being saline. Hydrate dissociation accompanied with fines migration caused up to 2.4% vertical strain contraction. The first-ever direct shear measurements on never-depressurized pressure-core specimens show hydrate-bearing sediments have higher sediment strength and peak friction angle than post-dissociation sediments, but the residual friction angle remains the same in both cases. Permeability measurements made before and after hydrate dissociation demonstrate that water permeability increases after dissociation, but the gain is limited by the transition from hydrate saturation

  12. Analysis of SBLOCA branch cold of a PWR without HPSI through the ISA methodology; Analisis de SBLOCA en Rama Fria de un PWR sin HPSI a traves de la Metodologia ISA

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayora, J.

    2012-07-01

    The objective of the study is the analysis of the appropriateness of the firing of the pumps of the primary and the time available for the beginning of the manual depressurization through steam generators.

  13. Radon remedial techniques in buildings - analysis of French actual cases

    International Nuclear Information System (INIS)

    Dupuis, M.

    2004-01-01

    The IRSN has compiled a collection of solutions from data provided by the various decentralised government services in 31 French departments. Contributors were asked to provide a description of the building, as well as details of measured radon levels, the type of reduction technique adopted and the cost. Illustrative layouts, technical drawings and photographs were also requested, when available. Of the cases recorded, 85% are establishments open to the public (schools (70%), city halls (4%) and combined city halls and school houses (26%)), 11% are houses and 4% industrial buildings. IRSN obtained 27 real cases of remedial techniques used. The data were presented in the form of fact sheets. The primary aim of this exercise was to illustrate each of the radon reduction techniques that can be used in the different building types (with basement, ground bearing slab, crawl space). This investigation not only enabled us to show that combining passive and active techniques reduces the operating cost of the installation, but above all that it considerably improves the efficiency. The passive technique reduces the amount of radon in the building and thus reduces the necessary ventilation rate, which directly affects the cost of operating the installation. For the 27 cases recorded, we noted:(a) the application of 7 passive techniques: sealing of floors and semi-buried walls, together with improved aeration by installing ventilation openings or ventilation strips in the windows. Radon concentrations were reduced on average by a factor of 4.7. No measurement in excess of 400 Bq.m -3 (the limit recommended by the French public authorities) was obtained following completion of the works; (b) the application of 15 active techniques: depressurization of the underlying ground, crawl space or basement and/or pressurization of the building. Radon concentrations were reduced on average by a factor of 13.8. Radon concentrations of over 400 Bq.m -3 were measured in only 4 cases

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

    International Nuclear Information System (INIS)

    Zentis, A.

    1993-01-01

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

  15. Effects of natural and forced basement ventilation on radon levels in single-family dwellings. Final report, May 90-Aug 91

    International Nuclear Information System (INIS)

    Cavallo, A.; Gadsby, K.; Reddy, T.A.

    1992-06-01

    The report gives, for the first time, results of an extensive study of the effect of ventilation on radon concentrations and radon entry rate in a single-family dwelling. Measurements of radon concentrations, building dynamics, and environmental parameters made in Princeton University research houses over several seasons and under different building operating conditions show the functional dependence of radon entry rate on basement depressurization. The work clarifies the role of natural ventilation in reducing indoor radon concentrations. The work shows conclusively that natural ventilation can decrease radon levels two ways: (1) by simple dilution, and (2) by providing a pressure break (defined as any opening in the building shell that reduces the outdoor/indoor differential pressure). This reduces building depressurization and thus the amount of radon-contaminated soil gas that is drawn into the building

  16. A preliminary study for the implementation of general accident management strategies

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung

    1997-01-01

    To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated

  17. PWR cold-leg small break loca with faulty HPI

    International Nuclear Information System (INIS)

    Kumamaru, H.; Kukita, Y.

    1991-01-01

    The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

  18. Natural circulation cooldown analysis for Yonggwang 3 and 4 per US NRC BTP RSB 5-1 requirements

    International Nuclear Information System (INIS)

    Seo, J.T.; Ko, C.S.; Ro, T.S.; Simoni, L.P.

    2004-01-01

    The Natural Circulation Cooldown (NCC) analysis from normal operations to shutdown cooling entry conditions for Yonggwang units 3 and 4 (YGN 3 and 4) was performed within the requirements of U.S. Nuclear Regulatory Commission (NRC) Branch Technical Position (BTP) RSB 5-1. The results showed that the YGN 3 and 4 can be cooled and depressurized to the shutdown entry conditions (350 deg F, 410 psia) within 16 hours under natural circulation condition requiring only 78% of the minimum condensate water storage capacity in conformance with BTP RSB 5-1 requirements. The results also demonstrated that the safety grade Reactor Coolant Gas Vent System (RCGVS) has sufficient capacity for the RCS depressurization as well as for the steam void control in the reactor vessel upper head region. (author)

  19. Methodology on the sparger development for Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K.

    1999-06-01

    In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloud acceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs

  20. Methodology on the sparger development for Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K

    1999-06-01

    In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloudacceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs.

  1. Effect of Mineral Dissolution/Precipitation and CO2 Exsolution on CO2 transport in Geological Carbon Storage.

    Science.gov (United States)

    Xu, Ruina; Li, Rong; Ma, Jin; He, Di; Jiang, Peixue

    2017-09-19

    Geological carbon sequestration (GCS) in deep saline aquifers is an effective means for storing carbon dioxide to address global climate change. As the time after injection increases, the safety of storage increases as the CO 2 transforms from a separate phase to CO 2 (aq) and HCO 3 - by dissolution and then to carbonates by mineral dissolution. However, subsequent depressurization could lead to dissolved CO 2 (aq) escaping from the formation water and creating a new separate phase which may reduce the GCS system safety. The mineral dissolution and the CO 2 exsolution and mineral precipitation during depressurization change the morphology, porosity, and permeability of the porous rock medium, which then affects the two-phase flow of the CO 2 and formation water. A better understanding of these effects on the CO 2 -water two-phase flow will improve predictions of the long-term CO 2 storage reliability, especially the impact of depressurization on the long-term stability. In this Account, we summarize our recent work on the effect of CO 2 exsolution and mineral dissolution/precipitation on CO 2 transport in GCS reservoirs. We place emphasis on understanding the behavior and transformation of the carbon components in the reservoir, including CO 2 (sc/g), CO 2 (aq), HCO 3 - , and carbonate minerals (calcite and dolomite), highlight their transport and mobility by coupled geochemical and two-phase flow processes, and consider the implications of these transport mechanisms on estimates of the long-term safety of GCS. We describe experimental and numerical pore- and core-scale methods used in our lab in conjunction with industrial and international partners to investigate these effects. Experimental results show how mineral dissolution affects permeability, capillary pressure, and relative permeability, which are important phenomena affecting the input parameters for reservoir flow modeling. The porosity and the absolute permeability increase when CO 2 dissolved water is

  2. CFD Analyses of Air-Ingress Accident for VHTRs

    Science.gov (United States)

    Ham, Tae Kyu

    The Very High Temperature Reactor (VHTR) is one of six proposed Generation-IV concepts for the next generation of nuclear powered plants. The VHTR is advantageous because it is able to operate at very high temperatures, thus producing highly efficient electrical generation and hydrogen production. A critical safety event of the VHTR is a loss-of-coolant accident. This accident is initiated, in its worst-case scenario, by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. Following the depressurization process, the air (i.e., the air and helium mixture) in the reactor cavity could enter the reactor core causing an air-ingress event. In the event of air-ingress into the reactor core, the high-temperature in-core graphite structures will chemically react with the air and could lose their structural integrity. We designed a 1/8th scaled-down test facility to develop an experimental database for studying the mechanisms involved in the air-ingress phenomenon. The current research focuses on the analysis of the air-ingress phenomenon using the computational fluid dynamics (CFD) tool ANSYS FLUENT for better understanding of the air-ingress phenomenon. The anticipated key steps in the air-ingress scenario for guillotine break of VHTR cross vessel are: 1) depressurization; 2) density-driven stratified flow; 3) local hot plenum natural circulation; 4) diffusion into the reactor core; and 5) global natural circulation. However, the OSU air-ingress test facility covers the time from depressurization to local hot plenum natural circulation. Prior to beginning the CFD simulations for the OSU air-ingress test facility, benchmark studies for the mechanisms which are related to the air-ingress accident, were performed to decide the appropriate physical models for the accident analysis. In addition, preliminary experiments were performed with a simplified 1/30th scaled down acrylic set-up to understand the air

  3. The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

    International Nuclear Information System (INIS)

    Hermansky, P.; Krajcovic, M.

    2012-01-01

    The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

  4. Scaling relation and regime map of explosive gas–liquid flow of binary Lennard-Jones particle system

    KAUST Repository

    Inaoka, Hajime; Yukawa, Satoshi; Ito, Nobuyasu

    2012-01-01

    liquid droplets, and gas particles, which remain supercritical gaseous states under the depressurization realized by simulations. The system has a pipe-like structure similar to the model of a shock tube. We observed physical quantities and flow regimes

  5. A preliminary study for the implementation of general accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

  6. A preliminary study for the implementation of general accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

  7. Analysis of Total Loss of Feedwater for APR1400 using SPACE

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Min; Park, Seok Jeong; Park, Chan Eok; Choi, Jong Ho; Lee, Gyu Cheon [KEPCO Engineering and Construction, Deajeon (Korea, Republic of)

    2016-10-15

    The Total Loss of FeedWater (TLOFW) event is an accident that main feedwater and auxiliary feedwater of secondary side are not supplied to steam generators. APR1400 uses the Safety Depressurization and Vent System (SDVS) for the F and B operation and SDVS is designed to perform the rapid depressurization function of Reactor Coolant System (RCS) through the remote manual operation when TLOFW is occurred. If RCS pressure falls below a Safety Injection Pump (SIP) working pressure, it can be possible to start the F and B operation which injects SIP flow to RCS and releases the RCS vapor and two-phase flow through Pilot Operated Safety Relief Valves (POSRVs) by opening the POSRVs, and then it can be possible to remove the decay heat. The design requirement of SDVS is that the core water level should be maintained at higher than 2 feet from the top of active core during the F and B operation. The TLOFW analysis was carried out to evaluate the capability of decay heat removal for APR1400 using newly developed SPACE code. The analysis results show that the F and B operation with 2 POSRVs and 2 SIPs and the F and B operation with 4 POSRVs and 4 SIPs meet the SDVS design requirement for the fuel cladding temperature. The comparison with RELAP5 shows good agreement and it validates the applicability of SPACE code for this type of accident analysis.

  8. Calculation of the frequency of excedence in Full Spectrum LOCA by FSR; Calculo de la Frecuencia de excedencia en Full Spectrum LOCA mediante metodologia ISA

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Magan, J. J.; Queral Salazar, C.; Sanchez Perea, M.

    2012-07-01

    In this application LOCA sequences was taken into account the uncertainty in the size of rupture and the operator action times as cooling and depressurization through steam generators. The simulations were performed using the tool SCAIS, dynamically coupled with MAAP code.

  9. Hip chondrolabral mechanics during activities of daily living: Role of the labrum and interstitial fluid pressurization.

    Science.gov (United States)

    Todd, Jocelyn N; Maak, Travis G; Ateshian, Gerard A; Maas, Steve A; Weiss, Jeffrey A

    2018-03-01

    Osteoarthritis of the hip can result from mechanical factors, which can be studied using finite element (FE) analysis. FE studies of the hip often assume there is no significant loss of fluid pressurization in the articular cartilage during simulated activities and approximate the material as incompressible and elastic. This study examined the conditions under which interstitial fluid load support remains sustained during physiological motions, as well as the role of the labrum in maintaining fluid load support and the effect of its presence on the solid phase of the surrounding cartilage. We found that dynamic motions of gait and squatting maintained consistent fluid load support between cycles, while static single-leg stance experienced slight fluid depressurization with significant reduction of solid phase stress and strain. Presence of the labrum did not significantly influence fluid load support within the articular cartilage, but prevented deformation at the cartilage edge, leading to lower stress and strain conditions in the cartilage. A morphologically accurate representation of collagen fibril orientation through the thickness of the articular cartilage was not necessary to predict fluid load support. However, comparison with simplified fibril reinforcement underscored the physiological importance. The results of this study demonstrate that an elastic incompressible material approximation is reasonable for modeling a limited number of cyclic motions of gait and squatting without significant loss of accuracy, but is not appropriate for static motions or numerous repeated motions. Additionally, effects seen from removal of the labrum motivate evaluation of labral reattachment strategies in the context of labral repair. Copyright © 2018 Elsevier Ltd. All rights reserved.

  10. Maximum Recoverable Gas from Hydrate Bearing Sediments by Depressurization

    KAUST Repository

    Terzariol, Marco; Goldsztein, G.; Santamarina, Carlos

    2017-01-01

    financial analyses; results highlight the need for innovative production strategies in order to make hydrate accumulations an economically-viable energy resource. Horizontal directional drilling and multi-wellpoint seafloor dewatering installations may lead

  11. Depressurization-filtration system of the containment of French PWR's

    International Nuclear Information System (INIS)

    L'homme, A.; Schektman, N.

    1987-01-01

    In the hypothetical event of a core meltdown occurring in a pressurized water reactor, and in order to preserve the integrity of the containment threatened by a build-up in pressure, EDF has developed, with the CEA, a decompression device which filters the containment internal atmosphere by using an unused containment penetration, and a sand-box, as filtering mechanism. This device and its procedure for utilization, constitute the U5 procedure. Check-tests on a semi-industrial scale have been carried out at the Nuclear Research Centre at Cadarache, by using columns of sand 80 cm high, according to following varying criteria: the granulometry of the sand, that of the aerosols, the flow-through speed, and the percentage steam content of the fluid to be filtered. The filtering material chosen is sand of a median diameter of 0.6 mm. (log normal distribution). The purification factor is above 10. The device tested meets the chosen targets, and is applied today to French units on condition to simple modifications concerning specific aspects of different series. The first is expected to be put into service during 1987

  12. Overview of the 2006-2008 JOGMEC/NRCan/Aurora Mallik Gas Hydrate Production Test Program

    Science.gov (United States)

    Yamamoto, K.; Dallimore, S. R.

    2008-12-01

    During the winters of 2007 and 2008 the Japan Oil, Gas and Metals National Corporation (JOGMEC) and Natural Resources Canada (NRCan), with Aurora Research Institute as the operator, carried out an on-shore gas hydrate production test program at the Mallik site, Mackenzie Delta, Northwest Territories, Canada. The prime objective of the program was to verify the feasibility of depressurization technique by drawing down the formation pressure across a 12m perforated gas hydrate bearing section. This project was the second full scale production test at this site following the 2002 Japex/JNOC/GSC et al Mallik research program in which seven participants organizatinos from five countries undertook a thermal test using hot water circulation Field work in 2007 was devoted to establishing a production test well, installing monitoring devices outside of casing, conducting base line geophysical studies and undertaking a short test to gain practical experience prior to longer term testing planned for 2008 . Hydrate-dissociated gas was produced to surface by depressurization achieved by lowering the fluid level with a dowhole pump. However, the operation was terminated 60 hours after the start of the pumping mainly due to sand production problems. In spite of the short period (12.5 hours of ellapsed pumping time), at least 830m3 of the gas was produced and accumulated in the borehole. Sand screens were installed across the perforated interval at the bottom hole for the 2008 program to overcome operational problems encountered in 2007 and achieve sustainable gas production. Stable bottom hole flowing pressures were successfully achieved during a 6 day test with continuous pump operation. Sustained gas production was achieved with rates between 2000- 4000m3/day and cummulative gas volume in the surface of approximately 13,000m3. Temperature and pressure data measured at the bottom hole and gas and water production rates gave positive evidence for the high efficiency of gas

  13. Rock fracture dynamics research at AECL's Underground Research Laboratory: applications to geological disposal of radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Young, R.P. [Univ. of Toronto, Toronto, ON (Canada); Haycox, J.R. [Applied Seismology Consultants Limited, Shrewsbury, Shropshire (United Kingdom); Martino, J. [Atomic Energy of Canada Limited, Pinawa, MB (Canada)

    2011-07-01

    Studies of rock fracture dynamics at AECL's Underground Research Laboratory (URL) have helped to provide a fundamental understanding of how crystalline rock responds to stresses induced from excavation, pressurization and temperature changes. The data acquired continue to provide insights into how a facility for the future geological disposal of radioactive waste could be engineered. Research into microseismic (MS), acoustic emission (AE), and ultrasonic velocity measurements has been performed on the full-scale sealed, pressurized, and heated horizontal elliptical tunnel at the Tunnel Sealing Experiment (TSX). The continuous monitoring of the experiment for 8 years provides a unique dataset for the understanding of the medium-term performance of an engineered disposal facility. This paper summarizes the results, interpretations and key findings of the experiment paying particular focus to the heating and cooling/depressurization of the chamber. Initial drilling of the tunnel and bulkheads causes microfracturing around the tunnel, mapped by MS and AEs, and is used as a benchmark for fracturing representing the excavated damaged zone (EDZ). There is no further extension to the volume during pressurization or heating of the tunnel suggesting an increase in crack density and coalescence of cracks rather than extension into unfractured rock. The dominant structure within the seismic cloud has been investigated using a statistical approach applying the three-point method. MS events in the roof exhibit a dominant pattern of sub-horizontal and shallow-dipping well defined planar features, but during cooling and depressurization a 45 degree dip normal to the tunnel axis is observed, which may be caused by movement in the rock-concrete interface due to differential cooling of the bulkhead and host rock. Cooling and depressurization of the TSX have not led to a significant increase in the number of MS or AE events. Ultrasonic results suggest the rock gets even stiffer

  14. Advancing Ruggedness of Nuclear Stations By Expanding Defence In Depth in Critical Areas

    International Nuclear Information System (INIS)

    Koshy, Thomas

    2015-01-01

    The nuclear industry continues to rise above the challenges it has faced over the years from external events and internal events. Fukushima event has shed light on a few vulnerabilities that could be overcome by utilizing the current state of technology. Common cause from sea water ingression was not conceived to have the entire electrical power system including AC and DC disabled beyond reasonable recovery. Rather than focusing on the solutions for lessons from Fukushima, it is better to address 'Fukushima type' events and advance the resilience of the NPPs. The effort needs to be on exploring different approaches to overcome such vulnerabilities so that a variety of solutions are available to make appropriate choices on improving NPP ruggedness based on anticipated challenges in the regions. In a technology neutral approach for light water reactors (LWR) there are 4 critical areas that are significant for ensuring nuclear safety. (1) Reactor trip, (2) Depressurization, (3) Emergency Core Cooling, and (4) Containment integrity. The reactor trip had not suffered any significant setbacks in the immediate past but provisions to address Anticipated Transients without Scram (ATWS) were generally included in most designs. While the technology has advanced, software driven/assisted trips are becoming popular and desirable. However, a diverse approach with least probability of potential interference needs to be provided in the control room and remote shutdown area to advance the ruggedness of rector trip. Depressurization is essential for passive as well as active cooling systems and therefore the approaches to de-pressurize should have more than one approach to ensure its success. In the absence of diverse approaches to de-pressurize, it is more important to consider RCS cooling capability during accidents or transients while the reactor is at a higher pressure. In the area of Emergency Core Cooling, the events history demonstrates greater success on diversity

  15. Assessment of Literature Related to Combustion Appliance Venting Systems

    Energy Technology Data Exchange (ETDEWEB)

    Rapp, V. H.; Less, B. D.; Singer, B. C.; Stratton, J. C.; Wray, C. P.

    2015-02-01

    In many residential building retrofit programs, air tightening to increase energy efficiency is often constrained by safety concerns with naturally vented combustion appliances. Tighter residential buildings more readily depressurize when exhaust equipment is operated, making combustion appliances more prone to backdraft or spill combustion exhaust into the living space. Several measures, such as installation guidelines, vent sizing codes, and combustion safety diagnostics, are in place with the intent to prevent backdrafting and combustion spillage, but the diagnostics conflict and the risk mitigation objective is inconsistent. This literature review summarizes the metrics and diagnostics used to assess combustion safety, documents their technical basis, and investigates their risk mitigations. It compiles information from the following: codes for combustion appliance venting and installation; standards and guidelines for combustion safety diagnostics; research evaluating combustion safety diagnostics; research investigating wind effects on building depressurization and venting; and software for simulating vent system performance.

  16. System transient response to loss of off-site power

    International Nuclear Information System (INIS)

    Sozer, A.

    1990-01-01

    A simultaneous trip of the reactor, main circulation pumps, secondary coolant pumps, and pressurizer pump due to loss of off-site power at the High Flux Isotope Reactor (HFIR) located at the Oak Ridge National Laboratory (ORNL) has been analyzed to estimate available safety margin. A computer model based on the Modular Modeling System code has been used to calculate the transient response of the system. The reactor depressurizes from 482.7 psia down to about 23 psia in about 50 seconds and remains stable thereafter. Available safety margin has been estimated in terms of the incipient boiling heat flux ratio. It is a conservative estimate due to assumed less than available primary and secondary flows and higher than normal depressurization rate. The ratio indicates no incipient boiling conditions at the hot spot. No potential damage to the fuel is likely to occur during this transient. 2 refs., 6 figs

  17. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  18. Steam Line Break Analysis in CAREM-25 Reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  19. Accident loads for a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Techy, Z. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Lajtha, G. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Taubner, R. [Institute for Electric Power Research (VEIKI), Budapest (Hungary)

    1995-08-01

    Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system. (orig.).

  20. A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

    International Nuclear Information System (INIS)

    Svenson, Ola

    1998-01-01

    This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences

  1. Reactive Skin Decontamination Lotion and Demilitarization: Partners in Global Safety

    International Nuclear Information System (INIS)

    Feghali, N.

    2007-01-01

    The work of demilitarization relies on procedures, equipment, and training to minimize exposure to Chemical Warfare agents. As an additional safeguard, immediate access to Reactive Skin Decontamination Lotion (RSDL) can minimize the harm of accidental exposure to CW agents. To illustrate the advantage of having RSDL available during demilitarization, this presentation will compare the outcomes of two patients who were exposed to Mustard, one of which was treated with RSDL and one that was not. Two individuals were exposed to mustard during a demilitarization operation at Defence Research and Development Canada; the two individuals received different treatments. Patient A: Treatment was not immediate, but RSDL was applied within the first 24 hours following exposure. Subsequently, treatment included depressurizing the blister, silver sulfadiazine dressing and the intake of oral antibiotics. Patient A healed over seven days and returned to work soon thereafter. Patient B: Treatment followed the then existing standard approach which included immersion in a detergent solution. Subsequently, treatment included depressurizing the blister, silver sulfadiazine dressing and oral antibiotics. Patient B required oral prophylactic antibiotics, topical and oral antihistamines and topical steroids and a recovery time of four weeks. While RSDL significantly increased the recovery time and reduced the suffering, it should be pointed out that if RSDL had been pre-positioned in the laboratory, an even greater impact on patient recovery should have been expected. Immediate access to RSDL is even more important for the faster acting agents and should be considered as an essential part of all demilitarization activities. (author)

  2. 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.)

  3. Solar radon reduction at six homes in northeast Iowa

    International Nuclear Information System (INIS)

    Rhoads, H.E.; Hoekje, P.L.

    1995-01-01

    Growing concern about radon lung cancer risks, carbon monoxide poisoning, and the sick building syndrome have increased demand for improved indoor air quality. Through solar pre-heating of ventilation air, the Solar Radon Reduction System (SRRS) provides energy benefits with lower installation costs than conventional air-to-air heat exchangers and sub-slab suction approaches. Indoor air quality is improved through dilution, combustion appliance make-up air, pressurization, and reduced radon infiltration through induced-draft solar air collectors drawing supply air from outdoors. Installed at six homes in Waterloo and Cedar Falls, Iowa, the SRRS was found to significantly reduce radon concentrations in all houses with energy benefits and improved overall indoor comfort. Up to 73% reductions from closed house levels as high as 20.9 pCi/L were achieved

  4. Source term estimation for small sized HTRs

    International Nuclear Information System (INIS)

    Moormann, R.

    1992-08-01

    Accidents which have to be considered are core heat-up, reactivity transients, water of air ingress and primary circuit depressurization. The main effort of this paper belongs to water/air ingress and depressurization, which requires consideration of fission product plateout under normal operation conditions; for the latter it is clearly shown, that absorption (penetration) mechanisms are much less important than assumed sometimes in the past. Source term estimation procedures for core heat-up events are shortly reviewed; reactivity transients are apparently covered by them. Besides a general literature survey including identification of areas with insufficient knowledge this paper contains some estimations on the thermomechanical behaviour of fission products in water in air ingress accidents. Typical source term examples are also presented. In an appendix, evaluations of the AVR experiments VAMPYR-I and -II with respect to plateout and fission product filter efficiency are outlined and used for a validation step of the new plateout code SPATRA. (orig.)

  5. On Small Disturbance Ascent Vent Behavior

    Science.gov (United States)

    Woronowicz, Michael

    2015-01-01

    As a spacecraft undergoes ascent in a launch vehicle, its ambient pressure environment transitions from one atmosphere to high vacuum in a matter of a few minutes. Venting of internal cavities is necessary to prevent the buildup of pressure differentials across cavity walls. These pressure differentials are often restricted to low levels to prevent violation of container integrity. Such vents usually consist of fixed orifices, ducts, or combinations of both. Duct conductance behavior is fundamentally different from that for orifices in pressure driven flows governing the launch vehicle ascent depressurization environment. Duct conductance is governed by the average pressure across its length, while orifice conductance is dictated by a pressure ratio. Hence, one cannot define a valid equivalent orifice for a given duct across a range of pressure levels. This presentation discusses development of expressions for these two types of vent elements in the limit of small pressure differentials, explores conditions for their validity, and compares their features regarding ascent depressurization performance.

  6. Assessment of thermalhydraulic phenomena for external water make-up

    International Nuclear Information System (INIS)

    Harwood, C.; Baschuk, J.

    2015-01-01

    Following the Fukushima Daiichi accident, Canadian NPP licensees implemented a number of changes, including additional provisions for water make-up to the reactor/boiler systems. The CNSC has placed a contract with CNL to model some of the make-up options, focusing on cooling via the boilers to prevent core damage. Such strategies have been credited with sustaining thermo syphoning in the primary system and thus prolonging the available time for the operator to provide pumped make-up to the boilers or emergency coolant injection to the core, thereby maintaining decay heat removal. This paper presents results of CATHENA calculations of an extended loss of all electrical power in which the operator manually depressurizes the boilers by crash-cooling, thus allowing water to flow to the boilers by gravity from the deaerator tank. The rapid cooling of the boilers promotes thermo-syphoning flow in the primary heat transport system and results in a corresponding cool-down and depressurization of this system. (author)

  7. Assessment of thermalhydraulic phenomena for external water make-up

    Energy Technology Data Exchange (ETDEWEB)

    Harwood, C., E-mail: Christopher.Harwood@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, ON (Canada); Baschuk, J. [Canadian Nuclear Laboratories, Chalk River, ON, (Canada)

    2015-07-01

    Following the Fukushima Daiichi accident, Canadian NPP licensees implemented a number of changes, including additional provisions for water make-up to the reactor/boiler systems. The CNSC has placed a contract with CNL to model some of the make-up options, focusing on cooling via the boilers to prevent core damage. Such strategies have been credited with sustaining thermo syphoning in the primary system and thus prolonging the available time for the operator to provide pumped make-up to the boilers or emergency coolant injection to the core, thereby maintaining decay heat removal. This paper presents results of CATHENA calculations of an extended loss of all electrical power in which the operator manually depressurizes the boilers by crash-cooling, thus allowing water to flow to the boilers by gravity from the deaerator tank. The rapid cooling of the boilers promotes thermo-syphoning flow in the primary heat transport system and results in a corresponding cool-down and depressurization of this system. (author)

  8. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  9. A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  10. A passive emergency heat sink for water cooled reactors with particular application to CANDU reg-sign reactors

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU reg-sign moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners

  11. Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

    Japan Atomic Energy Research Institute has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of a depressurization accident. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. (author)

  12. Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Hsu, Keng-Hsien, E-mail: hardlycampus@iner.gov.tw; Lin, Chin-Tsu, E-mail: jtling@iner.gov.tw

    2015-07-15

    Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

  13. The ultimate emergency measures to secure a NPP under an accidental condition with no designed power or water supply

    Energy Technology Data Exchange (ETDEWEB)

    Liang, K.S., E-mail: ksliang@alum.mit.edu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong-Chuan Road, Shanghai (China); Chiang, S.C. [Department of Nuclear Safety, Taiwan Power Company, 242 Sec. 3, Roosevelt Road, Taipei 10016, Taiwan (China); Hsu, Y.F.; Young, H.J.; Pei, B.S. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Wang, L.C. [Department of Nuclear Safety, Taiwan Power Company, 242 Sec. 3, Roosevelt Road, Taipei 10016, Taiwan (China)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer An ultimate measure to secure core was developed, if power or water supply cannot be restored in time. Black-Right-Pointing-Pointer This ultimate measure was simulated by RELAP5-3D to verify the concept of this emergency plan. Black-Right-Pointing-Pointer Quantification of the required raw water injection rate was performed for NPPS in Taiwan Black-Right-Pointing-Pointer Reactor controlled depressurization within the 1st hour is essential to reduce the required raw water injection rate. Black-Right-Pointing-Pointer For PWR, even heat sink can be developed, RCP seal leak might eventually cause core uncover 10 h after seal leak occurs. - Abstract: In the recent nuclear catastrophe which occurred in Japan on March 11, 2011, several units of Fukushima conventional BWR experienced a total loss of power and water supply triggered by a heavy earthquake and a following Tsunami beyond design basis. In Fukushima accident it was observed that sea water was injected into reactors only after hydrogen explosion took place and it was considered a little too late to prevent core from damage. With regard to this fact, the Taiwan power company develops an ultimate measure to prevent reactor from encountering core damage, if either designed AC power or reactor water supply cannot be restored in time. This ultimate measure was named as DIVing plan, abbreviated from system depressurization, water injection and containment venting. Once any designed AC power or reactor water supply is made available, this DIVing plan will be activated to (1) depressurize reactor first, (2) inject any available water into reactor by any available power supply if this critical status cannot be restored in time, and (3) vent the containment if necessary to maintain containment integrity. In this paper the DIVing plan was simulated by RELAP5-3D to verify the concept of it and also to quantify the required raw water injection rate to prevent core from damage for both

  14. The ultimate emergency measures to secure a NPP under an accidental condition with no designed power or water supply

    International Nuclear Information System (INIS)

    Liang, K.S.; Chiang, S.C.; Hsu, Y.F.; Young, H.J.; Pei, B.S.; Wang, L.C.

    2012-01-01

    Highlights: ► An ultimate measure to secure core was developed, if power or water supply cannot be restored in time. ► This ultimate measure was simulated by RELAP5-3D to verify the concept of this emergency plan. ► Quantification of the required raw water injection rate was performed for NPPS in Taiwan ► Reactor controlled depressurization within the 1st hour is essential to reduce the required raw water injection rate. ► For PWR, even heat sink can be developed, RCP seal leak might eventually cause core uncover 10 h after seal leak occurs. - Abstract: In the recent nuclear catastrophe which occurred in Japan on March 11, 2011, several units of Fukushima conventional BWR experienced a total loss of power and water supply triggered by a heavy earthquake and a following Tsunami beyond design basis. In Fukushima accident it was observed that sea water was injected into reactors only after hydrogen explosion took place and it was considered a little too late to prevent core from damage. With regard to this fact, the Taiwan power company develops an ultimate measure to prevent reactor from encountering core damage, if either designed AC power or reactor water supply cannot be restored in time. This ultimate measure was named as DIVing plan, abbreviated from system depressurization, water injection and containment venting. Once any designed AC power or reactor water supply is made available, this DIVing plan will be activated to (1) depressurize reactor first, (2) inject any available water into reactor by any available power supply if this critical status cannot be restored in time, and (3) vent the containment if necessary to maintain containment integrity. In this paper the DIVing plan was simulated by RELAP5-3D to verify the concept of it and also to quantify the required raw water injection rate to prevent core from damage for both PWR and BWR plants in Taiwan, after the loss of passive cooling mechanism. Provided the passive cooling mechanism is lost

  15. 76 FR 10482 - Special Conditions: Boeing Model 787-8 Airplane; Overhead Flightcrew-Rest Compartment Occupiable...

    Science.gov (United States)

    2011-02-25

    ... out. (b) For a manual, hand-held extinguishing system (designed as the sole means to fight a fire or..., except that Sec. 25.1309 remains at Amendment 25- 117 for cargo-fire protection systems. If the... to two flight crewmembers trained in the evacuation, fire fighting, and depressurization procedures...

  16. Development of technique for estimating primary cooling system break diameter in predicting nuclear emergency event sequence

    International Nuclear Information System (INIS)

    Tatebe, Yasumasa; Yoshida, Yoshitaka

    2012-01-01

    If an emergency event occurs in a nuclear power plant, appropriate action is selected and taken in accordance with the plant status, which changes from time to time, in order to prevent escalation and mitigate the event consequences. It is thus important to predict the event sequence and identify the plant behavior resulting from the action taken. In predicting the event sequence during a loss-of-coolant accident (LOCA), it is necessary to estimate break diameter. The conventional method for this estimation is time-consuming, since it involves multiple sensitivity analyses to determine the break diameter that is consistent with the plant behavior. To speed up the process of predicting the nuclear emergency event sequence, a new break diameter estimation technique that is applicable to pressurized water reactors was developed in this study. This technique enables the estimation of break diameter using the plant data sent from the safety parameter display system (SPDS), with focus on the depressurization rate in the reactor cooling system (RCS) during LOCA. The results of LOCA analysis, performed by varying the break diameter using the MAAP4 and RELAP5/MOD3.2 codes, confirmed that the RCS depressurization rate could be expressed by the log linear function of break diameter, except in the case of a small leak, in which RCS depressurization is affected by the coolant charging system and the high-pressure injection system. A correlation equation for break diameter estimation was developed from this function and tested for accuracy. Testing verified that the correlation equation could estimate break diameter accurately within an error of approximately 16%, even if the leak increases gradually, changing the plant status. (author)

  17. Mechanical model of human eye compliance for volumetric occlusion break surge measurements.

    Science.gov (United States)

    Dyk, David W; Miller, Kevin M

    2018-02-01

    To develop a mechanical model of human eye compliance for volumetric studies. Alcon Research, Ltd., Lake Forest, California, USA. Experimental study. Enucleated human eyes underwent pressurization and depressurization cycles with peak intraocular pressures (IOPs) of 60 to 100 mm Hg; anterior chamber pressure and volume changes were measured. Average net volume change curves were calculated as a function of IOP for each eye. Overall mean volumes were computed from each eye's average results at pressure points extrapolated over the range of 5 to 90 mm Hg. A 2-term exponential function was fit to these results. A fluid chamber with a displaceable piston was created as a mechanical model of this equation. A laser confocal displacement meter was used to measure piston displacement. A test bed incorporated the mechanical model with a mounted phacoemulsification probe and allowed for simulated occlusion breaks. Surge volume was calculated from piston displacement. An exponential function, V = C 1 × exp(C 2 × IOP) + C 3  × exp(C 4  × IOP) - V 0 , where V, the volume, was fit to the final depressurization curve obtained from 15 enucleated human eyes. The C 1 through C 4 values were -0.07141, -0.23055, -0.14972, and -0.02006, respectively. The equation was modeled using a piston system with 3 parallel springs that engaged serially. The mechanical model mimicked depressurization curves observed in human cadaver eyes. The resulting mechanical compliance model measured ocular volumetric changes and thus would be helpful in characterizing the postocclusion break surge response. Copyright © 2018 ASCRS and ESCRS. Published by Elsevier Inc. All rights reserved.

  18. Recent Ground Hold and Rapid Depressurization Testing of Multilayer Systems

    Science.gov (United States)

    Johnson, Wesley L.

    2014-01-01

    In the development of flight insulation systems for large cryogenic orbital storage (spray on foam and multilayer insulation), testing need include all environments that are experienced during flight. While large efforts have been expended on studying, bounding, and modeling the orbital performance of the insulation systems, little effort has been expended on the ground hold and ascent phases of a mission. Historical cryogenic in-space systems that have flown have been able to ignore these phases of flight due to the insulation system being within a vacuum jacket. In the development phase of the Nuclear Mars Vehicle and the Shuttle Nuclear Vehicle, several insulation systems were evaluated for the full mission cycle. Since that time there had been minimal work on these phases of flight until the Constellation program began investigating cryogenic service modules and long duration upper stages. With the inception of the Cryogenic Propellant Storage and Transfer Technology Demonstration Mission, a specific need was seen for the data and as such, several tests were added to the Cryogenic Boil-off Reduction System liquid hydrogen test matrix to provide more data on a insulation system. Testing was attempted with both gaseous nitrogen (GN2) and gaseous helium (GHe) backfills. The initial tests with nitrogen backfill were not successfully completed due to nitrogen liquefaction and solidification preventing the rapid pumpdown of the vacuum chamber. Subsequent helium backfill tests were successful and showed minimal degradation. The results are compared to the historical data.

  19. Linear accelerator section alignment in a vacuum chamber

    International Nuclear Information System (INIS)

    Vengrov, R.M.; Vinogradskij, N.N.; Danil'tsev, E.N.; Iosseliani, D.D.; Kosyak, V.S.; Porubaj, N.I.; Ugarov, S.B.

    1989-01-01

    Alignment technique for multisectional accelerating structures, that may be used in designing new accelerators for experimental and applied purposes, is described. The accuracy of the alignment of four-chamber resonator sections directly in an accelerator vacuum volume without its depressurization is not less than 100 μm. 8 refs.; 5 figs.; 5 tabs

  20. Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Jian Hongbing

    1992-01-01

    The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

  1. Source term estimation for small sized HTRs: status and further needs - a german approach

    International Nuclear Information System (INIS)

    Moormann, R.; Schenk, W.; Verfondern, K.

    2000-01-01

    The main results of German studies on source term estimation for small pebble-bed HTRs with their strict safety demands are outlined. Core heat-up events are no longer dominant for modern high quality fuel, but fission product transport during water ingress accidents (steam cycle plants) and depressurization is relevant, mainly due to remobilization of fission products which were plated-out in the course of normal operation or became dust borne. An important lack of knowledge was identified as concerns data on plate-out under normal operation, as well as on the behaviour of dust borne activity as a whole. Improved knowledge in this field is also important for maintenance/repair and design/shielding. For core heat-up events the influence of burn-up on temperature induced fission product release has to be measured for future high burn-up fuel. Also, transport mechanisms out of the He circuit into the environment require further examination. For water/steam ingress events mobilization of plated-out fission products by steam or water has to be considered in detail, along with steam interaction with kernels of particles with defective coatings. For source terms of depressurization, a more detailed knowledge of the flow pattern and shear forces on the various surfaces is necessary. In order to improve the knowledge on plate-out and dust in normal operation and to generate specimens for experimental remobilization studies, planning/design of plate-out/dust examination facilities which could be added to the next generation of HTRs (HTR10,HTTR) is proposed. For severe air ingress and reactivity accidents, behaviour of future advanced fuel elements has to be experimentally tested. (authors)

  2. Risk factors of methane hydrate resource development in the concentrated zones distributed in the eastern Nankai Trough

    Science.gov (United States)

    Yamamoto, K.; Nagakubo, S.

    2009-04-01

    Some environmental and safety concerns on the offshore methane hydrate development have been raised, but the ground of such allegations are sometime not fully reasonable. The risks of methane hydrate resource development to environment and safety should be discussed upon methane hydrate occurrences condition, the production methods, and the designs of production system, under comprehensively scientific manners. In the Phase 1 of the Methane Hydrate Exploitation Program in Japan (FY2001-2008), the Research Consortium for Methane Hydrate Resources in Japan (MH21 Research Consortium) found methane hydrate concentrated zones in the eastern Nankai Trough that are potential prospects for resource development. The concentrated zones are consisted of turbidite-derived sandy sediments and hydrate crystals in pore spaces of sand grains (pore-filling type structure). The MH21 Research Consortium proposed the depressurization method as prime technique due to its efficiency of gas production in such concentrated zones, and has tried to develop conceptual designs of production systems based on the information of existing devices and facilities. Under the condition and circumstances described above, the authors tried to extract and evaluate some risk factors concerning methane hydrate development using depressurization in the area. Leakage of methane gas, that is less harmful substance to ecosystem than heavier hydrocarbons, from production system can be one possible risk. However, in the case of gas production through wellbore, even if catastrophic damages happen in the subsea production system during gas production, the leakages do not continue because the borehole could be filled by seawater and depressurization is stopped immediately. Another possible risk is a leakage of produced gas through seafloor. If methane hydrate production makes high pressure or temperature zones in sediments, the risk should be considered. However, depressurization method makes opposite condition

  3. Modelling the effects of ice-sheet activity on CO2 outgassing by Icelandic volcanoes

    Science.gov (United States)

    Armitage, J. J.; Ferguson, D.; Petersen, K. D.; Creyts, T. T.

    2017-12-01

    Glacial cycles may play a significant role in mediating the flux of magmatic CO2 between the Earth's mantle and atmosphere. In Iceland, it is thought that late-Pleistocene deglaciation led to a significant volcanic pulse, evidenced by increased post-glacial lava volumes and changes in melt chemistry consistent with depressurization. Investigating the extent to which glacial activity may have affected volcanic CO2 emissions from Iceland, and crucially over what timescale, requires detailed knowledge of how the magma system responded to the growth and collapse of the ice-sheet before and after the LGM. To investigate this, we coupled a model of magma generation and transport with a history of ice-sheet activity. Our results show that the emplacement and removal of the LGM ice-sheet likely led to two significant pulses of magmatic CO2. The first, and most significant of these, is associated with ice-sheet growth and occurs as the magma system recovers from glacial loading. This recovery happens from the base of the melting region upwards, producing a pulse of CO2 rich magma that is predicted to reach the surface around 20 ka after the loading event, close in time to the LGM. The second peak in CO2 output occurs abruptly following deglaciation as a consequence of increased rates of melt generation and transport in the shallow mantle. Although these post-glacial melts are relatively depleted in CO2, the increase in magma flux leads to a short-lived period of elevated CO2 emissions. Our results therefore suggest a negative feedback, whereby ice-sheet growth produces a delayed pulse of magmatic CO2, which, in addition to increased geothermal heat flux, may contribute towards driving deglaciation, which itself then causes further magmatism and CO2 outgassing. This model is consistent with the seismic structure of the asthenosphere below Iceland, and the established compositional and volumetric trends for sub- and post-glacial volcanism in Iceland. These trends show that

  4. Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-03-16

    During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

  5. Direct observation of characteristic dissociation behaviors of hydrate-bearing cores by rapid-scanning X-ray CT imaging

    Energy Technology Data Exchange (ETDEWEB)

    Ebinuma, T.; Oyama, H.; Utiumi, T.; Nagao, J.; Narita, H. [National Inst. of Advanced Industrial Science and Technology, Toyohiraku, Sapporo (Japan)

    2008-07-01

    Methane hydrate has significant potential as a new source of energy. Major considerations in developing production methods of methane from hydrates are the fundamental properties of hydrate-bearing sediments, and the dissociation behavior of methane hydrate and the gas and water flow generated by its dissociation in sediments. Marine methane hydrates occur several hundred meters below the sea floor, in a variety of forms. The pore-space filling-type is considered to be the most suited to exploitation, as it is contained within the pore spaces of sandy sediments, and has relatively larger gas permeability compared to other forms. However, shallow sandy sediments are not usually consolidated, and methane hydrate is unstable at normal pressure and temperature. Therefore, common methods are not suitable, and new experimental methods have been developed to study the properties of hydrate-bearing sediment and its dissociation process. This paper presented the results of an experimental study involving the dissociation of artificial methane-hydrate-bearing sediments. The experiment was performed using X-ray computed tomography in order to directly observe dissociation behavior in the sediments and the gas and water flows generated by dissociation. The paper described the depressurization process and presented a schematic diagram of rapid scanning X-ray computed tomography scanner and core holder with tri-axial structure. The experimental apparatus for dissociation of methane hydrate was also illustrated. The thermal stimulation process and hot water injection process were explained. It was concluded that dissociation by depressurization demonstrated that the temperature reduction induced by depressurization depended on the phase equilibrium state of methane hydrate, and that dissociation preferentially occurred at the periphery of the core. This behavior was due to the heat flux from the outside of the core, where the heat flux controlled the dissociation rate. 10 refs

  6. Critical pressure of non-equilibrium two-phase critical flow

    Energy Technology Data Exchange (ETDEWEB)

    Minzer, U [Israel Electric Corp. Ltd., Haifa (Israel)

    1996-12-01

    Critical pressure is defined as the pressure existing at the exit edge of the piping, when it remains constant despite a decrease in the back. According to this definition the critical pressure is larger than the back pressure and for two-phase conditions below saturation pressure. The two-phase critical pressure has a major influence on the two-phase critical flow characteristics. Therefore it is of High significance in calculations of critical mass flux and critical depressurization rate, which are important in the fields of Nuclear Reactor Safety and Industrial Safety. At the Nuclear Reactor Safety field is useful for estimations of the Reactor Cooling System depressurization, the core coolant level, and the pressure build-up in the containment. In the Industrial Safety field it is helpful for estimating the leakage rate of toxic gases Tom liquefied gas pressure vessels, depressurization of pressure vessels, and explosion conditions due to liquefied gas release. For physical description of non-equilibrium two-phase critical flow it would be convenient to divide the flow into two stages. The first stage is the flow of subcooled liquid at constant temperature and uniform pressure drop (i.e., the case of incompressible fluid and uniform piping cross section). The rapid flow of the liquid causes a delay in the boiling of the liquid, which begins to boil below saturation pressure, at thermal non-equilibrium. The boiling is the beginning of the second stage, characterized by a sharp increase of the pressure drop. The liquid temperature on the second stage is almost constant because most of the energy for vaporization is supplied from the large pressure drop The present work will focus on the two-phase critical pressure of water, since water serves as coolant in the vast majority of nuclear power reactors throughout the world. (author).

  7. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  8. Critical pressure of non-equilibrium two-phase critical flow

    International Nuclear Information System (INIS)

    Minzer, U.

    1996-01-01

    Critical pressure is defined as the pressure existing at the exit edge of the piping, when it remains constant despite a decrease in the back. According to this definition the critical pressure is larger than the back pressure and for two-phase conditions below saturation pressure. The two-phase critical pressure has a major influence on the two-phase critical flow characteristics. Therefore it is of High significance in calculations of critical mass flux and critical depressurization rate, which are important in the fields of Nuclear Reactor Safety and Industrial Safety. At the Nuclear Reactor Safety field is useful for estimations of the Reactor Cooling System depressurization, the core coolant level, and the pressure build-up in the containment. In the Industrial Safety field it is helpful for estimating the leakage rate of toxic gases Tom liquefied gas pressure vessels, depressurization of pressure vessels, and explosion conditions due to liquefied gas release. For physical description of non-equilibrium two-phase critical flow it would be convenient to divide the flow into two stages. The first stage is the flow of subcooled liquid at constant temperature and uniform pressure drop (i.e., the case of incompressible fluid and uniform piping cross section). The rapid flow of the liquid causes a delay in the boiling of the liquid, which begins to boil below saturation pressure, at thermal non-equilibrium. The boiling is the beginning of the second stage, characterized by a sharp increase of the pressure drop. The liquid temperature on the second stage is almost constant because most of the energy for vaporization is supplied from the large pressure drop The present work will focus on the two-phase critical pressure of water, since water serves as coolant in the vast majority of nuclear power reactors throughout the world. (author)

  9. The use of stability indices in predicting asphaltene problems in upstream and downstream oil operations

    Energy Technology Data Exchange (ETDEWEB)

    Asomaning, S. [Baker Petrolite, Sugar Land, TX (United States)

    2003-07-01

    A series of test methods have been developed to determine the stability of asphaltenes in crude oils. They were developed due to the high cost of remediating asphaltene deposition in offshore operations. This study described the characteristics of the Oliensis Spot Test, two saturates, aromatics, resins and asphaltenes (SARA)-based screens (the Colloidal Instability Index and Asphaltene-Resin ratio), a solvent titration method with near infrared radiation (NIR) solids detection, and live oil depressurization. Each method is used to predict the stability of asphaltenes in crude oils with different API gravity. A complete description of the test methods was provided along with experimental data. The effectiveness of the different tests in predicting the stability of asphaltenes in crude oils was also assessed. Results indicate that the prediction of a crude oil's tendency towards asphaltene precipitation was more accurate with the Colloidal Instability Index and the solvent titration method. Live oil depressurization proved to be very effective in predicting the stability of asphaltenes for light oils, where most stability tests fail. tabs., figs.

  10. Gas production potential of disperse low-saturation hydrate accumulations in oceanic sediments

    International Nuclear Information System (INIS)

    Moridis, George J.; Sloan, E. Dendy

    2007-01-01

    In this paper, we evaluate the gas production potential of disperse, low-saturation (S H H hydrate-bearing sediments subject to depressurization-induced dissociation over a 10-year production period. We investigate the sensitivity of items (a)-(c) to the following hydraulic properties, reservoir conditions, and operational parameters: intrinsic permeability, porosity, pressure, temperature, hydrate saturation, and constant pressure at which the production well is kept. The results of this study indicate that, despite wide variations in the aforementioned parameters (covering the entire spectrum of such deposits), gas production is very limited, never exceeding a few thousand cubic meters of gas during the 10-year production period. Such low production volumes are orders of magnitude below commonly accepted standards of economic viability, and are further burdened with very unfavorable gas-to-water ratios. The unequivocal conclusion from this study is that disperse, low-S H hydrate accumulations in oceanic sediments are not promising targets for gas production by means of depressurization-induced dissociation, and resources for early hydrate exploitation should be focused elsewhere

  11. Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

    2017-12-15

    This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

  12. Variable reluctance displacement transducer temperature compensated to 6500F

    International Nuclear Information System (INIS)

    1975-01-01

    In pressurized water reactor tests, compact instruments for accurate measurement of small displacements in a 650 0 F environment are often required. In the case of blowdown tests such as the Loss of Fluid Test (LOFT) or Semiscale computer code development tests, not only is the initial environment water at 650 0 F and 2200 psi but it undergoes a severe transient due to depressurization. Since the LOFT and Semiscale tests are run just for the purpose of obtaining data during the depressurization, instruments used to obtain the data must not give false outputs induced by the change in environment. A LOFT rho v 2 probe and a Semiscale drag disk are described. Each utilizes a variable reluctance transducer (VRT) for indication of the drag-disk location and a torsion bar for drag-disk restoring force. The VRT, in addition to being thermally gain and null offset stable, is fabricated from materials known to be resistant to large nuclear radiation levels and has successfully passed a fast neutron radiation test of 2.7 x 10 17 nvt without failure

  13. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    International Nuclear Information System (INIS)

    Ruggles, A.E.; Cheng, L.Y.; Dimenna, R.A.; Griffith, P.; Wilson, G.E.

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis

  14. Pressure vessel SBLOCA simulation with trace: application to ISTF (Rosa V) - 151

    International Nuclear Information System (INIS)

    Abella, V.; Gallardo, S.; Verdu, G.

    2010-01-01

    In this work, an overview of the results obtained in the simulation of an Upper Head Small Break Loss-Of-Coolant-Accident (SBLOCA) under the assumption of total failure of High Pressure Injection System (HPIS) in the Large Scale Test Facility (LSTF) is provided. In previous works, an SBLOCA located in the Pressure Vessel (PV) Lower Plenum was simulated with TRACE. In that case, an asymmetrical steam generator secondary-side depressurization was produced as an accident management action at the Steam Generator in loop without pressurizer after the generation of safety injection signal to achieve a determined depressurization rate in the primary system. The new SBLOCA scenario has been simulated and results compared with experimental values, with the purpose of completing the analysis of PV SBLOCA. This study is developed in the frame of the OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA). Finally, the present paper represents a contribution for the study of safety analysis of vessel SBLOCAs and the assessment of the predictability of thermal-hydraulic codes like TRACE. (authors)

  15. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

    Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

  16. 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)

  17. Development of a feed-and-bleed operation strategy with hybrid-SIT under low pressure condition of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, In Seop, E-mail: jeoni@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States); Han, Sang Hoon, E-mail: shhan2@kaeri.re.kr [Advanced Research Group, Korea Atomic Energy Research Institute, 70 Daedeok-daero 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Kang, Sang Hee, E-mail: sanghee.kang@khnp.co.kr [NSSS Design Group, Korea Hydro & Nuclear Power Co., Ltd., Central Research Institute, 70, 1312-beongil, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States)

    2017-04-01

    Highlights: • The novel F&B operation strategy with H-SIT and LPSI is developed. • The effectiveness of the H-SITs is verified using thermo-hydraulic simulations. • Success criteria considered for the new F&B operation strategy is identified. • A PSA model of APR+ reflecting the new F&B strategy with H-SIT is developed. • A risk analysis of the proposed F&B operation strategy is performed. - Abstract: While safety functions in current nuclear power plants are mainly provided by active safety systems, recently passive safety systems are being combined with the active systems to strengthen accident mitigation capability and therefore enhance overall plant safety. To this end, securing long-term cooling of the core is of particular importance. This study considers the hybrid safety injection tank (H-SIT), a passive injection system, as a target component to develop a long-term cooling strategy using active and passive systems concurrently. In the feed-and-bleed (F&B) operation, one of the important long-term cooling strategies to maintain core safety in pressurized water reactors, low pressure safety injection (LPSI) pumps are typically considered inoperable as depressurization is first required, which leads to core dry-out before reaching LPSI operable pressure. This study investigates whether H-SITs, with the important design feature of passive coolant injection under any pressure condition of the primary coolant system, can make up the core during depressurization thereby allowing LPSI pumps to be used in F&B operation as an additional means of long-term cooling. The effectiveness of the H-SITs is verified using thermal-hydraulic simulations, and based on the results a novel F&B operation strategy with H-SITs and LPSI pumps is developed. A probabilistic safety assessment (PSA) model is then developed in order to assess the risk effect of the suggested strategy. PSA results demonstrate that the proposed strategy lowers core damage frequency in the target

  18. Broadband measurements of the sources of explosions at Stromboli Volcano, Italy

    Science.gov (United States)

    Chouet, B.; Saccorotti, G.; Dawson, P.; Martini, M.; Scarpa, R.; De Luca, G.; Milana, G.; Cattaneo, M.

    1999-01-01

    During September-October 1997, 21 three-component broadband seismometers were deployed on Stromboli Volcano at radial distances of 0.3-2.2 km from the active crater to investigate the source mechanisms of Strombolian explosions. In the 2-50 s band, the very-long period (VLP) signals associated with explosions are consistent with two stationary sources repeatedly activated in time. VLP particle motions are essentially linear and analyses of semblance and particle motions are consistent with a source centroid offset 300 m beneath and 300 m northwest of the active vents. Similar VLP waveforms are observed at all 21 stations, indicating that the seismograms are source-dominated. The VLP ground displacement response to each explosion may be qualitatively interpreted as: (1) pressurization of the conduit associated with the ascent of a slug of gas; (2) depressurization of the conduit in response to mass withdrawal during the eruption; and (3) repressurization of the conduit associated with the replenishment of the source with fluid. Copyright 1999 by the American Geophysical Union.

  19. [Electrochemical methods of control of iodine contents in drinks].

    Science.gov (United States)

    Zakharova, E A; Slepchenko, G B; Kolpakova, E Iu

    2001-01-01

    The simple and express methods of determination of iodide ions (0.01-0.20 mg/decimeter3) in iodine-enriched drinks by potentiometry and inversion voltamperometry were developed. The studies on influencing a storage time hermetically packaged carbonated beverages, a storage time of the depressurized drinks, stuff of ware on the contents of iodine in drinks are held.

  20. Experimental investigations on the deposition and remobilization of aerosol particles in turbulent flows

    International Nuclear Information System (INIS)

    Barth, Thomas

    2014-01-01

    Aerosol particle deposition and resuspension experiments in turbulent flows were performed to investigate the complex particle transport phenomena and to provide a database for the development and validation of computational fluid dynamics (CFD) codes. The background motivation is related to the source term analysis of an accidental depressurization scenario of a High Temperature Reactor (HTR). During the operation of former HTR pilot plants, larger amounts of radio-contaminated graphite dust were found in the primary circuit. This dust most likely arose due to abrasion between the graphitic core components and was deposited on the inner wall surfaces of the primary circuit. In case of an accident scenario, such as a depressurization of the primary circuit, the dust may be remobilized and may escape the system boundaries. The estimation of the source term being discharged during such a scenario requires fundamental knowledge of the particle deposition, the amount of contaminants per unit mass as well as the resuspension phenomena. Nowadays, the graphite dust distribution in the primary circuit of an HTR can be calculated for stationary conditions using one-dimensional reactor system codes. However, it is rather unknown which fraction of the graphite dust inventory may be remobilized during a depressurization of the HTR primary circuit. Two small-scale experimental facilities were designed and a set of experiments was performed to investigate particle transport, deposition and resuspension in turbulent flows. The facility design concept is based on the fluid dynamic downscaling of the helium pressure boundary in the HTR primary circuit to an airflow at ambient conditions in the laboratory. The turbulent flow and the particles were recorded by high-resolution, non-invasive imaging techniques to provide a spatio-temporal insight into the particle transport processes. The different investigations of this thesis can be grouped into three categories. Firstly, the

  1. Radon mitigation choices in the United States - a comparison of private and public sector developments

    International Nuclear Information System (INIS)

    Leovic, K.W.; Sanchez, D.C.; Craig, A.B.

    1988-01-01

    The paper compares private and public sector developments relating to radon mitigation in the US. In response to elevated levels of radon in many houses throughout the U.S., the Federal and State governments and the private sector have undertaken many varied mitigation and public information efforts resulting in a range of radon reduction installations in the U.S. Government sponsored research and development of radon mitigation techniques in single-family detached houses is compared with radon mitigation approaches undertaken by the private sector. An attempt is also made to compare premitigation diagnostics, the performance of mitigation installations, and the cost of the installations. The information collected indicates that sub-slab ventilation is the most common technique in both public and private sector radon mitigation; however, air-to-air heat exchangers and sealing are employed more frequently in the private sector than in government sponsored mitigation. (author)

  2. Cost effectiveness of radon mitigation in Canada

    International Nuclear Information System (INIS)

    Letourneau, E.G.; Krewski, D.; Zielinski, J.M.; McGregor, R.G.

    1992-01-01

    This paper examines the cost effectiveness of comprehensive strategies for reducing exposure to radon gas in indoor air in Canadian homes. The analysis is conducted within the context of a general framework for risk management programme evaluation which includes well-known evaluation techniques such as cost effectiveness and cost-benefit analyses as special cases. Based on this analysis, it is clear that any comprehensive programme to reduce exposure to environmental radon will be extremely expensive, and may not be justifiable in terms of health impact, particularly when considered in relation to other public health programmes. Testing of homes at the point of sale and installing sub-slab suction equipment to reduce exposure to indoor radon where necessary appears to be a relatively cost-effective radon mitigation strategy. In general, radon mitigation was found to be most cost effective in cities with relatively high levels of radon. (author)

  3. Parametric modelling of temporal variations in radon concentrations in homes

    International Nuclear Information System (INIS)

    Revzan, K.L.; Turk, B.H.; Harrison, J.; Nero, A.V.; Sextro, R.G.

    1988-01-01

    The 222 Rn concentrations in the living area, the basement, and the undelying soil of a New Jersey home have been measured at half-hour intervals over the course of a year, as have indoor and outdoor temperatures, wind speed and direction, and indoor-outdoor and basement-subslab pressures; in addition, periods of furnace opration have been logged. We generalize and extend an existing radon entry model in order to demonstrate the dependence of the radon concentration on the environmental variales and the extent of furnace use. The model contains parameters which are dependent on geological and structural factors which have not been measured or otherwise determined; statistical methods are used to find the best values of the parameters. The non-linear regression of the model predictions (over time) on the measured living area radon concentrations yields an R/aup 2/ of 0.88. 9 refs., 2 figs

  4. Radon reduction in wood foundation system

    International Nuclear Information System (INIS)

    Clark, R.J.

    1990-01-01

    Radon, an issue of growing concern to the building industry. Silently, invisibly, it invades existing structures as it will future foundation structures. This paper addresses the nature and causes of radon, and cost-effective prevention and retrofit techniques used for wood foundation systems. Radon also can enter homes with foundations that use the under-floor as an air distribution system. These building practices will be shown; even materials used in construction may release radon, for example, this may be a problem in a house that has a solar heating system in which its heat is stored in large beds of stone. Stone is most often used in wood foundation construction. The common radon entry points will be looked at, and the latest prevention techniques will be illustrated, such as natural and forced ventilation, sealing major radon sources and entry routes, and sub-slab and sump crock ventilations

  5. An Approach to the Flammability Testing of Aerospace Materials

    Science.gov (United States)

    Hirsch, David B.

    2012-01-01

    Presentation reviews: (1) Current approach to evaluation of spacecraft materials flammability (2) The need for and the approach to alternative routes (3) Examples of applications of the approach recommended a) Crew Module splash down b) Crew Module depressurization c) Applicability of NASA's flammability test data to other sample configurations d) Applicability of NASA's ground flammability test data to spacecraft environments

  6. Experiment data report for semiscale Mod-1 test S-04-2 (baseline ECC test)

    International Nuclear Information System (INIS)

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

    1976-09-01

    Recorded test data are presented for Test S-04-2 of the Semiscale Mod-1 Baseline ECC test series. This test is among Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-2 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using emergency core coolant injection parameters based on downcomer volume scaling. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such sime that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core coolant injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-04-2 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

  7. MHTGR inherent heat transfer capability

    International Nuclear Information System (INIS)

    Berkoe, J.M.

    1992-01-01

    This paper reports on the Commercial Modular High Temperature Gas-Cooled Reactor (MHTGR) which achieves improved reactor safety performance and reliability by utilizing a completely passive natural convection cooling system called the RCCS to remove decay heat in the event that all active cooling systems fail to operate. For the highly improbable condition that the RCCS were to become non-functional following a reactor depressurization event, the plant would be forced to rely upon its inherent thermo-physical characteristics to reject decay heat to the surrounding earth and ambient environment. A computational heat transfer model was created to simulate such a scenario. Plant component temperature histories were computed over a period of 20 days into the event. The results clearly demonstrate the capability of the MHTGR to maintain core integrity and provide substantial lead time for taking corrective measures

  8. Switchable and Tunable Aerodynamic Drag on Cylinders

    Science.gov (United States)

    Guttag, Mark; Lopéz Jiménez, Francisco; Upadhyaya, Priyank; Kumar, Shanmugam; Reis, Pedro

    We report results on the performance of Smart Morphable Surfaces (Smporhs) that can be mounted onto cylindrical structures to actively reduce their aerodynamic drag. Our system comprises of an elastomeric thin shell with a series of carefully designed subsurface cavities that, once depressurized, lead to a dramatic deformation of the surface topography, on demand. Our design is inspired by the morphology of the giant cactus (Carnegiea gigantea) which possesses an array of axial grooves, thought to help reduce aerodynamic drag, thereby enhancing the structural robustness of the plant under wind loading. We perform systematic wind tunnel tests on cylinders covered with our Smorphs and characterize their aerodynamic performance. The switchable and tunable nature of our system offers substantial advantages for aerodynamic performance when compared to static topographies, due to their operation over a wider range of flow conditions.

  9. Radon reduction in crawl-space houses

    International Nuclear Information System (INIS)

    Osborne, M.C.; Moore, D.G.; Southerlan, R.E.; Brennan, T.; Pyle, B.E.

    1989-01-01

    This paper gives results of an EPA study of radon-mitigation alternatives for crawl space houses in several houses in Nashville, TN. Application of one of these alternative mitigation options, suction under a polyethylene membrane, has been successful in significantly reducing radon levels in both the crawl space and the house. The large radon concentrations measured under unvented plastic ground covers and the moisture barriers found in many crawl spaces can act as radon-rich reservoirs capable of contaminating a crawl space and house during periods of depressurization. With the exhaust components of the mitigation system in place, radon levels below the plastic decreased by more than 95% under both passive and active suction conditions. Based on the study, the design of a cost-effective subplastic suction passive radon mitigation system for crawl spaces seems promising

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

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, G.

    1995-08-01

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

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

    International Nuclear Information System (INIS)

    Jacobs, G.

    1995-01-01

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

  12. Studies on air ingress for pebble bed reactors

    International Nuclear Information System (INIS)

    Moore, R.L.; Oh, C.H.; Merrill, B.J.; Petti, D.A.

    2002-01-01

    A loss-of-coolant accident (LOCA) has been considered a critical event for helium-cooled pebbled bed reactors. Following helium depressurization, it is anticipated that unless countermeasures are taken air will enter the core through the break and then by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure and graphite pebbles. Thus, without any mitigating features a LOCA will lead to an air ingress event. The INEEL is studying such an event with two well-respected light water reactor transient response codes: RELAP5/ATHENA and MELCOR. To study the degree of graphite oxidation occurring due to an air ingress event, a MELCOR model of a reference pebble bed design was constructed. A modified version of MELCOR developed at INEEL, which includes graphite oxidation capabilities, and molecular diffusion of air into helium was used for these calculations. Results show that the lower reflector graphite consumes all of the oxygen before reaching the core. The results also show a long time delay between the time that the depressurization phase of the accident is over and the time that natural circulation air through the core occurs. (author)

  13. Heat removal tests for pressurized water reactor containment spray by largescale facility

    International Nuclear Information System (INIS)

    Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

    1983-01-01

    Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

  14. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  16. Particle deposition and resuspension in gas-cooled reactors—Activity overview of the two European research projects THINS and ARCHER

    Energy Technology Data Exchange (ETDEWEB)

    Barth, T., E-mail: t.barth@hzdr.de [Institute of Fluid Dynamics, Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany); Lecrivain, G. [Institute of Fluid Dynamics, Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany); Jayaraju, S.T. [Nuclear Research and Consultancy Group (NRG), 1755ZG Petten (Netherlands); Hampel, U. [Institute of Fluid Dynamics, Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany); AREVA Endowed Chair of Imaging Techniques in Energy and Process Engineering, Technische Universität Dresden, 01062 Dresden (Germany)

    2015-08-15

    Highlights: • A summary on particle deposition and resuspension experiments is provided. • Similarities between single and multilayer particle deposits are found. • Numerical models for simulation of particle deposits are successfully developed. - Abstract: The deposition and resuspension behaviour of radio-contaminated aerosol particles is a key issue for the safety assessment of depressurization accidents of gas-cooled high temperature reactors. Within the framework of two European research projects, namely Thermal Hydraulics of Innovative Nuclear Systems (THINS) and Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D (ARCHER), a series of investigations was performed to investigate the transport, the deposition and the resuspension of aerosol particles in turbulent flows. The experimental and numerical tests can be subdivided into four different parts: (1) Monolayer particle deposition, (2) Monolayer particle resuspension, (3) Multilayer particle deposition and (4) Multilayer particle resuspension. The experimental results provide a new insight into the formation and removal of aerosol particle deposits in turbulent flows and are used for the development and validation of numerical procedures in gas-cooled reactors. Good agreement was found between the numerical and the experimental results.

  17. Annual meeting on nuclear technology 1980. Technical meeting: Shock wave propagation processes

    International Nuclear Information System (INIS)

    1980-01-01

    The papers deal with shock wave propagation processes in LWR-type reactors (licencing procedure) in cases of ruptures in vessels, in pipes and high-pressure valves, in case of loss of coolant accidents with dynamic structure coupling or fluid structure interactions in the reactor core jacket and the fuel rods, as well as with the stresses placed on reactor pressure vessel fittings by depressurization waves. (DG) [de

  18. Fuel elements and fuel element materials. Experimental facilities for fission products lift-off tests

    International Nuclear Information System (INIS)

    Blanchard, R.J.; Veyrat, J.F.

    1978-01-01

    One of the hypothetical accidents on the HTGR primary cooling circuits is the failure of a circuit resulting in a depressurization in the primary loops of the reactor. There is a risk of release of fission products in relation to the size of the failure. Experimental facilities for HTGR tests were developed: an in pile helium loop Comedie and an out of pile helium loop

  19. Radon mitigation choices in the United States: A comparison of private- and public-sector developments

    International Nuclear Information System (INIS)

    Witter, K.A.; Sanchez, D.C.; Craig, A.B.

    1988-01-01

    This paper compares private- and public-sector developments relating to radon mitigation in the U.S. In response to elevated radon levels in many U.S. houses, the Federal and State governments and the private sector have undertaken many varied mitigation and public-information efforts resulting in a range of radon reduction installations in the U.S. Government-sponsored research and development of radon-mitigation techniques in single-family detached houses is compared with radon mitigation approaches undertaken by the private sector. An attempt is also made to compare premitigation diagnostics, the application of mitigation techniques to various substructure types, the performance of mitigation installations, and the cost of the installations. The information collected indicates that subslab ventilation is the most common technique in both public- and private-sector radon mitigation; however, air-to-air heat exchangers and sealing are employed more frequently in the private sector than in government-sponsored mitigation

  20. A detailed evaluation of the individual health benefits arising in a domestic property following radon remediation - a case-study in Northamptonshire, U.K

    International Nuclear Information System (INIS)

    Denman, A.R.; Briggs, D.J.; Allison, C.C.; Groves-Kirkby, C.J.; Phillips, P.S.; Crockett, R.G.M.

    2008-01-01

    Radon gas occurs naturally in the environment with variable distribution, concentrating sufficiently in the built environment in some areas to pose a public health risk. Radon levels can be successfully reduced in affected buildings, and large-scale remediation programmes have been justified in terms of accrued costs and benefits. We present results from a house where radon levels in the main living-room and master bedroom were monitored on an hourly basis over extended periods before and after radon remediation by sub-slab depressurisation. These results were combined with results from a recent occupancy survey to estimate the health impact on occupants spending varying times in the home. Prior to remediation, mean hourly radon exposure is moderately linearly correlated (R 2 = 0.66-0.78) with time spent in the house. Following remediation, correlation is significantly enhanced (R 2 = 0.91-0.95), but the exposure reduction of an occupant following remediation is less than that predicted using the NRPB protocol

  1. Review of selected state-of-the-art applications of diagnostic measurements for radon-mitigation planning. Report for April 1986-June 1987

    International Nuclear Information System (INIS)

    Hubbard, L.M.; Harrje, D.T.; Gadsby, K.J.; Sanchez, D.C.; Turk, B.H.

    1987-09-01

    Since late-1984, EPA's AEERL has supported a program to develop and demonstrate radon-mitigation techniques for single-family detached dwellings. As part of the program, projects have been started, directed at developing and demonstrating the use of diagnostic measurements in all phases of the radon-mitigation process. Diagnostic measurements are used to assess: (1) the radon sources strengths, variability, and locations; and, (2) radon transport to the house and its entry and distribution in the house as influenced by environmental, house characteristics, and occupancy factors. The diagnostic measurements reported include: (1) soil-gas grab sampling; (2) communication (air flow or pressure-field extension) tests; (3) whole house infiltration; (4) differential pressure, (5) gamma radiation; and, (6) radon flux. The paper concludes that the above selected diagnostic measurements were especially useful in characterizing houses with indoor radon problems attributable to soil-gas-borne radon that may be amenable to mitigation through the use of subslab ventilation

  2. A detailed evaluation of the individual health benefits arising in a domestic property following radon remediation - a case-study in Northamptonshire, U.K

    Energy Technology Data Exchange (ETDEWEB)

    Denman, A.R. [Medical Physics Department, Northampton General Hospital NHS Trust, Cliftonville, Northampton NN1 5BD (United Kingdom); School of Health, University of Northampton, Boughton Green Road, Northampton NN2 7AL (United Kingdom)], E-mail: tony.denman@northampton.ac.uk; Briggs, D.J. [Department of Epidemiology and Public Health, Imperial College of Science, Technology and Medicine, London W1 2PG (United Kingdom)], E-mail: d.briggs@imperial.ac.uk; Allison, C.C. [Medical Physics Department, Northampton General Hospital NHS Trust, Cliftonville, Northampton NN1 5BD (United Kingdom)], E-mail: claire.allison@ngh.nhs.uk; Groves-Kirkby, C.J. [Medical Physics Department, Northampton General Hospital NHS Trust, Cliftonville, Northampton NN1 5BD (United Kingdom)], E-mail: chris.groves-kirkby@ngh.nhs.uk; Phillips, P.S. [School of Applied Sciences, The University of Northampton, Boughton Green Road, Northampton NN2 7AL (United Kingdom)], E-mail: paul.phillips@northampton.ac.uk; Crockett, R.G.M. [School of Applied Sciences, University of Northampton, Boughton Green Road, Northampton NN2 7AL (United Kingdom)], E-mail: robin.crockett@northampton.ac.uk

    2008-07-15

    Radon gas occurs naturally in the environment with variable distribution, concentrating sufficiently in the built environment in some areas to pose a public health risk. Radon levels can be successfully reduced in affected buildings, and large-scale remediation programmes have been justified in terms of accrued costs and benefits. We present results from a house where radon levels in the main living-room and master bedroom were monitored on an hourly basis over extended periods before and after radon remediation by sub-slab depressurisation. These results were combined with results from a recent occupancy survey to estimate the health impact on occupants spending varying times in the home. Prior to remediation, mean hourly radon exposure is moderately linearly correlated (R{sup 2} = 0.66-0.78) with time spent in the house. Following remediation, correlation is significantly enhanced (R{sup 2} = 0.91-0.95), but the exposure reduction of an occupant following remediation is less than that predicted using the NRPB protocol.

  3. Radon-mitigation choices in the United States: a comparison of private- and public-sector developments

    International Nuclear Information System (INIS)

    Witter, K.A.; Sanchez, D.C.; Craig, A.B.

    1988-01-01

    This paper compares private- and public-sector developments relating to radon mitigation in the U.S. In response to elevated radon levels in many U.S. houses, the Federal and State governments and the private sector have undertaken many varied mitigation and public-information efforts resulting in a range of radon-reduction installations in the U.S. Government-sponsored research and development of radon-mitigation techniques in single-family detached houses is compared with radon mitigation approaches undertaken by the private sector. An attempt is also made to compare premitigation diagnostics, the application of mitigation techniques to various substructure types, the performance of mitigation installations, and the cost of the installations. The information collected indicates that subslab ventilation is the most common technique in both public and private sector radon mitigation; however, air-to-air heat exchangers and sealing are employed more frequently in the private sector than in government-sponsored mitigation

  4. Gas Hydrates Research Programs: An International Review

    Energy Technology Data Exchange (ETDEWEB)

    Jorge Gabitto; Maria Barrufet

    2009-12-09

    Gas hydrates sediments have the potential of providing a huge amount of natural gas for human use. Hydrate sediments have been found in many different regions where the required temperature and pressure conditions have been satisfied. Resource exploitation is related to the safe dissociation of the gas hydrate sediments. Basic depressurization techniques and thermal stimulation processes have been tried in pilot efforts to exploit the resource. There is a growing interest in gas hydrates all over the world due to the inevitable decline of oil and gas reserves. Many different countries are interested in this valuable resource. Unsurprisingly, developed countries with limited energy resources have taken the lead in worldwide gas hydrates research and exploration. The goal of this research project is to collect information in order to record and evaluate the relative strengths and goals of the different gas hydrates programs throughout the world. A thorough literature search about gas hydrates research activities has been conducted. The main participants in the research effort have been identified and summaries of their past and present activities reported. An evaluation section discussing present and future research activities has also been included.

  5. Risk evaluation of accident management strategies

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, A.

    1992-01-01

    The use of Probabilistic Risk Assessment (PRA) methods to evaluate accident management strategies in nuclear power plants discussed in this paper. The PRA framework allows an integrated evaluation to be performed to give the full implications of a particular strategy. The methodology is demonstrated for a particular accident management strategy, intentional depressurization of the reactor coolant system to avoid containment pressurization during the ejection of molten debris at vessel breach

  6. INEL design studies in support of the Westinghouse EPRI small plant study

    International Nuclear Information System (INIS)

    Burtt, J.D.; Kullberg, C.M.

    1986-03-01

    In support of the design effort of a Westinghouse EPRI small plant study, several analyses were performed at the Idaho National Engineering Laboratory. An analysis was performed to study fuel behavior under conditions of a limiting flow coastdown transient. Depressurization capabilities for the reactor coolant system were studied. The post-accident heat removal for the current containment design was studied. The results of all three studies are reported. 31 figs

  7. Mitigation of houses with extremely high indoor radon concentrations

    International Nuclear Information System (INIS)

    Jiranek, M.; Neznal, M.

    2006-01-01

    Full text of publication follows: The paper reports on the experience of the Czech Technical University in dealing with mitigation of houses in which unusually high indoor radon concentrations were found. The whole process of remediation is illustrated by example of an old single-family house that was built in the area formed by highly permeable soils with high radon content in the soil air. T he house has a small cellar located under 1/5 of the ground floor area. Two types of floors, i.e. timber floors and cracked concrete slabs were found in the house. As a result of extremely high radon concentration in the sub-floor region (up to 600 kBq/m 3 ) and leaky structures in contact with soil, radon concentrations around 100 kBq/m 3 in the cellar and up to 60 kBq/m 3 in the living rooms on the ground floor were measured prior to mitigation. Mitigation measures that were carried out in the house consist of reconstruction of timber floors and installation of active soil depressurization. Timber floors were replaced with concrete slab fitted with damp proof membrane, thermal insulation and floor covering. The soil depressurization system was made up of two sections. The first section is composed of the network of perforated pipes inserted in the drainage layer placed under the new floors and four perforated tubes drilled under the existing floors. The soil air from this section is extracted by means of a roof fan installed at the top of the vertical exhaust pipe running inside the living space and terminating above the roof. The second section was designed to withdraw by means of a small fan radon-laden air from the filling in the floor above the cellar and from perforated tubes drilled into the sub-floor region under the rooms adjacent to the cellar. It serves also for the active ventilation of the cellar. Pressure, temperature and radon concentration sensors were installed into the drainage layer during the reconstruction of floors to record variations in these

  8. Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

    International Nuclear Information System (INIS)

    Hanson, G.H.

    1978-01-01

    The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. LTR 113-47 has shown that the LOFT ECCS can be safely bypassed or disabled when the total core power does not exceed 25 kW. A modified policy involves disabling the automatic actuation of the LOFT ECCS, but still retaining the manual activation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 70 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS

  9. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    International Nuclear Information System (INIS)

    Ball, S.J.

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR

  10. High-temperature reactor developments in the Netherlands

    International Nuclear Information System (INIS)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van.

    1996-01-01

    The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.)

  11. Building America Case Study: Simplified Air Distribution, Desuperheaters, and Sub-Slab Geothermal Heat Exchangers, Pittsburgh, Pennsylvania

    Energy Technology Data Exchange (ETDEWEB)

    N. Mittereder, A. Poerschke

    2017-05-01

    This report presents a cold-climate project that examines an alternative approach to ground source heat pump (GSHP) ground loop design. The innovative ground loop design is an attempt to reduce the installed cost of the ground loop heat exchange portion of the system by containing the entire ground loop within the excavated location beneath the basement slab.

  12. Building America Case Study: Simplified Air Distribution, Desuperheaters, and Sub-Slab Geothermal Heat Exchangers, Pittsburgh, Pennsylvania

    Energy Technology Data Exchange (ETDEWEB)

    2017-06-08

    This report presents a cold-climate project that examines an alternative approach to ground source heat pump (GSHP) ground loop design. The innovative ground loop design is an attempt to reduce the installed cost of the ground loop heat exchange portion of the system by containing the entire ground loop within the excavated location beneath the basement slab.

  13. Analysis of effect of late water injection on RCS repressurization

    International Nuclear Information System (INIS)

    Tao Jun; Cao Xuewu

    2011-01-01

    Effect of late water injection on RCS repressurization during high pressure severe accident sequence in a typical PWR was analyzed. As the results shown, late water injection could increase RCS pressure when RPV failed without RCS passive depressurization. Especially in the condition of opening one PORV, RCS pressure could reach high pressure limit when RPV failed and the risk of HPME and DCH was dramatically increased. Integrity of containment could be threatened. However, in the condition of RCS passive depressurization induced by pressurizer surge line creep failure, RCS pressure could be decreased to very low level even only one PORV was opened and two trains of emergency core cooling were implemented. The risk of HPME and DCH was eliminated. The more PORVs were opened, the faster accident progression was and the earlier RPV failed. RCS pressure was a little higher when PRV failed if two trains of emergency core cooling was implemented comparing with the condition with only one train of emergency core cooling. However the time of RPV failure was obviously delayed. From the point of delaying RPV failure and preventing containment early failure of view, the optimized late water injection was opening three PORVs and implementing two trains of emergency core cooling. (authors)

  14. Test methods for determining asphaltene stability in crude oils

    Energy Technology Data Exchange (ETDEWEB)

    Asomaning, S. [Baker Petrolite, Sugar Land, TX (United States)

    2001-07-01

    The development of test methods for the determination of the stability of asphaltenes in crude oils was rendered necessary, due to the high cost of remediating asphaltene deposition in harsh production environments, namely the underwater systems in offshore deepwater. The Oliensis Spot Test, two saturates, aromatics, resins and asphaltenes (SARA)-based screens (the Colloidal Instability Index and Asphaltene-Resin ratio), a solvent titration method with near infrared radiation (NIR) solids detection, and live oil depressurization were used for the purposes of this study, to predict the stability of asphaltenes in crude oils with different API gravity. A complete description of the test methods was provided, and the experimental data obtained as a result was presented. Correlation with data on the deposition histories of the oils was used to validate the experimental stability data. The author discussed the effectiveness of the different tests for the prediction of the stability of asphaltenes in crude oils. The prediction of a crude oil's propensity towards asphaltene precipitation was more accurate with the Colloidal Instability Index and the solvent titration method. Live oil depressurization proved to be very effective for the prediction of the stability of asphaltenes for light oils, where most stability tests fail. tabs., 31 figs.

  15. 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

  16. Space suit glove design with advanced metacarpal phalangeal joints and robotic hand evaluation.

    Science.gov (United States)

    Southern, Theodore; Roberts, Dustyn P; Moiseev, Nikolay; Ross, Amy; Kim, Joo H

    2013-06-01

    One area of space suits that is ripe for innovation is the glove. Existing models allow for some fine motor control, but the power grip--the act of grasping a bar--is cumbersome due to high torque requirements at the knuckle or metacarpal phalangeal joint (MCP). This area in particular is also a major source of complaints of pain and injury as reported by astronauts. This paper explores a novel fabrication and patterning technique that allows for more freedom of movement and less pain at this crucial joint in the manned space suit glove. The improvements are evaluated through unmanned testing, manned testing while depressurized in a vacuum glove box, and pressurized testing with a robotic hand. MCP joint flex score improved from 6 to 6.75 (out of 10) in the final glove relative to the baseline glove, and torque required for flexion decreased an average of 17% across all fingers. Qualitative assessments during unpressurized and depressurized manned testing also indicated the final glove was more comfortable than the baseline glove. The quantitative results from both human subject questionnaires and robotic torque evaluation suggest that the final iteration of the glove design enables flexion at the MCP joint with less torque and more comfort than the baseline glove.

  17. MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

    International Nuclear Information System (INIS)

    Choi, Wonjun; Jeon, Joongoo; Kim, Nam Kyung; Kim, Sung Joong

    2017-01-01

    The representative example of the severe accident studies using the severe accident code is investigation of effectiveness of developed severe accident management (SAM) strategy considering the positive and adverse effects. In Korea, some numerical studies were performed to investigate the SAM strategy using various severe accident codes. Seo et.al performed validation of RCS depressurization strategy and investigated the effect of severe accident management guidance (SAMG) entry condition under small break loss of coolant accident (SBLOCA) without safety injection (SI), station blackout (SBO), and total loss of feed water (TLOFW) scenarios. The SGTR accident with the sequential mitigation actions according to the flow chart of SAMG was simulated by the MELCOR 1.8.6 code. Three scenariospreventing the RPV failure were investigated in terms of fission product release, hydrogen risk, and the containment pressure. Major conclusions can be summarized as follows: (1) According to the flow chart of SAMG, RPV failure can be prevented depending on the method of RCS depressurization. (2) To reduce the release of fission product during the injecting into SGs, a temporary opening of SDS before the injecting into SGs was suggested. These modified sequences of mitigation actions can reduce the release of fission product and the adverse effect of SDS.

  18. ATWS analysis for total loss of feedwater sequence in UCN 3 and 4

    International Nuclear Information System (INIS)

    Park, S. H.; Song, Y. M.; Kim, D. H.; Kim, S. D.; Park, S. Y.

    1999-01-01

    ATWS is a trip-failed severe accident initiated from the transients like a turbine trip, a control bank withdrawal, and a loss of feedwater which are expected to occur comparatively often (one or two occurrences / year). In this study, an ATWS sequence in Ulchin 3 and 4 is analyzed and the effects of the important systems are studied for accident management purpose using a MIDAS/PK computer code. The MIDAS/PK code has been developed via coupling a point kinetics module with the MELCOR code. The code calculates a primary peak pressure of about 24MPa at 240 seconds for the ATWS initiated by a TLOF (Total Loss of Feedwater) transient. Along with the basic ATWS analysis, several sensitivity runs are performed. From these, the turbines and the safety depressurization system (SDS) are judged to be important. The turbine trip resulting in a loss of offsite power and a RCP trip, degrades primary heat transfer to the secondary sides, and in turn, increases primary coolant temperature which reduces the reactor power due to the negative moderator temperature coefficient. Manual operation of SDS has an effect to lower the primary peak pressure considerably via supplementary depressurization in addition to the PORVs

  19. STEAM GENERATOR TUBE INTEGRITY ANALYSIS OF A TOTAL LOSS OF ALL HEAT SINKS ACCIDENT FOR WOLSONG NPP UNIT 1

    Directory of Open Access Journals (Sweden)

    HEOK-SOON LIM

    2014-02-01

    Full Text Available A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS and the steam generator (SG secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  20. Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo [Korea Htydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Seoungrae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

    2014-02-15

    A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  1. Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

    International Nuclear Information System (INIS)

    Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo; Kim, Seoungrae

    2014-01-01

    A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident

  2. Analysis of cold leg LOCA with failed HPSI by means of integrated safety assessment methodology

    International Nuclear Information System (INIS)

    Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayorga, J.

    2014-01-01

    Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties

  3. Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

    2012-06-01

    The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

  4. Assessment of Gas Production Potential from Hydrate Reservoir in Qilian Mountain Permafrost Using Five-Spot Horizontal Well System

    Directory of Open Access Journals (Sweden)

    Yun-Pei Liang

    2015-09-01

    Full Text Available The main purpose of this study is to investigate the production behaviors of gas hydrate at site DK-2 in the Qilian Mountain permafrost using the novel five-spot well (5S system by means of numerical simulation. The whole system is composed of several identical units, and each single unit consists of one injection well and four production wells. All the wells are placed horizontally in the hydrate deposit. The combination method of depressurization and thermal stimulation is employed for hydrate dissociation in the system. Simulation results show that favorable gas production and hydrate dissociation rates, gas-to-water ratio, and energy ratio can be acquired using this kind of multi-well system under suitable heat injection and depressurization driving forces, and the water production rate is manageable in the entire production process under current technology. In addition, another two kinds of two-spot well (2S systems have also been employed for comparison. It is found that the 5S system will be more commercially profitable than the 2S configurations for gas production under the same operation conditions. Sensitivity analysis indicates that the gas production performance is dependent on the heat injection rate and the well spacing of the 5S system.

  5. Plugging Effects on Depressurization Time in Dry Storage Containers with Pinhole Breaches

    International Nuclear Information System (INIS)

    Casella, Andrew M.; LOYALKA, SUDARSHAN K.; Hanson, Brady D.

    2006-01-01

    As continuation on previous work, we now examine the effect that aerosol deposition may have on plugging pinhole breaches in spent fuel containers. A model is developed considering only diffusive settling

  6. Diffusion modeling of fission product release during depressurized core conduction cooldown conditions

    International Nuclear Information System (INIS)

    Martin, R.C.

    1991-01-01

    A simple model for diffusion through the silicon carbide layer of TRISO particles is applied to the data for accident condition testing of fuel spheres for the High-Temperature Reactor program of the Federal Republic of Germany (FRG). Categorization of sphere release of 137 Cs based on fast neutron fluence permits predictions of release with an accuracy comparable to that of the US/FRG accident condition fuel performance model. Calculations are also performed for 85 Kr, 90 Sr, and 110m Ag. Diffusion of cesium through SiC suggests that models of fuel failure should consider fuel performance during repeated accident condition thermal cycling. Microstructural considerations in models of fission product release are discussed. The neutron-induced segregation of silicon within the SiC structure is postulated as a mechanism for enhanced fission product release during accident conditions. As oxygen-enhanced SiC decomposition mechanism is also discussed. (author). 12 refs, 11 figs, 2 tabs

  7. Study of steam condensation in SG tubes with large amount of nitrogen to be accumulated

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S.A.; Sitnik, Y.K. [EDO Gidropress, Podolsk (Russian Federation)

    1997-12-31

    The effect of nitrogen during SG heat transfer under SBLOCA conditions have been studied. Depressurization of the primary side leads to release of nitrogen dissolved in the hydroaccumulator water. Nitrogen can accumulate in SGs and affect adversely heat transfer under reflux condenser conditions. The main objective of the study has been to show that nitrogen does not prevent heat transfer in SGs of the VVER-640 which is reactor plant of new generation. (orig.).

  8. Pitot tube and drag body measurements in transient steam--water flows

    International Nuclear Information System (INIS)

    Fincke, J.R.; Deason, V.A.; Dacus, M.W.

    1979-01-01

    The use of full-flow drag devices and rakes of water-cooled Pitot tubes to measure the transient two-phase mass flow during loss-of-coolant experiments in pressurized water reactor (PWR) environments has been developed. Mass flow rate measurements have been obtained in high temperature and pressure environments, similar to PWRs, under transient conditions. Comparisons of the measured time integrated value of mass flow to the known system mass before depressurization are made

  9. Use of the supercritical fluid technology to prepare efficient nanocomposite foams for environmental protection purpose

    OpenAIRE

    Urbanczyk, Laetitia; Thomassin, Jean-Michel; Huynen, Isabelle; Alexandre, Michaël; Jérôme, Christine

    2009-01-01

    This work reports on the preparation of novel nanocomposite foams that are efficient broadband microwave absorbers. Carbon nanotubes are first successfully dispersed into PCL and PMMA by melt blending. Then, foaming is promoted by supercritical CO2 by depressurization. Regular cellular structures are obtained in both cases with cells size around 10-50µm. The electromagnetic interference (EMI) shielding efficiency of these materials are then evaluated and compared to the non-foamed nanocomposi...

  10. A case of death of the driver due to environmental asphyxia by liquid nitrogen leakage in the cabin of the car during a road accident

    Science.gov (United States)

    Raczkowska, Zuzanna; Samojłowicz, Dorota

    2013-01-01

    Nitrogen causes environmental asphyxia by displacing oxygen in the air leading to death. The study presents a case of a death of a driver death who was transporting flasks with liquid nitrogen that depressurized during an accident. The mechanism and cause of death were determined based on the result of the autopsy and histopathologic examination. The authors emphasize the relevance of accident scene inspection during establishing the cause of death in similar cases.

  11. Investigations on the propagation of free surface boiling in a vertical superheated liquid column

    International Nuclear Information System (INIS)

    Das, P.K.; Bhat, G.S.; Arakeri, V.H.

    1987-01-01

    Some experimental studies on boiling propagation in a suddenly depressurized superheated vertical liquid column are reported. The propagation velocity of this phase change has been measured using an optical method. This velocity is strongly dependent on liquid superheat, liquid purity and test section size. The measured velocities of less than 5 m s -1 are significantly lower than the sonic velocity. Present observations suggest that the dominant mechanism for boiling propagation is convection. (author)

  12. Simulation model of dynamical behaviour of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Planchard, J.

    1994-01-01

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

  13. Unconventional Fossil-Based Fuels. Economic and Environmental Trade-Offs

    Science.gov (United States)

    2008-01-01

    concern with using water for in situ operations is that the water is generally drawn from aquifers (fresh or saline) due to the location of operations...depressurization of freshwater aquifers , changes in groundwater levels, changes in underground water storage or flow due to voidage zones left by bitumen removal...More recently, Sasol converted a CTL facility to accept natural gas from Mozambique. In Qatar , a large GTL facility has recently begun operating, and a

  14. Test plan for In Situ Vitrification Engineering-Scale Test No. 6, EG ampersand G Idaho, Inc., Job Number 318230

    International Nuclear Information System (INIS)

    1991-03-01

    The objectives of the test included the effects of in situ vitrification on containerized sludge contained in a simulated randomly-disposed array. From this arrangement, the test results obtained the following data applicable to Idaho National Engineering Laboratory Large Field Testing: canister burst pressure and temperature, canister depressurization rate, melt encapsulation rate of the canister and the hood area plenum temperatures, pressures, compositional analyses, and flows as affected by gas releases. 10 figs., 1 tab

  15. Investigations on the propagation of free surface boiling in a vertical superheated liquid column

    Energy Technology Data Exchange (ETDEWEB)

    Das, P.K.; Bhat, G.S.; Arakeri, V.H.

    1987-04-01

    Some experimental studies on boiling propagation in a suddenly depressurized superheated vertical liquid column are reported. The propagation velocity of this phase change has been measured using an optical method. This velocity is strongly dependent on liquid superheat, liquid purity and test section size. The measured velocities of less than 5 m s/sup -1/ are significantly lower than the sonic velocity. Present observations suggest that the dominant mechanism for boiling propagation is convection.

  16. Study of steam condensation in SG tubes with large amount of nitrogen to be accumulated

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S A; Sitnik, Y K [EDO Gidropress, Podolsk (Russian Federation)

    1998-12-31

    The effect of nitrogen during SG heat transfer under SBLOCA conditions have been studied. Depressurization of the primary side leads to release of nitrogen dissolved in the hydroaccumulator water. Nitrogen can accumulate in SGs and affect adversely heat transfer under reflux condenser conditions. The main objective of the study has been to show that nitrogen does not prevent heat transfer in SGs of the VVER-640 which is reactor plant of new generation. (orig.).

  17. Containment hydrogen and atmosphere activity control to mitigate severe accidents in VVERs and Western PWRs. Design and status of implementation

    International Nuclear Information System (INIS)

    Feuerbach, R.

    2002-01-01

    For accident management nuclear power plants in Europe have been or will be back-fitted with supplementary systems for monitoring the containment hydrogen concentration, for the early removal and reduction of hydrogen and filtered venting systems to retain radioactive aerosols and iodine. The hydrogen monitoring system (HMS) provides the information of local H 2 concentration in the containment during DBA and severe accident situations. The new HMS contains of overall H 2 -sensors and is installed inside the confinement. It provides continuos information about the local and temporal distribution of hydrogen, reported directly to the Emergency Response Team in case of severe accident. The hydrogen Reduction System (HRS) consists of several Passive Autocatalytic Recombiners (PAR) located in several compartments in the containment. The number of PARs to be installed depends on the type of NPP, structure of containment and the investigated accident scenario e.g. DBA conditions - approx. 6 to 20 PARs; severe accident conditions - 20-60 PARs). In case of severe accident it does not need any operator actions. The Filtered Venting System (FVS) is is especially important for WWER-440/230 maintaining sub atmospheric pressure in the confinement. For severe accident the on-site Emergency Response Team has to take the necessary strategic decisions for containment depressurization via the FVS

  18. Performance Evaluation of SMART Passive Safety System for Small Break LOCA Using MARS Code

    International Nuclear Information System (INIS)

    Chun, Ji Han; Lee, Guy Hyung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

    2013-01-01

    SMART has significantly enhanced safety by reducing its core damage frequency to 1/10 that of a conventional nuclear power plant. KAERI is developing a passive safety injection system to replace the active safety injection pump in SMART. It consists of four trains, each of which includes gravity-driven core makeup tank (CMT) and safety injection tank (SIT). This system is required to meet the passive safety performance requirements, i.e., the capability to maintain a safe shutdown condition for a minimum of 72 hours without an AC power supply or operator action in the case of design basis accidents (DBAs). The CMT isolation valve is opened by the low pressurizer pressure signal, and the SIT isolation valve is opened at 2 MPa. Additionally, two stages of automatic depressurization systems are used for rapid depressurization. Preliminary safety analysis of SMART passive safety system in the event of a small-break loss-of-coolant accident (SBLOCA) was performed using MARS code. In this study, the safety analysis results of a guillotine break of safety injection line which was identified as the limiting SBLOCA in SMART are given. The preliminary safety analysis of a SBLOCA for the SMART passive safety system was performed using the MARS code. The analysis results of the most limiting SI line guillotine break showed that the collapsed liquid level inside the core support barrel was maintained sufficiently high above the top of core throughout the transient. This means that the passive safety injection flow from the CMT and SIT causes no core uncovery during the 72 hours following the break with no AC power supply or operator action, which in turn results in a consistent decrease in the fuel cladding temperature. Therefore, the SMART passive safety system can meet the passive safety performance requirement of maintaining the plant at a safe shutdown condition for a minimum of 72 hours without AC power or operator action for a representing accident of SBLOCA

  19. Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

    International Nuclear Information System (INIS)

    Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun; Sun, Xiaodong; Christensen, Richard N.; Oh, Chang H.

    2015-01-01

    Highlights: • A 1/8th geometric-scale test facility that models the VHTR hot plenum is proposed. • Geometric scaling analysis is introduced for VHTR to analyze air-ingress accident. • Design calculations are performed to show that accident phenomenology is preserved. • Some analyses include time scale, hydraulic similarity and power scaling analysis. • Test facility has been constructed and shake-down tests are currently being carried out. - Abstract: A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to depend largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time

  20. Verification of the computer code ATHLET in the framework of the external verification group ATHLET BETHSY test 9.3 - steam generator U-tube rupture with failure of the high pressure injection. Final report; Verifikation des ATHLET-Rechenprogramms im Rahmen der externen Verifikationsgruppe ATHLET BETHSY Test 9.3 - Heizrohrbruch mit Versagen der Hochdruck-Noteinspeisung. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E.; Schaefer, F. [Forschungszentrum Rossendorf e.V. (FZR) (Germany). Inst. fuer Sicherheitsforschung

    1998-08-01

    In the framework of the external validation of the thermalhydraulic code ATHLET MOD 1.1 CYCLE D, which is being developed by the GRS, post test analyses of two experiments were done, which were performed at the French integral test facility BETHSY. During test 9.3 the consequences of a steam generator U-tube rupture with failure of the high pressure injection and of the auxiliary feedwater supply were investigated. As accident management measures, the depressurization of the secondary sides, first of the two intact steam generators, then of the damaged steam generator and finally the primary depressurization by opening of the pressurizer valve were performed. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. The safety relevant statement could be reproduced. Deviations, which did not impose the general results, occurred concerning the break mass flow during the depressurization of the damaged steam generator and the description of the failure of the heat transfer to the damaged steam generator. Reasons are hardly to find, because these processes are highly complex. (orig.) [Deutsch] Im Rahmen der externen Validierung des von der Gesellschaft fuer Anlagen- und Reaktorsicherheit entwickelten Stoerfallcodes ATHLET, der in der Version Mod 1.1 Cycle D vorlag, wurden zwei Experimente nachgerechnet und analysiert, die an der franzoesischen Versuchsanlage BETHSY durchgefuehrt wurden. Im Test 9.3 werden die Konsequenzen untersucht, wenn bei einem Heizrohrbruch die Hochdruckeinspeisung sowie die Not-Speisewasserversorgung der Dampferzeuger versagen und nur die Druckspeicher sowie die Niederdruckeinspeisung zur Verfuegung stehen. Als Accident Management Massnahmen wurde die sekundaere Druckentlastung und schliesslich die primaere Entlastung ueber den Druckhalter untersucht. Die Analyse kommt zu dem Ergebnis, dass der Code ATHLET in der Lage ist, dieses komplexe Szenario recht gut zu beschreiben. Die

  1. Numerical simulation of gas hydrate exploitation from subsea reservoirs in the Black Sea

    Science.gov (United States)

    Janicki, Georg; Schlüter, Stefan; Hennig, Torsten; Deerberg, Görge

    2017-04-01

    Natural gas (methane) is the most environmental friendly source of fossil energy. When coal is replace by natural gas in power production the emission of carbon dioxide is reduced by 50 %. The vast amount of methane assumed in gas hydrate deposits can help to overcome a shortage of fossil energy resources in the future. To increase their potential for energy applications new technological approaches are being discussed and developed worldwide. Besides technical challenges that have to be overcome climate and safety issues have to be considered before a commercial exploitation of such unconventional reservoirs. The potential of producing natural gas from subsea gas hydrate deposits by various means (e. g. depressurization and/or carbon dioxide injection) is numerically studied in the frame of the German research project »SUGAR - Submarine Gas Hydrate Reservoirs«. In order to simulate the exploitation of hydrate-bearing sediments in the subsea, an in-house simulation model HyReS which is implemented in the general-purpose software COMSOL Multiphysics is used. This tool turned out to be especially suited for the flexible implementation of non-standard correlations concerning heat transfer, fluid flow, hydrate kinetics, and other relevant model data. Partially based on the simulation results, the development of a technical concept and its evaluation are the subject of ongoing investigations, whereby geological and ecological criteria are to be considered. The results illustrate the processes and effects occurring during the gas production from a subsea gas hydrate deposit by depressurization. The simulation results from a case study for a deposit located in the Black Sea reveal that the production of natural gas by simple depressurization is possible but with quite low rates. It can be shown that the hydrate decomposition and thus the gas production strongly depend on the geophysical properties of the reservoir, the mass and heat transport within the reservoir, and

  2. Development of gas-cooled fast reactor and its thermo-hydraulics

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi

    1977-10-01

    Development, thermo-hydraulics and safety of GCFR are reviewed. The Development of Gas-Cooled Fast Reactor (GCFR) utilizes helium technology of HTGR and fuel technology of LMFBR. The breeding ratio of GCFR will be larger than that of LMFBR by about 0.2. Features of GCFR are a fuel with roughened surface to raise the heat transfer and vent system for the pressure equalization in the fuel rod. Helium as coolant of GCFR is chemically stable and stays in the single phase. So, there is no fuel-coolant interaction unlike the case of LMFBR. Since the helium must be pressurized, possibility of a depressurization accident is not negligible. In the United States, a 300MWe demonstration plant program is about to start; the collaboration with European countries is now quite active in this field. Though the development of GCFR started behind that of LMFBR, GCFR is equally promising as a fast breeder reactor. When realized, it will present possibility of a choice between these two. (auth.)

  3. FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

    Energy Technology Data Exchange (ETDEWEB)

    Paul, D.D.

    1976-06-01

    FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm.

  4. FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

    International Nuclear Information System (INIS)

    Paul, D.D.

    1976-06-01

    FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm

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

    International Nuclear Information System (INIS)

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

    1981-06-01

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

  6. Development and industrial tests of the first LNG hydraulic turbine system in China

    OpenAIRE

    Jie Chen; Yihuai Hua; Qingbo Su; Xueli Wan; Zhenlin Li

    2016-01-01

    The cryogenic hydraulic turbine can be used to replace the conventional J–T valve for LNG or mixed refrigerant throttling and depressurization in a natural gas liquefaction plant. This advanced technology is not only to enhance the efficiency of the liquefaction plant, but to usher a new trend in the development of global liquefaction technologies. China has over 136 liquefaction plants, but the cryogenic hydraulic turbines have not been deployed in industrial utilization. In addition, these ...

  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. Radiation effluent suppression system

    International Nuclear Information System (INIS)

    Watanabe, Atsushi.

    1992-01-01

    In a radiation release suppression system upon accident, an electromotive valve, a pneumatic operation valve or a manual operation valve is disposed to gas ventilation pipelines which are extended from both of a dry well and a wet well of a reactor container to a stuck. In addition, a combination filter of a metal fiber filter made of stainless steel etc. and an activated carbon fiber filter is disposed in the midway of pipelines in a reactor building. With such a constitution, the inside of the container can be depressurized (prevention of ruptures) and the amount of radioactive substances released to circumstances is remarkably suppressed by the effect of radioactive substance capturing effect of the metal fiber filter made of stainless steel etc. disposed in the vent pipe in the container and a radioactive substance capturing effect by the combination filter of the metal fiber filter made of stainless steel, etc. and the activated carbon fiber filter disposed in the gas ventilation pipelines even upon occurrence of an accident exceeding design basis. Systems can be simplified and minimized, and cost down can also be attained. (N.H.)

  9. Control of indoor radon and radon progeny concentrations

    International Nuclear Information System (INIS)

    Sextro, R.G.

    1985-05-01

    There are three general categories of techniques for the control of radon and radon progeny concentrations in indoor air - restriction of radon entry, reduction of indoor radon concentrations by ventilation or air cleaning, and removal of airborne radon progeny. The predominant radon entry process in most residences appears to be pressure driven flow of soil gas through cracks or other openings in the basement, slab, or subfloor. Sealing these openings or ventilation of the subslab or subfloor space are methods of reducing radon entry rates. Indoor radon concentrations may be reduced by increased ventilation. The use of charcoal filters for removal of radon gas in the indoor air by adsorption has also been proposed. Concentrations of radon progeny, which are responsible for most of the health risks associated with radon exposures, can be controlled by use of electrostatic or mechanical filtration. Air circulation can also reduce radon progeny concentrations in certain cases. This paper reviews the application and limitations of each of these control measures and discusses recent experimental results

  10. The effects of HVAC system design and operation on radon entry into school buildings

    International Nuclear Information System (INIS)

    Turner, W.A.; Leovic, K.W.; Craig, A.B.

    1990-01-01

    Heating, ventilating, and air conditioning (HVAC) systems in schools vary considerably and tend to have a greater impact on pressure differentials--and consequently radon levels--than do heating and air-conditioning systems in houses. If the HVAC system induces a negative pressure relative to the subslab area, radon can be pulled into the building. If the HVAC system pressurizes the building, it can prevent radon entry as long as the fan is running. However, school HVAC systems are normally set back or turned off on evenings and weekends and, even if the HVAC system pressurizes the school during operation, indoor radon levels may build up during setback periods. In this paper many of the historical methods utilized to deliver ventilation air (outdoor air) over the past 40 years are summarized. In addition, for each type of system presented, the possible impact the ventilation system might be expected to have (positive or negative) on the pressure of the building envelope (and subsequent radon levels in the building) is discussed

  11. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    International Nuclear Information System (INIS)

    Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

    2015-01-01

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  12. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of); Chang, Soon Heung [Handong Global University, 558, Handong-ro, Buk-gu, Pohang Gyeongbuk 37554 (Korea, Republic of); Choi, Yu Jung [Korea Hydro and Nuclear Power Co.—Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 34101 (Korea, Republic of); Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of)

    2015-12-15

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  13. Analysis on engineering application of CNP1000 in-containment refueling water storage tank

    International Nuclear Information System (INIS)

    Wang Bin; Wang Yong; Qiu Jian; Weng Minghui

    2005-01-01

    Based on the basic design of CNP1000 (three loops), which is self-reliance designed by China National Nuclear Cooperation, and investigation results from abroad advanced nuclear power plant design of In-containment Refueling Water Storage tank, this paper describe the system flowsheet, functional requirements, structural design and piping arrangement about In-containment Refueling Water Storage Tank. The design takes the lower structural space as the IRWST. Four areas are configured to meet the diverse functional requirements, including depressurization area, water collection area, safety injection and/or containment spray suction area, TSP storage area / reactor cavity flooding holdup tank. Also the paper depict the corresponding analysis and demonstration, such as In-containment Refueling Water Storage Tank pressure transient on depressurization area of IRWST, suction and internal flow stream of IRWST, configuration of strains, the addition method and amount of chemical addition, design and engineering applicant of Reactor Cavity Flooding System. All the analysis results show the basic design of IRWST meeting with the Utility Requirement Document's requirements on performance of safety function, setting of overfill passage, overpressure protection, related interference, etc., and show the reliability of Engineering Safety Features being improved for CNP1000 (three loops). Meanwhile, it is demonstrated that the design of In-containment Refueling Water Storage Tank can apply on the future nuclear power plant project in China. (authors)

  14. First use of ECOSIM in air management system

    International Nuclear Information System (INIS)

    Perez Vara, R.; Torroglosa Ponce, V.; Lebru, A.; Novara, M.

    1993-01-01

    ECOSIM is a software tool for the simulation of Environmental Control and Life Support (ECLS) systems which has been developed for the European Space Agency. A preliminary model of the Hermes Air Management System has been developed during the ECOSIM testing in order to assess the functionality of the software and to verify its results with those obtained from previous simulation tools. The model represents the Hermes cabin with its crew and it includes submodels for the sub-systems performing the following functions: - Temperature and Humidity Control; - Total Pressure and Composition Control; - Air revitalisation. The interactions between these different subsystem are taken into account by the model, while many of the previous simulations made assumptions to decouple the different subsystems (e.g: a constant cabin temperature has been assumed during cabin depressurization transients, to decouple the pressure control section from the air conditioning section). The model allows to check the validity of these kind of assumption. The paper also presents the results obtained with this model for the following experiments: - Normal 24 hour cycle; - Re-entry; - Depressurization due to a micro-meteorite hole; - Failure of the condensate water separators. The first version of ECOSIM will be released to the space industry before the end of year 1992 and the user's feedback will lead to improvements being implemented in later versions. (author)

  15. Rapid Evaporation of microbubbles

    Science.gov (United States)

    Gautam, Jitendra; Esmaeeli, Asghar

    2008-11-01

    When a liquid is heated to a temperature far above its boiling point, it evaporates abruptly. Boiling of liquid at high temperatures can be explosive and destructive, and poses a potential hazard for a host of industrial processes. Explosive boiling may occur if a cold and volatile liquid is brought into contact with a hot and non-volatile liquid, or if a liquid is superheated or depressurized rapidly. Such possibilities are realized, for example, in the depressurization of low boiling point liquefied natural gas (LNG) in the pipelines or storage tanks as a result of a leak. While boiling of highly heated liquids can be destructive at macroscale, the (nearly) instantaneous pace of the process and the release of large amount of kinetic energy make the phenomena extremely attractive at microscale where it is possible to utilize the released energy to derive micromechanical systems. For instance, there is currently a growing interest in micro-explosion of liquid for generation of micro bubbles for actuation purposes. The aim of the current study is to gain a fundamental understanding of the subject using direct numerical simulations. In particular, we seek to investigate the boundary between stable and unstable nucleus growth in terms of the degree of liquid superheat and to compare the dynamics of unstable and stable growth.

  16. The structure of hydrate bearing fine grained marine sediments

    Energy Technology Data Exchange (ETDEWEB)

    Priest, J.; Kingston, E.; Clayton, C. [Southampton Univ., Highfield (United Kingdom). School of Civil Engineering and the Environment; Schultheiss, P.; Druce, M. [Geotek Ltd., Daventry (United Kingdom)

    2008-07-01

    This paper discussed the structure of naturally occurring methane gas hydrates in fine-grained sediments from core samples recovered using in situ pressures from the eastern margin of the Indian Ocean. High resolution X-ray computed tomography (CT) images were taken of gas hydrate cores. The hydrate structure was examined and comparisons were made between low resolution X-ray images obtained on the cores prior to sub-sectioning and depressurization procedures. The X-ray images showed the presence of high-angle, sub-parallel veins within the recovered sediments. The scans indicated that the hydrates occurred as fracture filing veins throughout the core. Fracture orientation was predominantly sub-vertical. Thick millimetric hydrate veins were composed of sub-millimetric veins with variations in fracture angle. The analysis indicated that hydrate formation was episodic in nature and subject to changes in the stress regime. Results of the study showed that depressurization and subsequent freezing alter the structure of the sediment even when the gas hydrate has not been altered. A large proportion of the hydrate survived when outside of its stability region. The self-preserving behaviour of the hydrate was attributed to the endothermic nature of gas hydrate dissociation. It was concluded that the accurate physical characterization of gas hydrates can only be conducted when the core section remains under in situ stress conditions. 13 refs., 9 figs.

  17. Passive safety system of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

    2015-08-15

    Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

  18. Impact of recovery actions on IPE back-end results

    International Nuclear Information System (INIS)

    Guey, Ching; Kabadi, Jay

    1993-01-01

    This paper presents a sensitivity study of the impact of various recovery/accident management actions on the IPE results. The human actions studied include: (1) RWST replenishment; (2) depressurization; (3) AC power recovery; (4) Component Cooling Water (CCW) cross-connection; and (5) manual actuation of containment spray (cavity flooding). The results indicate that the total frequency of severe accident sequences involving containment failure is lower than 1.0 x E-6/yr when all these recovery/accident management actions are taken credit. It is concluded that with appropriate considerations of operator actions in Emergency Operating Procedures (EOPs) and future accident management policy, the uncertain phenomenological issues may not play as important a role as they do now. Although the actual nonrecovery probability used in this study may be refined during future implementation of the accident management policy, the results of this study indicate that the frequency of severe accidents in which the reactor vessel fails is low when the recovery/accident management actions are taken credit. It is also clear from the study that with major recovery actions (such as RWST replenishment, AC power recovery and depressurization) being taken credit, additional accident management type of actions (e.g. manual actuation of containment spray and CCW cross-connection) may not be as effective in reducing the risk

  19. Preliminary investigation of turbulent reactive mixing in PCRV/CV gas mixtures

    International Nuclear Information System (INIS)

    Boccio, J.L.

    1978-08-01

    Relaxation of the prima facie assumption of complete mixing of primary containment and secondary containment gases during postulated depressurization accidents within gas cooled reactors has led to a study program designed to identify and selectively quantify the relevant gas dynamic processes which are manifest during the depressurization event. Uncertainty in the degree of gas mixedness naturally leads to uncertainty in containment vessel design pressure and heat loads and possible combustion hazards therein. This report details an analytical approach in the modeling of the exhaust-jet structure during a penetration failure. A chemical kinetics model is also described for the possibility of examining diffusive flame structure assuming the exhaust jet is composed of combustibles as well. The salient features of the mixing model and associated reaction kinetics are embodied in the classical problem of a turbulent, chemically reacting jet exhausting into a stationary ambient atmosphere capable of supporting combustion. A so-called ''two equation'' turbulence model is linked to a chemical kinetics code describing the production of CO 2 and H 2 O with He and N 2 considered as inert diluents. The usefulness of the model is exemplified by experimental/numerical comparisons presented in the open literature and within this report. The need for such a calculational tool in HTGR safety research is stressed as well

  20. Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

  1. A study on the influence diagrams for the application to containment performance analysis

    International Nuclear Information System (INIS)

    Park, Joon Won

    1995-02-01

    Influence diagrams have been applied to containment performance analysis of Young-Gwang 3 and 4 in an effort to explicitly display the dependencies between events and to treat operator intervention more generally. This study has been initiated to remove the three major drawbacks of the current event tree methodology: 1) Event tree cannot express dependency between events explicitly. 2) Accident Progression Event Tree (APET) cannot represent entire containment system. 3) It is difficult to consider operator intervention with event tree. To resolve these problems, a new approach, i.e., influence diagrams, are proposed. In the present work, the applicability of the influence diagrams have been demonstrated to YGN 3 and 4 containment performance analysis and an assessment of accident management strategies. To show that the results of the application of influence diagrams are reasonable, results are compared with that of YGN 3 and 4 IPE. Both results are in good agreement. In addition, influence diagrams are used to assess two accident management strategies: 1) RCS depressurization, 2) cavity flooding. Cavity flooding has a favorable effect to late containment failure and basemat melt-through, and depressurization of RCS is good for steam generator tube rupture. However, early containment failure probability is worse in both cases. As a result of the present study, it is shown that influence diagrams can be applied to the containment performance analysis

  2. Current Status of Aerosol Generation and Measurement Facilities for the Verification Test of Containment Filtered Venting System in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, the design of aerosol generation and measurement systems are explained and present circumstances are also described. In addition, the aerosol test plan is shown. Containment Filtered Venting System (FCVS) is one of the safety features to reduce the amount of released fission product into the environment by depressurizing the containment. Since Chernobyl accident, the regulatory agency in several countries in Europe such as France, Germany, Sweden, etc. have been demanded the installation of the CFVS. Moreover, the feasibility study on the CFVS was also performed in U.S. After the Fukushima accident, there is a need to improve a containment venting or installation of depressurizing facility in Korea. As a part of a Ministry of Trade, Industry and Energy (MOTIE) project, KAERI has been conducted the integrated performance verification test of CFVS. As a part of the test, aerosol generation system and measurement systems were designed to simulate the fission products behavior. To perform the integrated verification test of CFVS, aerosol generation and measurement system was designed and manufactured. The component operating condition is determined to consider the severe accident condition. The test will be performed in normal conditions at first, and will be conducted under severe condition, high pressure and high temperature. Undesirable difficulties which disturb the elaborate test are expected, such as thermophoresis on the pipe, vapor condensation on aerosol, etc.

  3. Optimal valve location in long oil pipelines

    OpenAIRE

    Grigoriev, A.; Grigorieva, N.V.

    2007-01-01

    We address the valve location problem, one of the basic problems in design of long oil pipelines. Whenever a pipeline is depressurized, the shutoff valves block the oil flow and seal the damaged part of the pipeline. Thus, the quantity of oil possibly contaminating the area around the pipeline is determined by the volume of the damaged section of the pipeline between two consecutive valves. Then, ecologic damage can be quantified by the amount of leaked oil and the environmental characteristi...

  4. Fuel bundle impact velocities due to reverse flow

    International Nuclear Information System (INIS)

    Wahba, N.N.; Locke, K.E.

    1996-01-01

    If a break should occur in the inlet feeder or inlet header of a CANDU reactor, the rapid depressurization will cause the channel flow(s) to reverse. Depending on the gap between the upstream bundle and shield plug, the string of bundles will accelerate in the reverse direction and impact with the upstream shield plug. The reverse flow impact velocities have been calculated for various operating states for the Bruce NGS A reactors. The sensitivity to several analysis assumptions has been determined. (author)

  5. PKL

    International Nuclear Information System (INIS)

    Probst, P.; Szabo, I.

    1985-01-01

    This study concerns the simulation of the behavior of a PWR during a LOCA. One studies more particularly the refilling and reflooding phase (PKLI) and the end of the depressurization phase (PKLII). The aim of this test program is to provide the information needed for the validation/verification of computer codes (RELAP4 MOD6, TRAC PF1) and to test an instrumentation realized for the 2D/3D program. A comparison calculation/experiment is presented. 5 refs [fr

  6. Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-02

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Hirata, Kazuo; Gotou, Hiroki

    1990-03-01

    Experimental data for the 10% hot leg break test, Run SB-HL-02, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on June 30, 1987, are presented. This test assumed total failure of both high pressure injection (HPI) and auxiliary feedwater (AFW) systems. The test results were characterized by asymmetric loop responses, flashing in the cold legs and upper downcomer, as well as condensation depressurization in the cold legs following injection of emergency core coolant (ECC) from accumulators. (author)

  7. Final work plan : supplemental upward vapor intrusion investigation at the former CCC/USDA grain storage facility in Hanover, Kansas.

    Energy Technology Data Exchange (ETDEWEB)

    LaFreniere, L. M.; Environmental Science Division

    2008-12-15

    chloroform into four homes located on or adjacent to the former CCC/USDA facility. The technical objective will be accomplished by collecting sub-slab vapor samples. The preliminary data collected during the July 2007 investigation did not fully address the source of or migration pathway for the carbon tetrachloride detected in the four homes. The scope of work proposed here will generate additional data needed to help evaluate whether the source of the detected carbon tetrachloride is vapor intrusion attributable to activities of the CCC/USDA. The additional vapor sampling at Hanover will be performed, on behalf of the CCC/USDA, by the Environmental Science Division of Argonne National Laboratory and H&P Mobile Geochemistry of San Diego (http://www.handpmg.com). Argonne is a nonprofit, multidisciplinary research center operated by UChicago Argonne, LLC, for the U.S. Department of Energy (DOE). The CCC/USDA has entered into an interagency agreement with DOE, under which Argonne provides technical assistance to the CCC/USDA with environmental site characterization and remediation at its former grain storage facilities. The professional staff members of H&P Mobile Geochemistry are nationally leading experts in soil gas sampling and vapor intrusion investigations.

  8. Neogene volcanism and extension in Western Anatolian-Aegean area: A new geodynamic model

    Energy Technology Data Exchange (ETDEWEB)

    Agostini, S; Tonarini, S [Istituto di Geoscienze e Georisorse, Consiglio Nazionale delle Ricerche, Via G. Moruzzi 1, 56124 Pisa (Italy); Doglioni, C [Dipartimento di Scienze della Terra, Universita La Sapienza, Roma (Italy); Innocenti, F [Dipartimento di Scienze della Terra, Universita di Pisa, Pisa (Italy); Manetti, P [Dipartimento di Scienze della Terra, Universita di Firenze, Firenze (Italy)], E-mail: s.agostini@igg.cnr.it

    2008-07-01

    The widespread Western Anatolian-Aegean Neogene volcanism presents a complex geochemical evolution reflecting the uncommon space-time variability of the geodynamic setting of the region. In the Western Anatolian and Central Aegean, a widespread supra-subduction magmatism, with calc-alkaline to shoshonitic affinity, took place from Early to Middle Miocene; this phase of activity ends with spots of ultra-K lavas and dykes. From Late Miocene onwards scattered alkali basaltic lavas with intraplate affinity were emitted, while calc-alkaline activity occurred in the South Aegean arc. Since Late Oligocene-Early Miocene, the region was, and still is, affected by extensional tectonics generally ascribed to a backarc rift. However the Aegean region should rather be considered as an unconventional backarc since its characteristics rather differ from 'typical' backarcs. In fact, in spite of a long lasting(>40Ma) active NE-directed subduction of Africa, the backarc area still maintains a relatively thick continental crust (>20-25 km). Moreover, the upper Eurasian plate is overriding the lower Africa plate with separate segments, with Greece moving faster, and Turkey moving slower. The differential velocity between Greece and Turkey determines extension in the upper plate, unrelated to the loss of subducted retreating lithosphere, which is the usual setting for the origin of 'classic' backarc settings. The geodynamic framework is supported by the geochemical and isotopic features of the supra-subduction magmas revealing the occurrence of a trapped, drying slab, with progressive decreasing of Fluid Mobile Elements/Fluid Immobile Elements ratios, {delta}{sup 11}B and {delta}{sup 7}Li, coupled with scarce variations of Sr and Nd isotopes. Moreover, the differential motion between the Greek and Anatolian micro-plates creates tear zones with the formation of slab ruptures or vertical slab windows. The occurrence of such windows is, in fact, outlined by the

  9. Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

  10. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  11. Application of the TRAC-PD2 code to the simulation of the CANON experiment

    International Nuclear Information System (INIS)

    Neves Conti, T. das; Freitas, R.L.

    1985-01-01

    A comparison between the TRAC -PD2 code calculations and results from the CANON experiment is presented. The CANON experiment simulates the loss of coolant accident through the depressurization of a horizontal tube containing water at different temperatures. The experiment consist of the instantaneous rupture at one end of the tubing and the corresponding pressure and void fraction measurements during the transient. The comparison shows that the TRAC-PD2 code predicts satisfactorily the pressure and void fraction evolution in the CANON experiment. (F.C.) [pt

  12. 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.

  13. Introduction of the 2007-2008 JOGMEC/NRCan/Aurora Mallik Gas Hydrate Production Research Program, NWT, Canada

    Science.gov (United States)

    Yamamoto, K.; Dallimore, S. R.; Numasawa, M.; Yasuda, M.; Fujii, T.; Fujii, K.; Wright, J.; Nixon, F.

    2007-12-01

    Japan Oil, Gas and Metals National Corporation (JOGMEC) and Natural Resource Canada (NRCan) have embarked on a new research program to study the production potential of gas hydrates. The program is being carried out at the Mallik gas hydrate field in the Mackenzie Delta, a location where two previous scientific investigations have been carried in 1998 and 2002. In the 2002 program that was undertaken by seven partners from five countries, 468m3 of gas flow was measured during 124 hours of thermal stimulation using hot warm fluid. Small-scale pressure drawdown tests were also carried out using Schlumberger's Modular Dynamics Tester (MDT) wireline tool, gas flow was observed and the inferred formation permeabilities suggested the possible effectiveness of the simple depressurization method. While the testing undertaken in 2002 can be cited as the first well constrained gas production from a gas hydrate deposit, the results fell short of that required to fully calibrate reservoir simulation models or indeed establish the technical viability of long term production from gas hydrates. The objectives of the current JOGMEC/NRCan/Aurora Mallik production research program are to undertake longer term production testing to further constrain the scientific unknowns and to demonstrate the technical feasibility of sustained gas hydrate production using the depressurization method. A key priority is to accurately measure water and gas production using state-of-art production technologies. The primary production test well was established during the 2007 field season with the re-entry and deepening of JAPEX/JNOC/GSC Mallik 2L-38 well, originally drilled in 1998. Production testing was carried out in April of 2007 under a relatively low drawdown pressure condition. Flow of methane gas was measured from a 12m perforated interval of gas-hydrate-saturated sands from 1093 to 1105m. The results establish the potential of the depressurization method and provide a basis for future

  14. Thermal hydraulic research on next generation PWRs using ROSA/LSTF

    International Nuclear Information System (INIS)

    Yonomoto, T.; Anoda, Y.

    2000-01-01

    A thermal-hydraulic research on next generation PWRs has been conducted at JAERI using the ROSA-V/Large Scale Test Facility (LSTF), focusing on phenomena related to passive safety systems. This paper describes two test results conducted for this research: a small break loss-of-coolant accident (SBLOCA) test and a low pressure steady-state natural circulation (NC) test. The former test investigated a combined use of a SG secondary-side automatic depressurization system (SADS) and a gravity-driven injection system (GDIS) to mitigate a SBLOCA. The results have shown that the primary loop can be depressurized to the GDIS actuation pressure of 0.2 MPa by the SADS alone, and then the stable long-term core cooling can be established by NC. Results of both tests showed a complicated nonuniform flow behavior among SG U-tubes during NC, which was characterized by the coexistence of concurrent condensing two-phase flow in some tubes and stagnant two-phase stratification in the others. The mechanism for the stratification was understood from the measured secondary side temperature distribution showing the lowest temperature at the top and bottom regions and the highest around the midplane. This was caused by the saturation temperature difference corresponding to the static pressure difference, and the recirculation in the secondary. This secondary side temperature distribution enabled the condensation occurring around the tube top to be balanced with evaporation occurring around the midplane in the U-tube with the stratification. Since the heat transfer occurs primarily through tubes with the concurrent flow, the nonuniform behavior directly affects the effective heat transfer area at SG. When the SG primary side was modeled with one lumped flow channel, the RELAP5 significantly over predicted the primary depressurization rate, and could not predict the stable long-term core cooling behavior at low pressure. In order to understand the mechanism of the nonuniform behavior, the

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

    International Nuclear Information System (INIS)

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

    1976-09-01

    Recorded test data are presented for Test S-04-1 of the Semiscale Mod-1 Baseline ECC Test Series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-1 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using system volume scaled coolant injection parameters. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

  16. Investigation of the Methane Hydrate Formation by Cavitation Jet

    Science.gov (United States)

    Morita, H.; Nagao, J.

    2015-12-01

    Methane hydrate (hereafter called "MH") is crystalline solid compound consisting of hydrogen-bonded water molecules forming cages and methane gas molecules enclosed in the cage. When using MH as an energy resource, MH is dissociated to methane gas and water and collect only the methane gas. The optimum MH production method was the "depressurization method". Here, the production of MH means dissociating MH in the geologic layers and collecting the resultant methane gas by production systems. In the production of MH by depressurization method, MH regeneration was consider to important problem for the flow assurance of MH production system. Therefore, it is necessary to clarify the effect of flow phenomena in the pipeline on hydrate regeneration. Cavitation is one of the flow phenomena which was considered a cause of MH regeneration. Large quantity of microbubbles are produced by cavitation in a moment, therefore, it is considered to promote MH formation. In order to verify the possible of MH regeneration by cavitation, it is necessary to detailed understanding the condition of MH formation by cavitation. As a part of a Japanese National hydrate research program (MH21, funded by METI), we performed a study on MH formation using by cavitation. The primary objective of this study is to demonstrate the formation MH by using cavitation in the various temperature and pressure condition, and to clarify the condition of MH formation by using observation results.

  17. Implementation of a new interfacial mass and energy transfer model in RETRAN-3D

    International Nuclear Information System (INIS)

    Macian, R.; Cebulh, P.; Coddington, P.; Paulsen, M.

    1999-01-01

    The RETRAN-3D MOD002.0 best estimate code includes a five-equation flow field model developed to deal with situations in which thermodynamic non-equilibrium phenomena are important. Several applications of this model to depressurization and pressurization transients showed serious convergence problems. An analysis of the causes for the numerical instabilities identified the models for interfacial heat and mass transfer as the source of the problems. A new interfacial mass and energy transfer model has thus been developed and implemented in RETRAN-3D. The heat transfer for each phase is equal to the product of the interfacial area density, a heat transfer coefficient and the temperature difference between the interface at saturation and the bulk temperature of the respective phase. However, in the context of RETRAN-3D, the vapor remains saturated in a two-phase volume, and no vapor heat transfer is thus calculated. The values of interfacial area density and heat transfer coefficient are obtained based on correlations appropriate for different flow regimes. A flow regime map, based on the work of Taitel and Dukler, with void fraction and mixture mass flux as map coordinates, is used to identify the flow regime present in a given volume. The new model has performed well when assessed against data from four experimental facilities covering depressurization, condensation and steady state void distribution. The results also demonstrate the viability of the approach followed to develop the new model for a five-equation based code. (author)

  18. Self-leveling onset criteria in debris beds

    International Nuclear Information System (INIS)

    Zhang, Bin; Harada, Tetsushi; Hirahara, Daisuke; Matsumoto, Tatsuya; Morita, Koji; Fukuda, Kenji; Yamano, Hidemasa; Suzuki, Tohru; Tobita, Yoshiharu

    2010-01-01

    In a core-disruptive accident of a sodium-cooled fast breeder reactor, core debris may settle on the core-support structure and/or in the lower inlet plenum of the reactor vessel because of rapid quenching and fragmentation of molten core materials in the subcooled sodium plenum. Coolant boiling is the mechanism driving the self-leveling of a debris bed that causes significant changes in the heat-removal capability of the beds. In the present study, we develop criteria establishing the onset of this self-leveling behavior that we base on a force balance model assuming a debris bed with a single-sized spherical particle. The model considers drag, buoyancy, and gravity acting on each particle. A series of experiments with simulant materials verified the applicability of this description of self-leveling. Particle size (between 0.5-6 mm), shape (spherical and nonspherical), density (namely of alumina, zirconia, lead, and stainless steel), along with boiling intensity, bed volume, and even experimental methods were taken into consideration to obtain general characteristics of the self-leveling process. We decided to use depressurization boiling to simulate an axially increasing void distribution in the debris bed, although bottom heating was also used to validate the use of the depressurization method. On the self-leveling onset issues, we obtained good agreement between model predictions and experimental results. Extrapolation of our model to actual reactor conditions is discussed. (author)

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

    International Nuclear Information System (INIS)

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

    1976-07-01

    Recorded test data are presented for Test S-29-1 of the Semiscale Mod-1 special heat transfer test series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident (LOCA) in a pressurized-water reactor system. Test S-29-1 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,260 psia. An asymmetrical offset shear cold leg break was used to investigate the system response to a depressurization transient with a flow distribution different from that associated with a symmetrical cold leg break. System flow was set to achieve a core fluid temperature differential of 66 0 F at full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling (DNB) might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

  20. Irradiated ignition over solid materials in reduce pressure environment: Fire safety issue in man-made enclosure system

    Science.gov (United States)

    Nakamura, N.; Aoki, A.

    Effects of ambient pressure and oxygen yield on irradiated ignition characteristics over solid combustibles have been studied experimentally Aim of the present study is to elucidate the flammability and chance of fire in depressurized enclosure system and give ideas for the fire safety and fire fighting strategies in such environment Thin cellulosic paper is considered as the solid combustible since cellulose is one of major organic compounds and flammables in the nature Applied atmosphere consists of inert gas either CO2 or N2 and oxygen and various mixture ratios are of concerned Total ambient pressure level is varied from 0 1MPa standard atmospheric pressure to 0 02MPa Ignition is initiated by external thermal flux exposed into the solid surface as a model of unexpected thermal input to initiate the localized fire Thermal degradation of the solid induces combustible gaseous products e g CO H2 or other low class of HCs and the gas mixes with ambient oxygen to form the combustible mixture over the solid Heat transfer from the hot irradiated surface into the mixture accelerates the local exothermic reaction in the gas phase and finally thermal runaway ignition is achieved Ignition event is recorded by high-speed digital video camera to analyze the ignition characteristics Flammable map in partial pressure of oxygen Pox and total ambient pressure Pt plane is made to reveal the fire hazard in depressurized environment Results show that wider flammable range is obtained depending on the imposed ambient

  1. Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

    International Nuclear Information System (INIS)

    Henninger, R.J.; Boyack, B.E.

    1986-01-01

    A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

  2. Drive transmission system between a driving organ and a receiver organ

    International Nuclear Information System (INIS)

    Guillot, J.F.

    1985-01-01

    The present invention applies to the control rods of a water cooled nuclear reactor. The drive transmission system is disposed on the internal kinematic chain, between the control rod which is the receiver organ, and the driving organ. The control rod translation is obtained from a motion of rotation transformed in a motion of translation by means of a screw-nut system. The present invention prevents from control rod ejection in case of depressurization of the vessel containing the control rod drives or in case of reactor upsetting when it is embarked [fr

  3. Anxiety and anxiety disorders. Toward a conceptual reorientation.

    Science.gov (United States)

    Curtis, G C

    1985-03-01

    Traditionally, it has been assumed that there is only one type of anxiety; recent pharmacologic evidence suggests that there may be several. The psychoanalytic concept of "neurotic" symptoms as depressurizing mechanisms is out of keeping with most evidence now available. Spontaneous or "free-floating" anxiety may be partly biologic and genetic in origin. Anxiety symptoms evoked by specific stimuli behave in part like conditioned responses. Where conditioning theory has failed to propose a plausible unconditioned stimulus for pathologic anxiety, biology, ethology, and psychoanalysis may have been more successful.

  4. Introduction of the bubble rise dynamic model into the ALMOD 3 code pressurizer

    International Nuclear Information System (INIS)

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

    1985-01-01

    A new evaporation model for the ALMOD 3 code pressurizer is implemented in order to estimate more accurately the water level behaviour and its influence in the pressure transient for very fast depressurization cases. For the inclusion of the bubble rise dynamic model it was necessary to consider a two-phase mixture in the water volume. The modifications don't require additional input data and virtually had not modified the processing time. The results and processing time for the original and the new models are presented. (F.E.) [pt

  5. Method for Evaluating the Corrosion Resistance of Aluminum Metallization of Integrated Circuits under Multifactorial Influence

    Science.gov (United States)

    Kolomiets, V. I.

    2018-03-01

    The influence of complex influence of climatic factors (temperature, humidity) and electric mode (supply voltage) on the corrosion resistance of metallization of integrated circuits has been considered. The regression dependence of the average time of trouble-free operation t on the mentioned factors has been established in the form of a modified Arrhenius equation that is adequate in a wide range of factor values and is suitable for selecting accelerated test modes. A technique for evaluating the corrosion resistance of aluminum metallization of depressurized CMOS integrated circuits has been proposed.

  6. Rehabilitation of a house with high radon level, using a ground ventilation system with double barrier

    International Nuclear Information System (INIS)

    Bonnefous, Y.C.; Richon, P.; Arnautou, J.C.; Sabroux, J.C.

    1995-01-01

    A ground ventilation system has been designed and implemented in a town hall in Brittany. Radon concentration in the heating unit room of this building has been reduced from 10000 Bq/m 3 to less than 200 Bq/m 3 by the means of a depressurization system using a 32 W fan, which blows air into a permeable gravel layer intercalated between two radon barrier mylar films. Results show that passive systems should be applicable; for new buildings, very low energy consumption systems with 10 W fans, are easily implemented if designed before construction

  7. Assessment of a pressurizer spray valve faulty opening transient at Asco Nuclear Power Plant with RELAP5/MOD2

    International Nuclear Information System (INIS)

    Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P.

    1993-12-01

    The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ''Pressurizer spray valve faulty opening'' presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data

  8. Analysis of fission product release from HTGR core during transient temperature excursion

    International Nuclear Information System (INIS)

    Saito, Takao; Yamatoya, Naotoshi; Onuma, Mamoru

    1978-01-01

    The computer program ''FRANC'' was developed to calculate the release activity of fission products from a high-temperature gas cooled reactor (HTGR) core during transient temperature excursions such as a hypothetical loss of forced circulation combined with design basis depressurization. The program utilizes a segmented cylindrical core spatial model with the associated values of the prior fuel irradiation history and temperature conditions. The fission product transport and decay chain behavior is expressed by a set of differential equations. This set of equations describes the entire core inventory of fission products by means of calculated parameters based on the detailed spatial core conditions. The program determines the time-dependent amounts of fission product nuclides escaping from the core into the coolant. Coded in Continuous System Simulation Language (CSSL) with double precision, FRANC showed appropriate results for both short- and long-lived fission product nuclides. The sample calculation conducted by applying the program to a large HTGR indicated that it would take about one hour for noble gases and volatile nuclides to be released to the coolant, and several hours for metalic nuclides. (auth.)

  9. High capacity Venturi scrubber to separate aerosol-borne radioactivity from an air-gas-steam mixture. Final report

    International Nuclear Information System (INIS)

    Mayinger, F.; Glueckert, U.

    1993-01-01

    All German LWR are equipped with devices which in the case of a hypothetic accident permit a filtered depressurization of the containment precluding failure of the latter and minimizing the release of radioactive materials into the environment. To filter the aerosol charged air-steam mixture from the containment also a venturi scrubber is used. It has the great advantage that it can remove safely and over a certain period of time, even without active cooling systems, the after-heat released from the separated radioactive materials. Those separated radioactive materials are trapped in a scrubbing liquid which, in the event of a temporary failure of all active cooling systems, may partly evaporate and thus remove the heat in a completely passive way. The venturi scrubbers conceived earlier by the reactor manufacturer are of a very simple design and not optimized to achieve highest separation degrees. Therefore development work was started to optimize the separation behaviour of the venturi scrubber precisely with regard to submicron aerosols which are to be expected after a core meltdown accident. To achieve this, a special concept of scrubbing liquid addition developed by the contractor, the so-called multistage concept, was applied adapting it to the specific requirements. (orig./HP) [de

  10. A review of radon mitigation in large buildings in the US

    International Nuclear Information System (INIS)

    Craig, A.B.

    1994-01-01

    The Environmental Protection Agency of the US carried out its initial research on radon mitigation in houses, both existing and new. A review of this work is presented in another paper at this workshop. Four years ago, this work was expanded to include the study of radon in schools, both new and existing, and now includes studies in other large buildings, as well. Factors affecting ease of mitigation of existing schools using active soil depressurisation (ASD) have been identified and quantified. Examination of the building and architectural plans makes it possible to predict the ease of mitigation of a specific building. Many schools can be easily and inexpensively mitigated using ASD. However, examination of a fairly large number of schools has shown that a significant percentage of existing schools will be hard to mitigate with ASD. In some cases, the heating, ventilating, and air conditioning (HVAC) system can be used to pressurise the building and retard radon entry. However, in some cases no central HVAC system exists and the school is difficult and/or expensive to mitigate by any technique. Prevention of radon entry is relatively easy and inexpensive to accomplish during construction of schools and other large buildings. It is also possible to control radon to near ambient levels in new construction, a goal which is much more difficult to approach in existing large buildings. The preferred method of radon prevention in the construction of large buildings is to design the HVAC system for building pressurisation, install a simple ASD system, and seal all entry routes between the sub-slab and the building interior. (author)

  11. New-construction techniques and HVAC overpressurization for radon reduction in schools

    International Nuclear Information System (INIS)

    Saum, D.; Witter, K.A.; Craig, A.B.

    1988-01-01

    Construction of a school in Fairfax County, Virginia, is being carefully monitored since elevated indoor radon levels have been identified in many existing houses near the site. Soil gas radon concentrations measured prior to pouring of the slabs were also indicative of a potential radon problem should the soil gas enter the school; however, subslab radon measurements collected thus far are lower than anticipated. Radon-resistant features have been incorporated into construction of the school and include the placing of at least 100 mm of clean coarse aggregate under the slab and a plastic film barrier between the aggregate and the slab, the sealing of all expansion joints, the sealing or plugging of all utility penetrations where possible, and the painting of interior block walls. In addition, the school's heating, ventilating, and air-conditioning (HVAC) system has been designed to operate continuously in overpressurization to help reduce pressure-driven entry of radon-containing soil gas into the building. Following completion, indoor radon levels in the school will be monitored to determine the effectiveness of these radon-resistant new-construction techniques and HVAC overpressurization in limiting radon entry into the school

  12. Reasons for increasing radon concentrations in radon remediated houses

    International Nuclear Information System (INIS)

    Clavensjoe, B.

    1997-01-01

    The study comprises 31 single-dwelling houses where remedial actions were carried out in the 1980s. In all of them the radon concentrations have increased more than 30% according to recent control measurements. Radon sources are building material as well as the soil. The remedial actions dealt with ventilation systems, leakage through the basement floor, air cushions, sub-slab suction or radon wells according to the original problems. Causes for the increase varied: In many houses with soil radon problems, the installation of a normal mechanical ventilation system is not a good remedial action. In some houses on a ground with high permeability and high radon content in the soil air, the radon concentration may increase by the lowering of the indoor air pressure. In other houses the increase was a measurement effect, where sites/rooms were confused. Living related causes were identified in a number of cases, where fan speeds were reduced for energy conservation/noise reduction purposes or different use of windows airing had occurred. Extension of the dwelling space without changing the ventilation system caused the increase in one house. 23 refs

  13. Control of radon and its progeny concentration in indoor atmosphere

    International Nuclear Information System (INIS)

    Ramachandran, T.V.; Subbaramu, M.C.

    1986-01-01

    Exposure to radon daughter concentration in indoor atmosphere can result in a significant risk to the general public. There are two generally used methods for the control of radon and progeny concentration in the indoor atmosphere, namely restriction of radon entry and reduction of indoor radon and its progeny concentration by ventilation or by air cleaning. Predominant radon entry process in most of the dwellings appears to be by pressure driven flow of soil gas through cracks or other openings in the basement slab or subfloors. Sealing these openings or ventilation of the subslab or subfloor space are the methods for reducing the radon entry rates. Indoor radon concentration can also be reduced by increasing the ventilation and by using charcoal filters for the removal of radon gas in indoor air by absorption. Concentration of radon progeny, which are responsible for most of the health risks associatd with radon exposure can also be controlled by the use of electrostatic or mechanical filters. This study describes briefly the above control strategies used for reducing the inhalation doses to persons in dwellings. (author). 9 refs., 2 tables

  14. Radiation and Heterogeneous processes and hydrogen safety of nuclear reactors

    International Nuclear Information System (INIS)

    Agayev, T.N.; Eyubov, K.T.; Aliyev, S.M.; Faradjzade, I.A.; Imanova, G.T.

    2017-01-01

    Due to the development of the quantitative and probabilistic analysis of safety of atomic power stations, interest in major accidents which can lead to overheating and fusion of an active zone has increased now. One of the major processes from the point of view of assessment of accident consequences with damage of an active zone is process of hydrogen formation. In the real work sources of hydrogen formation at various stages of accident with loss of the coolant of water-to-water power reactors are considered. The role of different processes of hydrogen formation depends on temperature, an amount of water and steam in an active zone and some other parameters. In this regard we have tried to formulate approach to creation of mathematical model of dynamics of hydrogen formation at accident in which the factors mentioned above would be considered. At the first stage of accident which lasted several tens of seconds depressurization of the first contour and loss of pressure took place. Water of the first contour under normal conditions of operation contained radiolytic hydrogen which concentration significantly exceeded its solubility with an atmospheric pressure. Therefore the dissolved hydrogen was emitted in a gas phase at a rupture of the pipeline. The second stage of accident is characterized by water vaporization from the first contour. During this period the amount of water in an active zone is constant and also water temperature in an active zone is constant. At last, at the third stage of accident there is water vaporization from an active zone also a warming up of the heat allocating assembly and constructional materials of an active zone.

  15. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Bradley, J.G.

    1982-01-01

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception

  16. Transition from depressurization to long term cooling in AP600 scaled integral test facilities

    International Nuclear Information System (INIS)

    Bessette, D.E.; Marzo, M. di

    1999-01-01

    A novel light water reactor design called the AP600 has been proposed by the Westinghouse Electric Corporation. In the evaluation of this plant's behavior during a small break loss of coolant accident (LOCA), the crucial transition to low pressure, long-term cooling is marked by the injection of the gravitationally driven flow from the in-containment refueling water storage tank (IRWST). The onset of this injection is characterized by intermittency in the IRWST flow. This happens at a time when the reactor vessel reaches its minimum inventory. Therefore, it is important to understand and scale the behavior of the integral experimental test facilities during this portion of the transient. The explanation is that the periodic liquid drains and refills of the pressurizer are the reason for the intermittent behavior. The momentum balance for the surge line yields the nondimensional parameter controlling this process. Data from one of the three experimental facilities represent the phenomena well at the prototypical scale. The impact of the intermittent IRWST injection on the safe plant operation is assessed and its implications are successfully resolved. The oscillation is found to result from, in effect, excess water in the primary system and it is not of safety significance. (orig.)

  17. FINAL PROJECT REPORT - EVALUATION AND TESTING OF HTGR REACTOR BUILDING RESPONSE TO DEPRESSURIZATION ACCIDENTS

    Energy Technology Data Exchange (ETDEWEB)

    ALLIANCE LIMITED, NGNP INDUSTRY

    2017-07-25

    This report provides a description of the project, summarizes each phase of the project, and ends with project conclusions. In addition, the report contains a descriptive index of the technical reports generated during the course of the project.

  18. A hybrid method of prediction of the void fraction during depressurization of diabatic systems

    International Nuclear Information System (INIS)

    Inayatullah, G.; Nicoll, W.B.; Hancox, W.T.

    1977-01-01

    The variation in vapour volumetric fraction during transient pressure, flow and power is of considerable importance in water-cooled nuclear power-reactor safety analysis. The commonly adopted procedure to predict the transient void is to solve the conservation equations using finite differences. This present method is intermediate between numerical and analytic, hence 'hybrid'. Space and time are divided into discrete intervals. Their size, however, is dictated by the imposed heat flux and pressure variations, and not by truncation error, stability or convergence, because within an interval, the solutions applied are analytic. The relatively simple hybrid method presented here can predict the void distribution in a variety of transient, diabatic, two-phase flows with simplicity, accuracy and speed. (Auth.)

  19. Release of Dissolved CO2 from Water in Laboratory Porous Media Following Rapid Depressurization

    Science.gov (United States)

    Crews, J. B.; Cooper, C. A.

    2011-12-01

    A bench-top laboratory study is undertaken to investigate the effects of seismic shocks on brine aquifers into which carbon dioxide has been injected for permanent storage. Long-term storage in deep saline aquifers has been proposed and studied as one of the most viable near-term options for sequestering fossil fuel-derived carbon dioxide from the atmosphere to curb anthropogenic climate change. Upon injection into the subsurface, it is expected that CO2, as either a gas or supercritical fluid, will mix convectively with the formation water. The possibility exists, however, that dissolved CO2 will come out of solution as a result of an earthquake. The effect is similar to that of slamming an unsealed container of carbonated beverage on a table; previously dissolved CO2 precipitates, forms bubbles, and rises due to buoyancy. In this study, we measure the change in gas-phase CO2 concentration as a function of the magnitude of the shock and the initial concentration of CO2. In addition, we investigate and seek to characterize the nucleation and transport of CO2 bubbles in a porous medium after a seismic shock. Experiments are conducted using a Hele-Shaw cell and a CCD camera to quantify the fraction of dissolved CO2 that comes out of solution as a result of a sharp mechanical impulse. The data are used to identify and constrain the conditions under which CO2 comes out of solution and, further, to understand the end-behavior of the precipitated gas-phase CO2 as it moves through or is immobilized in a porous medium.

  20. Scaling relation and regime map of explosive gas–liquid flow of binary Lennard-Jones particle system

    KAUST Repository

    Inaoka, Hajime

    2012-02-01

    We study explosive gasliquid flows caused by rapid depressurization using a molecular dynamics model of Lennard-Jones particle systems. A unique feature of our model is that it consists of two types of particles: liquid particles, which tend to form liquid droplets, and gas particles, which remain supercritical gaseous states under the depressurization realized by simulations. The system has a pipe-like structure similar to the model of a shock tube. We observed physical quantities and flow regimes in systems with various combinations of initial particle number densities and initial temperatures. It is observed that a physical quantity Q, such as pressure, at position z measured along a pipe-like system at time t follows a scaling relation Q(z,t)=Q(zt) with a scaling function Q(ζ). A similar scaling relation holds for time evolution of flow regimes in a system. These scaling relations lead to a regime map of explosive flows in parameter spaces of local physical quantities. The validity of the scaling relations of physical quantities means that physics of equilibrium systems, such as an equation of state, is applicable to explosive flows in our simulations, though the explosive flows involve highly nonequilibrium processes. In other words, if the breaking of the scaling relations is observed, it means that the explosive flows cannot be fully described by physics of equilibrium systems. We show the possibility of breaking of the scaling relations and discuss its implications in the last section. © 2011 Elsevier B.V. All rights reserved.

  1. Validation of CFD modeling for VGM loss-of-forced-cooling accidents

    International Nuclear Information System (INIS)

    Wysocki, Aaron; Ahmed, Bobby; Charmeau, Anne; Anghaie, Samim

    2009-01-01

    Heat transfer and fluid flow in the VGM reactor cavity cooling system (RCCS) was modeled using Computational Fluid Dynamics (CFD). The VGM is a Russian modular-type high temperature helium-cooled reactor. In the reactor cavity, heat is removed from the pressure vessel wall through natural convection and radiative heat transfer to water-cooled vertical pipes lining the outer cavity concrete. The RCCS heat removal capability under normal operation and accident scenarios needs to be assessed. The purpose of the present study is to validate the use of CFD to model heat transfer in the VGM RCCS. Calculations were based on a benchmark problem which defines a two-dimensional temperature distribution on the pressure vessel outer wall for both Depressurized and Pressurized Loss-of-Forced Cooling events. A two-dimensional axisymmetric model was developed to determine the best numerical modeling approach. A grid sensitivity study for the air region showed that a 20 mm mesh size with a boundary layer giving a maximum y+ of 2.0 was optimal. Sensitivity analyses determined that the discrete ordinates radiative model, the k-omega turbulence model, and the ideal gas law gave the best combination for capturing radiation and natural circulation in the air cavity. A maximum RCCS pipe wall temperature of 62degC located 6 m from the top of the cavity was predicted. The model showed good agreement with previous results for both Pressurized and Depressurized Loss-of-Forced-Cooling accidents based on RCCS coolant outlet temperature, relative contributions of radiative and convective heat transfer, and RCCS heat load profiles. (author)

  2. Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

    International Nuclear Information System (INIS)

    Rimpault, G.; Sugino, K.; Hayashi, H.

    2005-01-01

    In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

  3. AIR INGRESS ANALYSIS: COMPUTATIONAL FLUID DYNAMIC MODELS

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Richard Schultz; Hans Gougar; David Petti; Hyung S. Kang

    2010-08-01

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high temperature reactors (VHTRs). Phenomena Identification and Ranking Studies to date have ranked an air ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority. Following a loss of coolant and system depressurization incident, air will enter the core of the High Temperature Gas Cooled Reactor through the break, possibly causing oxidation of the in-the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of the lower plenum graphite can lead to a loss of structural support. Excessive oxidation of core graphite can also lead to the release of fission products into the confinement, which could be detrimental to a reactor safety. Computational fluid dynamic model developed in this study will improve our understanding of this phenomenon. This paper presents two-dimensional and three-dimensional CFD results for the quantitative assessment of the air ingress phenomena. A portion of results of the density-driven stratified flow in the inlet pipe will be compared with results of the experimental results.

  4. Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

    International Nuclear Information System (INIS)

    Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

    2009-01-01

    The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

  5. Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

    Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  6. Response of ventilation dampers to large airflow pulses

    International Nuclear Information System (INIS)

    Gregory, W.S.; Smith, P.R.

    1985-04-01

    The results of an experiment program to evaluate the response of ventilation system dampers to simulated tornado transients are reported. Relevant data, such as damper response time, flow rate and pressure drop, and flow/pressure vs blade angle, were obtained, and the response of one tornado protective damper to simulated tornado transients was evaluated. Empirical relationships that will allow the data to be integrated into flow dynamics codes were developed. These flow dynamics codes can be used by safety analysts to predict the response of nuclear facility ventilation systems to tornado depressurization. 3 refs., 21 figs., 6 tabs

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  8. Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

    Directory of Open Access Journals (Sweden)

    Roman Mukin

    2018-04-01

    Full Text Available A series of tests dedicated to station blackout (SBO accident scenarios have been recently performed at the Primärkreislauf-Versuchsanlage (primary coolant loop test facility; PKL facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4 of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for

  9. Effects of periodic atmospheric pressure variation on radon entry into buildings

    Science.gov (United States)

    Tsang, Y. W.; Narasimhan, T. N.

    1992-06-01

    Using a mathematical model, we have investigated the temporal variations of radon entry into a house basement in the presence of time-dependent periodic variations of barometric pressure as well as a persistent small steady depressurization within the basement. The tool for our investigation is an integral finite difference numerical code which can solve for both diffusive and advective flux of radon in the soil gas which is treated as a slightly compressible fluid. Two different boundary conditions at the house basement are considered: (1) a dirt floor basement so that diffusion is equally or more important than advective transport, and (2) an "impermeable" cement basement except for a 1-cm-wide crack near the perimeter of the basement floor; in which case, advective transport of radon flux dominates. Two frequencies of barometric pressure fluctuation with representative values of amplitudes, based on a Fourier decomposition of barometric pressure data, were chosen in this study: one with a short period of 0.5 hour with pressure amplitude of 50 Pa, the other a diurnal variation with a period of 24 hours with the typical pressure amplitude of 250 Pa. For a homogeneous soil medium with soil permeability to air between 10-13 and 10-10 m2, we predict that the barometric fluctuations increase the radon entry into the basement by up to 120% of the steady radon inflow into the basement owing to a steady depressurization of 5 Pa. If soil permeability heterogeneity is present, such as the presence of a thin layer of higher permeability aggregate immediately below the basement floor, radon flux due to atmospheric pumping is further increased. Effects of pressure pumping on radon entry are also compared to diffusion-only transport when the steady depressurization is absent. It is found that contribution to radon entry is significant for the basement crack configuration. In particular, for pressure pumping at 0.5-hour period and for a homogeneous medium of permeability of 10

  10. The rise and fall of periodic 'drumbeat' seismicity at Tungurahua volcano, Ecuador

    Science.gov (United States)

    Bell, Andrew F.; Hernandez, Stephen; Gaunt, H. Elizabeth; Mothes, Patricia; Ruiz, Mario; Sierra, Daniel; Aguaiza, Santiago

    2017-10-01

    Highly periodic 'drumbeat' long period (LP) earthquakes have been described from several andesitic and dacitic volcanoes, commonly accompanying incremental ascent and effusion of viscous magma. However, the processes controlling the occurrence and characteristics of drumbeat, and LP earthquakes more generally, remain contested. Here we use new quantitative tools to describe the emergence, evolution, and degradation of drumbeat LP seismicity at the andesitic Tungurahua volcano, Ecuador, in April 2015. The signals were recorded during an episode of minor explosive activity and ash emission, without lava effusion, and are the first to be reported at Tungurahua during the ongoing 17 yrs of eruption. Following four days of high levels of continuous and 'pulsed' tremor, highly-periodic LP earthquakes first appear on 10 April. Over the next four days, inter-event times and event amplitudes evolve through a series of step-wise transitions between stable behaviors, each involving a decrease in the degree of periodicity. Families of similar waveforms persist before, during, and after drumbeat activity, but the activity levels of different families change coincidentally with transitions in event rate, amplitude, and periodicity. A complex micro-seismicity 'initiation' sequence shows pulse-like and stepwise changes in inter-event times and amplitudes in the hours preceding the onset of drumbeat activity that indicate a partial de-coupling between event size and rate. The observations increase the phenomenology of drumbeat LP earthquakes, and suggest that at Tungurahua they result from gas flux and rapid depressurization controlled by shear failure of the margins of the ascending magma column.

  11. Implementation of condition-dependent probabilistic risk assessment using surveillance data on passive components

    International Nuclear Information System (INIS)

    Lewandowski, Radoslaw; Denning, Richard; Aldemir, Tunc; Zhang, Jinsuo

    2016-01-01

    Highlights: • Condition-dependent probabilistic risk assessment (PRA). • Time-dependent characterization of plant-specific risk. • Containment bypass involving in secondary system piping and SCC in SG tubes. - Abstract: A great deal of surveillance data are collected for a nuclear power plant that reflect the changing condition of the plant as it ages. Although surveillance data are used to determine failure probabilities of active components for the plant’s probabilistic risk assessment (PRA) and to indicate the need for maintenance activities, they are not used in a structured manner to characterize the evolving risk of the plant. The present study explores the feasibility of using a condition-dependent PRA framework that takes a first principles approach to modeling the progression of degradation mechanisms to characterize evolving risk, periodically adapting the model to account for surveillance results. A case study is described involving a potential containment bypass accident sequence due to the progression of flow-accelerated corrosion in secondary system piping and stress corrosion cracking of steam generator tubes. In this sequence, a steam line break accompanied by failure to close of a main steam isolation valve results in depressurization of the steam generator and induces the rupture of one or more faulted steam generator tubes. The case study indicates that a condition-dependent PRA framework might be capable of providing early identification of degradation mechanisms important to plant risk.

  12. Using a Reactive Transport Simulator to Simulate CH4 Production from Bear Island Basin in the Barents Sea Utilizing the Depressurization Method†

    Directory of Open Access Journals (Sweden)

    Khadijeh Qorbani

    2017-02-01

    Full Text Available The enormous amount of methane stored in natural gas hydrates (NGHsworldwide offers a significant potential source of energy. NGHs will be generally unable to reach thermodynamic equilibrium at their in situ reservoir conditions due to the number of active phases involved. Lack of reliable field data makes it difficult to predict the production potential and safety of CH4 production from NGHs. While the computer simulations will never be able to replace field data, one can apply state-of-the-artmodellingtechniquestoevaluateseveralpossiblelong-termscenarios. Realistic kinetic models for hydrate dissociation and reformation will be required, as well as analysis of all phase transition routes. This work utilizes our in-house extension of RetrasoCodeBright (RCB, a reactive transport simulator, to perform a gas hydrate production case study of the Bjørnøya (Bear Island basin, a promising field with very limited geological data reported by available field studies. The use of a reactive transport simulator allowed us to implement non-equilibrium thermodynamics for analysisofCH4 production from the gas hydrates by treating each phase transition involving hydrates as a pseudo reaction. Our results showed a rapid propagation of the pressure drop through the reservoir following the imposition of pressure drawdown at the well. Consequently, gas hydrate dissociation and CH4 production began in the early stages of the five-year simulation period.

  13. Best practices for the abandonment of pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Mackean, M; Reed, R; Snow, B [Nabors Canada, Calgary, AB (Canada). Abandonrite Service

    2006-07-01

    Pipeline regulations implemented in 2006 require that licensees register all pipelines. Training must also be provided for ground disturbance supervisors. In addition, signage must be maintained on abandoned pipelines, and discontinued pipelines must be properly isolated. Corrosion control and internal inhibition is required for discontinued lines. However, pipelines are often neglected during the well abandonment process. This presentation provided recommendations for coordinating well and pipeline abandonment processes. Pipeline ends can be located, depressurized, flushed and purged while wells are being abandoned. Contaminated soils around the wells can also be identified prior to reclamation activities. Administrative reviews must be conducted in order to provide accurate information on pipeline location, reclamation certification, and line break history. Field operation files must be reviewed before preliminary field work is conducted. Site inspections should be used to determine if all ends of the line are accessible. Landowners and occupants near the line must also be notified, and relevant documentation must be obtained. Skilled technicians must be used to assess the lines for obstructions as well as to cut and cap the lines after removing risers. The presentation also examined issues related to pressure change, movement, cold tapping, and live dead legs. tabs., figs.

  14. Safety design

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Shiozawa, Shusaku

    2004-01-01

    JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

  15. Computer simulation of the hydroelastic response of a pressurized water reactor to a sudden depressurization

    International Nuclear Information System (INIS)

    Dienes, J.K.; Hirt, C.W.; Stein, L.R.

    1977-03-01

    A computer program is being developed to analyze the response of the core support barrel to a sudden loss of coolant in a pressurized water reactor. This program, SOLA-FLX, combines SOLA-DF, a two-dimensional, two-phase, hydrodynamic code with FLX, a finite-difference code that integrates the Timoshenko equations of elastic shell motion. The programs are coupled so that the shell motion determined by FLX is used as a boundary condition by SOLA. In turn, the pressure determined by SOLA is the forcing term that controls the shell motion. An axisymmetric version was first developed to provide a basis for comparing with a simple set of experiments and to serve as a test case for the more general, unsymmetric version. The unsymmetric version is currently under development. The report describes the hydrodynamic code, the symmetric shell code, the unsymmetric shell code, and the method of coupling. Test problems used to verify the shell codes and coupled codes are also reported. Work is continuing to verify both the symmetric and unsymmetric codes by making comparisons with experimental data and with theoretical test problems

  16. The application of air pressure difference in reducing indoor radon concentration

    International Nuclear Information System (INIS)

    Leung, J.K.C.; Tso, M.Y.W.

    2000-01-01

    In densely populated tropical cities like Hong Kong, people usually live and work inside high-rise buildings. And because of the hot and humid climate, air conditioning systems are used throughout the year, particularly in commercial buildings. Previous territory-wide surveys have shown that over 10% of commercial buildings in Hong Kong have indoor radon concentrations above 200 Bq m -3 . Since the major source of indoor radon in high-rise buildings is the building materials, increasing ventilation and applying radon barriers on wall surfaces seem to be the only ways to reduce the indoor radon concentration. But it was noted that the ventilation rate the many commercial buildings are not efficient enough to remove the radon because of various reasons such as energy saving, lack of maintenance, etc. In this study, radon mitigation was achieved by reducing the rate of radon exhaled from the building materials. A special laboratory, which has the capability of simulating any meteorological conditions that could be faced by high-rise buildings in Hong Kong, was built. The reduction of radon exhalation rate by applying pressure difference and temperature difference across walls was studied in the laboratory. This paper summarizes the results and tactics for applying pressure difference in existing commercial buildings. A new technique of reducing radon exhalation rate in new buildings by depressurizing the interior of walls was also developed. Tunnels can be embedded in the concrete walls of new buildings during construction. By using simple vacuum pumps, radon exhalation rate from the walls can be reduced significantly by depressurizing the tunnels. The feasibility and applicability of the technique is presented in this paper. (author)

  17. Entry of soil gas and radon into houses

    International Nuclear Information System (INIS)

    Andersen, C.E.

    1992-04-01

    Entry of soil gas and radon into houses has been investigated by experiments conducted at radon test structures and numerical or analytical modelling. The numerical model solves the steady-state equations for Darcy flow of soil-gas and combined diffusive and advective transport of radon. Model calculations were compared with results from field experiments conducted at Risoe National Laboratory, and it was found that there was good agreement between measured and modelled pressure coupling and radon concentration profiles. Discrepancies regarding absolute values of soil-gas entry rates and radon concentrations were observed. The numerical model has been used to study the importance of soil and building related factors on radon entry rates into slab-on-grade houses. It was found that, for a house with a 3 mm perimeter crack along the floor-wall joint, the entry was mainly determined by the soil permeability and building related factors such as house depressurization and presence of a capillary breaking layer of gravel below the slab. In a house with a bare soil floor, the diffusivity of the soil was found to be of principal importance for the entry rate even for moderate permeabilities. An analytical model was developed for the purpose of studying soil-gas entry rates into houses in response to non-static driving forces. It is based on the analogy between a 'buried drain' and a basement house with a perimeter crack. The structure was depressurized sinusoidally in time and the frequency dependent pressure couplings were measured. There was fairly good agreement between theoretical and experimental results. (LN) (26 tabs., 30 ills., 66 refs.)

  18. Numerical studies of gas production from several CH4 hydrate zones at the Mallik site, Mackenzie Delta, Canada

    Science.gov (United States)

    Moridis, G.J.; Collett, T.S.; Dallimore, S.R.; Satoh, T.; Hancock, S.; Weatherill, B.

    2004-01-01

    The Mallik site represents an onshore permafrost-associated gas hydrate accumulation in the Mackenzie Delta, Northwest Territories, Canada. A gas hydrate research well was drilled at the site in 1998. The objective of this study is the analysis of various gas production scenarios from five methane hydrate-bearing zones at the Mallik site. In Zone #1, numerical simulations using the EOSHYDR2 model indicated that gas production from hydrates at the Mallik site was possible by depressurizing a thin free gas zone at the base of the hydrate stability field. Horizontal wells appeared to have a slight advantage over vertical wells, while multiwell systems involving a combination of depressurization and thermal stimulation offered superior performance, especially when a hot noncondensible gas was injected. Zone #2, which involved a gas hydrate layer with an underlying aquifer, could yield significant amounts of gas originating entirely from gas hydrates, the volumes of which increased with the production rate. However, large amounts of water were also produced. Zones #3, #4 and #5 were lithologically isolated gas hydrate-bearing deposits with no underlying zones of mobile gas or water. In these zones, thermal stimulation by circulating hot water in the well was used to induce dissociation. Sensitivity studies indicated that the methane release from the hydrate accumulations increased with the gas hydrate saturation, the initial formation temperature, the temperature of the circulating water in the well, and the formation thermal conductivity. Methane production appears to be less sensitive to the specific heat of the rock and of the hydrate, and to the permeability of the formation. ?? 2004 Published by Elsevier B.V.

  19. Toward resolving model-measurement discrepancies of radon entry into houses

    International Nuclear Information System (INIS)

    Garbesi, K.; Lawrence Berkeley Lab., CA

    1994-10-01

    Analysis of the literature indicated that radon transport models significantly and consistently underpredict the advective entry into houses of soil-gas borne radon. Advective entry is the dominant mechanism resulting in high concentrations of radon indoors. The author investigated the source of the model-measurement discrepancy via carefully controlled field experiments conducted at an experimental basement located in natural soil in Ben Lomond, California. Early experiments at the structure confirmed the existence and magnitude of the model-measurement discrepancy, ensuring that it was not merely an artifact of inherently complex and poorly understood field sites. The measured soil-gas entry rate during structure depressurization was found to be an order of magnitude larger than predicted by a current three-dimensional numerical model of radon transport. The exact magnitude of the discrepancy depends on whether the arithmetic or geometric mean of the small-scale measurements of permeability is used to estimate the effective permeability of the soil. This factor is a critical empirical input to the model and was determined for the Ben Lomond site in the typical fashion using single-probe static depressurization measurements at multiple locations. The remainder of the dissertation research tests a hypothesis to explain the observed discrepancy: that soil permeability assessed using relatively small-scale probe measurements does not reflect bulk soil permeability for flows that is likely to occur at larger scales of several meters or more in real houses and in the test structure. The idea is that soil heterogeneity is of a nature that, as flows occur over larger scales, larger scales of heterogeneity are encountered that facilitate larger flux rates, resulting in a scale dependence of effective soil permeability

  20. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

    Braun, Matthias; Löffler, Micha; Plank, Hermann; Asse, Dietmar; Dimmelmeier, Harald

    2014-01-01

    The highest priority for mitigating the consequences of a severe accident with core melt lies in securing containment integrity, as this represents the last barrier against fission product release to the environment. Containment integrity is endangered by several physical phenomena, especially highly transient phenomena following high-pressure reactor pressure vessel failure (like direct containment heating or steam explosions which can lead to early containment failure), hydrogen combustion, quasi-static over-pressure, temperature failure of penetrations, and basemat penetration by core melt. Each of these challenges can be counteracted by dedicated severe accident mitigation hardware, like dedicated primary circuit depressurization valves, hydrogen recombiners or igniters, filtered containment venting, containment cooling systems, and core melt stabilization systems (if available). However, besides their main safety function these systems often have also secondary effects that need to be considered. Filtered containment venting causes (though limited) fission product release into the environment, primary circuit depressurization leads to loss of coolant, and an ex-vessel core melt stabilization system as well as hydrogen igniters can generate high pressure and temperature loads on the containment. To ensure that during a severe accident any available systems are used to their full beneficial extent while minimizing their potential negative impact, AREVA has implemented a severe accident management for German nuclear power plants. This concept makes use of extensive numerical simulations of the entire plant, quantifying the impact of system activations (operational systems, safety systems, as well as dedicated severe accident systems) on the accident progression for various scenarios. Based on the knowledge gained, a handbook has been developed, allowing the plant operators to understand the current state of the plant (supported by computational aids), to predict

  1. Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

    International Nuclear Information System (INIS)

    Verfondern, K.; Martin, R.C.; Moormann, R.

    1993-01-01

    The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.) [de

  2. CONTEMPT-G computer program and its application to HTGR containments

    International Nuclear Information System (INIS)

    Macnab, D.I.

    1976-03-01

    The CONTEMPT-G computer program has been developed by General Atomic Company to simulate the temperature-pressure response of a containment atmosphere to postulated depressurization of High-Temperature Gas-Cooled Reactor (HTGR) primary or secondary coolant circuits. The mathematical models currently used in the code are described, and applications of the code in examples of the atmospheric response of a representative containment to a variety of postulated HTGR accident conditions are presented. In particular, maximum containment temperature and pressure, equilibrated long-term prestressed concrete reactor vessel and containment pressures, and peak containment conditions following steam pipe ruptures are examined for a representative 770-MW(e) HTGR

  3. Triaxial testing system for pressure core analysis using image processing technique

    Science.gov (United States)

    Yoneda, J.; Masui, A.; Tenma, N.; Nagao, J.

    2013-11-01

    In this study, a newly developed innovative triaxial testing system to investigate strength, deformation behavior, and/or permeability of gas hydrate bearing-sediments in deep sea is described. Transport of the pressure core from the storage chamber to the interior of the sealing sleeve of a triaxial cell without depressurization was achieved. An image processing technique was used to capture the motion and local deformation of a specimen in a transparent acrylic triaxial pressure cell and digital photographs were obtained at each strain level during the compression test. The material strength was successfully measured and the failure mode was evaluated under high confining and pore water pressures.

  4. Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-04

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Nakamura, Hideo; Saeki, Hiroyuki

    1991-03-01

    Experimental data for the 10% hot leg break test, Run SB-HL-04, conducted on March 29, 1988 at the ROSA-IV Large Scale Test Facility (LSTF), are presented. This test was conducted as part of test series which studied the effect of break orientation on 10% hot leg break transient, and represented a vertical upward break. Other two tests in this test series represented horizontal break and vertical downward break, respectively. The results of these tests were characterized by asymmetric loop responses, flashing in the cold legs as well as upper downcomer, and condensation depressurization in the cold legs following injection of emergency core coolant (ECC) from accumulators. (author)

  5. Thermal-hydraulic behavior on break simulation of steam generator U-tube

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Bang, Young Seok; Lee, Sukho; Kim, Hho Jung

    1995-01-01

    The thermal-hydraulic behavior depending on the break simulation in a steam generator U-tube was investigated and identified the code predictability on plant responses during SGTR accident. The calculated results were compared and assessed with LSTF SB-SG-06 test data. The RELAP5/MOD3.1 code well predicted the sequence of events and the significant phenomena, such as the asymmetric loop behavior, the RCS cooldown and heat transfer by the natural circulation, and system depressurization, even though there were some differences from the experimental data. The break flowrate was found to be sensitive to the break model and affected the system behavior

  6. Development of aluminothermic welding technology for assembly of reinforcement blocks for nuclear power plants

    International Nuclear Information System (INIS)

    Janicko, S.

    1984-01-01

    The newly developed technology of aluminothermic welding of reinforcing bars is used for the construction of reinforcements, turbine foundations, platforms, bubbling depressurization towers, etc. The method makes possible a good welding of the ends of the reinforcement and reproducibility of welded joints properties without the influence of the human factor on quality. The chemical composition of the weld has a higher content of C and Si which is important with regard to strength. Tensile strength, hardness (ranging from 188 to 270 HV 300), notch toughness (11 to 113 J.cm -2 ) and integrity were assessed. The savings achieved by introduction of the new technology are 1.783 h/weld. (J.H.)

  7. CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application

    International Nuclear Information System (INIS)

    Mankamo, T.; Bjoere, S.; Olsson, Lena

    1992-12-01

    Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems

  8. Alternate method of decay-heat removal in a C-E plant following a SBLOCA

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1986-01-01

    The use of an atmospheric steam-dump procedure to cool and depressurize a Combustion-Engineering plant, Calvert Cliffs-1, following small-break loss-of-coolant accidents with failure of the high-pressure injection system to operate has been investigated. The procedure was effective in depressuizing the primary to the low-pressure injection system operating pressure and design temperature using water supplies from the safety-grade condensate water storage tank only. The procedure was found to be effective even if additional failures occurred. Specifically, low-pressure injection conditions were attained if only a single atmospheric dump valve was available or if the safety-injection tanks (accumulators) were not available

  9. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

  10. The use of supercritical carbon dioxide for contaminant removal from solid waste

    International Nuclear Information System (INIS)

    Adkins, C.L.J.; Russick, E.M.; Smith, H.M.; Olson, R.B.

    1994-01-01

    Supercritical carbon dioxide is being explored as a waste minimization technique for separating oils, greases and solvents from solid waste. The containments are dissolved into the supercritical fluid and precipitated out upon depressurization. The carbon dioxide solvent can then be recycled for continued use. Definitions of the temperature, pressure, flowrate and potential co-solvents are required to establish the optimum conditions for hazardous contaminant removal. Excellent extractive capability for common manufacturing oils, greases, and solvents has been observed in both supercritical and liquid carbon dioxide. Solubility measurements are being used to better understand the extraction process, and to determine if the minimum solubility required by federal regulations is met

  11. A geotechnical investigation of a deep ocean site

    International Nuclear Information System (INIS)

    Freeman, T.J.; Schuttenhelm, R.T.E.

    1990-01-01

    A site investigation in deep water often relies solely on laboratory tests to evaluate the geotechnical properties of the sediments. This imposes two fundamental limitations on the investigation: The maximum depth to which the properties can be profiled and the uncertainty of sample disturbance and de-pressurization effects on the measured data. This paper uses results from investigations performed in a water depth of 5.4 km at an abyssal plain site, Great Meteor East (GME), to illustrate how ambiguities can arise in laboratory measurements of strength, and discusses how a simple in-situ test, the expendable penetrator, can be used to corroborate the laboratory data

  12. Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

    International Nuclear Information System (INIS)

    Hadid Subki, M.; )

    2014-01-01

    The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

  13. Water spray interaction with air-steam mixtures under containment spray conditions: experimental study in the TOSQAN facility

    Energy Technology Data Exchange (ETDEWEB)

    Porcheron, E.; Lemaitre, P.; Malet, J.; Nuboer, A.; Brun, P.; Bouilloux, L.; Vendel, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Direction de la Surete des Usines, des laboratoires, des transports et des dechets, Saclay, BP 68 - 91192 Gif-sur-Yvette cedex (France)

    2005-07-01

    . Measurements are performed during the depressurization transient state. Experimental results show vessel depressurization combined with steam concentration decrease and gas temperature decrease induced by spray injection. Detailed measurements are performed inside and outside water spray region in order to characterize droplets interaction with air/steam mixture. First, we perform a temporal analysis far away from the nozzle and we show that steam condensation on water droplets occurs very quickly at the beginning of spray injection that corresponds to a strong vessel depressurization. Secondly, we study the spatial evolution of droplets temperature when vessel pressure tends toward equilibrium. Droplet temperatures are measured at different distances from the nozzle and we identify, using Spalding parameter, the spray regions where occur steam condensation and droplet vaporization. The full paper will present extended local results performed on two different spray test conditions and interpretation on heat and mass transfers between a water spray and the surrounding air/steam atmosphere. (authors)

  14. 布尔台矿42201工作面底板砂岩承压水防治技术%Prevention and Control Technique for Confined Water of Floor Sandstone at 42201 Working Face of Buertai Coal Mine

    Institute of Scientific and Technical Information of China (English)

    高振宇; 何渊; 任建刚; 刘小雄

    2017-01-01

    为了解决布尔台矿42201工作面底板砂岩承压水带压开采问题.根据勘探钻孔资料及上部22201工作面开采过程中封闭不良钻孔出水情况,分析了工作面带压开采水文地质条件.通过计算煤层底板采动破坏深度、有效保护层厚度、阻水能力及突水系数,评价了工作面带压开采威胁程度.采取直流电法探测对底板砂岩含水层富水性异常垂向、平面分布范围、特征进行探查,并对重点富水异常区域进行钻探验证及疏水降压.对工作面内钻孔封孔质量进行了评价,预计了工作面开采时封闭不良钻孔的透水量,并对封闭不良钻孔区域的底板砂岩含水层实施疏水降压及注浆改造.经注浆检查孔验证,该区域疏水降压及注浆改造效果较好,可实现工作面安全开采.%In order to solve the problems of mining under pressure of floor sandstone confined water existing at 42201 working face of Buertai Coal Mine,this paper analyzes the hydrogeology condition of 42201 working face based on the data of exploration drilling and effluent of poor sealing holes at upper 22201 working face in the process of mining.And the threatening degree of mining under pressure is evaluated through the calculation data including damage depth of destroyed floor,thickness of effective protection layer,water-resisting ability and water burst coefficient.The direct current prospecting technique is used to detect abnormal vertical and plane distribution scope,characteristics of the watery of floor sandstone aquifer,and drilling verification and draining depressurization are carried out in key water enrichment anomaly area.We also evaluated the sealing quality and predicted the permeable capacity of poor sealing holes of working face,and carried out draining depressurization and grouting measures to seal the area of poor sealing holes for floor sandstone confined water.Finally,the measures of grouting and draining depressurization have

  15. Fission product plateout and liftoff in the MHTGR primary system: A review

    International Nuclear Information System (INIS)

    Wichner, R.P.

    1991-04-01

    A review is presented of the technical basis for predicting radioactivity release resulting from depressurization of an MHTGR primary system. Consideration is restricted to so called dry events with no involvement of the steam system. The various types of deposition mechanisms effective for iodine, cesium, strontium, and silver are discussed in terms of their chemical characteristics and the nature of the materials in the primary system. Emphasis is given to iodine behavior, including means for estimating the quantity available for release, the types of plateout locations in the primary system, and the effect of dust on distribution and release. The behavior of fission products cesium, strontium, and silver in such accidents is presented qualitatively. A major part of the review deals with expected dust levels, types, and transport. Available information on the level and nature of dust in the HTGR primary system is reviewed. A summary is presented of dust deposition and liftoff mechanisms. It was concluded that recent approaches to dust liftoff modeling, based on turbulent burst concepts for removal from surfaces, probably offer advantages over the current shear ratio approach. This study concludes that iodine releases from dry depressurization events are likely to be extremely low, on the order of millicuries, due to a predictably low degree of chemical desorption, a low degree of dust liftoff, and a low involvement of iodine with dust. It was also concluded that deposition mechanisms controlling the distribution of fission product material in the primary system, and hence also controlling the degree of liftoff, depend strongly on the chemical nature of the individual elements. Therefore contrary to the current practice, both plateout and liftoff models should reflect those unique chemical and physical properties. 56 refs., 16 figs., 23 tabs

  16. Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

    International Nuclear Information System (INIS)

    Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

    2011-01-01

    Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

  17. Behaviour of gas production from type 3 hydrate reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Pooladi-Darvish, M. [Calgary Univ., AB (Canada). Dept. of Chemical and Petroleum Engineering]|[Fekete Associates Inc., Calgary, AB (Canada); Zatsepina, O. [Calgary Univ., AB (Canada). Dept. of Chemical and Petroleum Engineering; Hong, H. [Fekete Associates Inc., Calgary, AB (Canada)

    2008-07-01

    The possible role of gas hydrates as a potential energy resource was discussed with particular reference to methods for estimating the rate of gas production from hydrate reservoirs under different operating conditions. This paper presented several numerical simulations studies of gas production from type 3 hydrate reservoirs in 1-D and 2-D geometries. Type 3 reservoirs include gas production from hydrate-reservoirs that lie totally within the hydrate stability zone and are sandwiched by impermeable layers on top and bottom. The purpose of this study was to better understand hydrate decomposition by depressurization. The study questioned whether 1-D modeling of type 3 hydrate reservoirs is a reasonable approximation. It also determined whether gas rate increases or decreases with time. The important reservoir characteristics for determining the rate of gas production were identified. Last, the study determined how competition between fluid and heat flow affects hydrate decomposition. This paper also described the relation and interaction between the heat and fluid flow mechanisms in depressurization of type 3 hydrate reservoirs. All results of 1-D and 2-D numerical simulation and analyses were generated using the STARS simulator. It was shown that the rate of gas production depends on the initial pressure/temperature conditions and permeability of the hydrate bearing formation. A high peak rate may be achieved under favourable conditions, but this peak rate is obtained after an initial period where the rate of gas production increases with time. The heat transfer in the direction perpendicular to the direction of fluid flow is significant, requiring 2D modeling. The hydraulic diffusivity is low because of the low permeability of hydrate-bearing formations. This could result in competition between heat and fluid flow, thereby influencing the behaviour of decomposition. 6 refs., 3 tabs., 12 figs.

  18. Structural relaxation and thermal conductivity of high-pressure formed, high-density di-n-butyl phthalate glass and pressure induced departures from equilibrium state.

    Science.gov (United States)

    Johari, G P; Andersson, Ove

    2017-06-21

    We report a study of structural relaxation of high-density glasses of di-n-butyl phthalate (DBP) by measuring thermal conductivity, κ, under conditions of pressure and temperature (p,T) designed to modify both the vibrational and configurational states of a glass. Various high-density glassy states of DBP were formed by (i) cooling the liquid under a fixed high p and partially depressurizing the glass, (ii) isothermal annealing of the depressurized glass, and (iii) pressurizing the glass formed by cooling the liquid under low p. At a given low p, κ of the glass formed by cooling under high p is higher than that of the glass formed by cooling under low p, and the difference increases as glass formation p is increased. κ of the glass formed under 1 GPa is ∼20% higher at ambient p than κ of the glass formed at ambient p. On heating at low p, κ decreases until the glass to liquid transition range is reached. This is the opposite of the increase in κ observed when a glass formed under a certain p is heated under the same p. At a given high p, κ of the low-density glass formed by cooling at low p is lower than that of the high-density glass formed by cooling at that high p. On heating at high p, κ increases until the glass to liquid transition range is reached. The effects observed are due to a thermally assisted approach toward equilibrium at p different from the glass formation p. In all cases, the density, enthalpy, and entropy would change until the glasses become metastable liquids at a fixed p, thus qualitatively relating κ to variation in these properties.

  19. FINAL REPORT on Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Hee C. NO; Nam Z. Cho

    2011-01-01

    The U.S. Department of Energy is performing research and development that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Generation IV very high temperature reactor (VHTR). Phenomena Identification and Ranking studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important. Consequently, the development of advanced air ingress-related models and verification & validation are of very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air ingress will occur through the break, leading to oxidation of the in-core graphite structure and fuel. This study indicates that depending on the location and the size of the pipe break, the air ingress phenomena are different. In an effort to estimate the proper safety margin, experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model are required. It will also require effective strategies to mitigate the effects of oxidation, eventually. This 3-year project (FY 2008–FY 2010) is focused on various issues related to the VHTR air-ingress accident, including (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the core bottom structures, (d) structural tests of the oxidized core bottom structures, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

  20. TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Sekhri, A.; Baumann, P. [KernkraftwerkLeibstadt AG, 5325 Leibstadt (Switzerland); Wicaksono, D. [Swiss Federal Inst. of Technology Zurich ETH, 8092 Zurich (Switzerland); Miro, R.; Barrachina, T.; Verdu, G. [Inst. for Industrial, Radiophysical and Environmental Safety ISIRYM, Universitat Politecnica de Valencia UPV, Cami de Vera s/n, 46021 Valencia (Spain)

    2012-07-01

    In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

  1. TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

    International Nuclear Information System (INIS)

    Sekhri, A.; Baumann, P.; Wicaksono, D.; Miro, R.; Barrachina, T.; Verdu, G.

    2012-01-01

    In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

  2. Toward resolving model-measurement discrepancies of radon entry into houses

    International Nuclear Information System (INIS)

    Garbesi, K.

    1993-01-01

    My dissertation research investigated the source of the model-measurement discrepancy via carefully controlled field experiments conducted at an experimental basement located in natural soil in Ben Lomond, California. Early experiments at the structure (Chapter II) confirmed the existence and magnitude of the model-measurement discrepancy, ensuring that it was not merely an artifact of inherently complex and poorly understood field sites. The measured soil-gas entry rate during structure depressurization was found to be an order of magnitude larger than predicted by a current three-dimensional numerical model of radon transport. The exact magnitude of the discrepancy depends on whether the arithmetic or geometric mean of the small-scale measurements of permeability is used to estimate the effective permeability of the soil. This factor is a critical empirical input to the model and was determined for the Ben Lomond site in the typical fashion using single-probe static depressurization measurement at multiple locations. The remainder of the dissertation research tests a hypothesis to explain the observed discrepancy: that soil permeability assessed using relatively small-scale probe measurements (0.1-0.5 m) does not reflect bulk soil permeability for flows that is likely to occur at larger scales of several meters or more in real houses and in the test structure. The idea is that soil heterogeneity is of a nature that, as flows occur over larger scales, larger scales of heterogeneity are encountered that facilitate larger flux rates, resulting in a scale dependence of effective soil permeability. In Chapter III, I describe the development of a dual-probe dynamic pressure technique to measure soil permeability to air (and anisotropy of permeability) at various length scales. Preliminary field tests of the apparatus indicated that soil permeability was indeed scale dependent

  3. The establishment of MELCOR/SNAP model of Chinshan nuclear power plant for Ultimate Response Guideline

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Wen-Sheng, E-mail: wshsu@ess.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Chiang, Yu, E-mail: s101013702@m101.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Jong-Rong, E-mail: jongrongwang@gmail.com [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Ting-Yi, E-mail: minired1119@gmail.com [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Te-Chuan, E-mail: tcwang@iner.gov.tw [Institute of Nuclear Energy Research Atomic Energy Council, R.O.C., 1000, Wenhua Road Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong, E-mail: jyhtong@cycu.edu.tw [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd, Chung Li 32023, Taiwan, ROC (China); Chen, Shao-Wen, E-mail: chensw@mx.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); and others

    2017-01-15

    Highlights: • The establishment of a MELCOR/SNAP model of Chinshan (BWR/4). • MELCOR/SNAP model was used to estimate the effectiveness of URG for Chinshan. • The MELCOR results were compared to MAAP, TRACE and PCTRAN. • URG is a new method to prevent a Fukushima-like accident. • The low raw water (150 GPM) can make the cladding temperature below 1088.7 K. - Abstract: After Fukushima Daiichi disaster, the safety analysis of severe accidents became one of the safety concerns in Taiwan. The Emergency Operating Procedure (EOP) cannot cope with a multiple system failure situation under a severe accident since it is a “Symptom-basis” procedure. To deal with that, Taiwan Power Company built up a new strategy for Fukushima-like accident called Ultimate Response Guideline (URG). It is a simple strategy with three main conditions: loss of regular motor driven injection system, loss of all AC power and tsunami/earthquake warning. If two of three happen, the operating procedure will change from EOP to URG and start the main works by following the strategy. There are three main works in URG: controlled-depressurization, line up low pressure injection water and prepare containment venting. In this study, MELCOR2.1 was used to calculate the cases of URG and checked the goal of the strategy that prevents the accident or not. There were three steps in this research. First, a model of Chinshan nuclear power plant (NPP) was built. Second, one was the case with URG and the other was not by using the above MELCOR model. The results were compared to MAAP5.0, TRACE and PCTRAN. Finally, some sensitivity studies of depressurization and water injection rate were done.

  4. Thermal-hydraulic modeling needs for passive reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

    1997-07-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

  5. Thermal-hydraulic modeling needs for passive reactors

    International Nuclear Information System (INIS)

    Kelly, J.M.

    1997-01-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken

  6. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  7. Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-11-01

    GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

  8. 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

  9. Recommendations of the Reaktorsicherheitskommission, adopted at the 226th meeting on October 21, 1987

    International Nuclear Information System (INIS)

    1988-01-01

    Internal emergency protection covers all measures that use in a flexible manner the existing operating and safety systems in the plant, for controlling or limiting design-exceeding incidents, and the consequences of such incidents. The following additional measures are planned or already installed: - Depressurization of the containment through a filter system. - Filtering of control room supply air. - Discharge time of the batteries of the emergency power system to be between 2 and 3 hours. - Use of the turbo-generator driven feed system of the BWR design type 69, in case of assumed, simultaneous failure of the auxiliary electrical system and of the diesel emergency sets. - Inerting of the containment of BWR of the 69 design type. (orig./HP) [de

  10. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  11. TRAC analysis of the Crystal River Unit-3 Plant transient of February 26, 1980

    International Nuclear Information System (INIS)

    Coddington, P.; Willcutt, G.J.E. Jr.

    1983-01-01

    This paper describes the application of the TRAC-PD2 and TRAC-PF1 codes to analyze the Crystal River transient. The PD2 and PF1 analyses used the three-dimensional and one-dimensional vessel models, respectively. Both calculations predicted the plant depressurization caused by the open PORV and the subsequent repressurization caused by closing the PORV and continuing high-pressure injection flow. Also, natural circulation was calculated in loop B following reestablishment of feedwater to the loop-B steam generator. After system repressurization, the codes calculated that pressure was relieved through the safety valves, and an intermittent flow occurred in loop A because of high-pressure-injection-driven density variations

  12. Thermosyphon Phenomenon as an alternate heat sink of Shutdown Cooling System for the CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jonghyun [GNEST, Seoul (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

    2006-07-01

    During the outage(overhaul) of the CANDU plant, there is a period when the coolant is partially drained to the reactor header level and the coolant is cooled and depressurized by Shutdown Cooling System(SDCS) other than PHTS pump. In the postulated accident of the loss of SDCS-the PHTS pump failure, the primary coolant system should be cooled by the alternate heat sink using the thermosyphon pheonomenon(TS) through the steam generator(SG) This study was aimed at verification and analyzing the core cooling ability of the TS. And the sensitivity analysis was done for the number of SGs used in the TS. As an analysis tool, RELAP5/CANDU was used.

  13. Reactor coolant system hydrostatic test and risk analysis for the first AP1000 unit

    International Nuclear Information System (INIS)

    Cao Hongjun; Yan Xiuping

    2013-01-01

    The cold hydrostatic test scheme of the primary coolant circuit, of the first AP1000 unit was described. Based on the up-stream design documents, standard specifications and design technical requirements, the select principle of test boundary was identified. The design requirements for water quality, pressure, temperature and temporary hydro-test pump were proposed. A reasonable argument for heating and pressurization rate, and cooling and depressurization rate was proposed. The possible problems and risks during the hydrostatic test were analyzed. This test scheme can provide guidance for the revisions and implementations of the follow-up test procedures. It is a good reference for hydrostatic tests of AP1000 units in the future in China. (authors)

  14. Interaction of Al2O3xSiO2 alloyed uranium oxide with pyrocarbon coating of fuel particles under irradiation

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Khromov, Yu.F.; Svistunov, D.E.; Chujko, E.E.

    1989-01-01

    Method of comparative data analysis for P O2 and P CO was used to consider interaction in fuel particle between pyrocarbon coating and fuel sample, alloyed with alumosilicate addition. Equations of interaction reactions for the case of hermetic and depressurized fuel particle are presented. Calculations of required xAl 2 O 3 XySiO 2 content, depending on oxide fuel burnup, were conducted. It was suggested to use silicon carbide for limitation of the upper level of CO pressure in fuel particle. Estimation of thermal stability of alumosilicates under conditions of uranium oxide burnup equals 1100 and 1500 deg C for Al/Si ratio in addition 1/1 and 4/1 respectively

  15. Fission product release from HTGR coated microparticles and fuel elements

    International Nuclear Information System (INIS)

    Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

    1991-01-01

    The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

  16. Retrogradation of Maize Starch after High Hydrostatic Pressure Gelation: Effect of Amylose Content and Depressurization Rate

    KAUST Repository

    Yang, Zhi; Swedlund, Peter; Gu, Qinfen; Hemar, Yacine; Chaieb, Saharoui

    2016-01-01

    High hydrostatic pressure (HHP) has been employed to gelatinize or physically modify starch dispersions. In this study, waxy maize starch, normal maize starch, and two high amylose content starch were processed by a HHP of the order of 600 MPa

  17. Depressurization and boiling of a single magmatic fluid as a mechanism for tin-tungsten deposit formation

    OpenAIRE

    Maximilian Korges; Philipp Weis; Volker Lüders; Oscar Laurent

    2018-01-01

    Tin (Sn) and tungsten (W) mineralization are commonly associated with each other in relation to highly evolved granites, but economical ore grades are restricted to rare global occurrences and mineralization styles are highly variable, indicating different mechanisms for ore formation. The Sn-W Zinnwald deposit in the Erzgebirge (Germany and the Czech Republic) in the roof zone of a Variscan Li-F granite hosts two contrasting styles of mineralization: (1) cassiterite (Sn) in greisen bodies, a...

  18. Retrogradation of Maize Starch after High Hydrostatic Pressure Gelation: Effect of Amylose Content and Depressurization Rate

    KAUST Repository

    Yang, Zhi

    2016-05-24

    High hydrostatic pressure (HHP) has been employed to gelatinize or physically modify starch dispersions. In this study, waxy maize starch, normal maize starch, and two high amylose content starch were processed by a HHP of the order of 600 MPa, at 25°C for 15min. The effect of HHP processing on the crystallization of maize starches with various amylose content during storage at 4°C was investigated. Crystallization kinetics of HHP treated starch gels were investigated using rheology and FTIR. The effect of crystallization on the mechanical properties of starch gel network were evaluated in terms of dynamic complex modulus (G*). The crystallization induced increase of short-range helices structures were investigated using FTIR. The pressure releasing rate does not affect the starch retrogradation behaviour. The rate and extent of retrogradation depends on the amylose content of amylose starch. The least retrogradation was observed in HHP treated waxy maize starch. The rate of retrogradation is higher for HHP treated high amylose maize starch than that of normal maize starch. A linear relationship between the extent of retrogradation (phase distribution) measured by FTIR and G* is proposed.

  19. Retrogradation of Maize Starch after High Hydrostatic Pressure Gelation: Effect of Amylose Content and Depressurization Rate

    Science.gov (United States)

    Yang, Zhi; Swedlund, Peter; Gu, Qinfen; Hemar, Yacine; Chaieb, Sahraoui

    2016-01-01

    High hydrostatic pressure (HHP) has been employed to gelatinize or physically modify starch dispersions. In this study, waxy maize starch, normal maize starch, and two high amylose content starch were processed by a HHP of the order of 600 MPa, at 25°C for 15min. The effect of HHP processing on the crystallization of maize starches with various amylose content during storage at 4°C was investigated. Crystallization kinetics of HHP treated starch gels were investigated using rheology and FTIR. The effect of crystallization on the mechanical properties of starch gel network were evaluated in terms of dynamic complex modulus (G*). The crystallization induced increase of short-range helices structures were investigated using FTIR. The pressure releasing rate does not affect the starch retrogradation behaviour. The rate and extent of retrogradation depends on the amylose content of amylose starch. The least retrogradation was observed in HHP treated waxy maize starch. The rate of retrogradation is higher for HHP treated high amylose maize starch than that of normal maize starch. A linear relationship between the extent of retrogradation (phase distribution) measured by FTIR and G* is proposed. PMID:27219066

  20. Retrogradation of Maize Starch after High Hydrostatic Pressure Gelation: Effect of Amylose Content and Depressurization Rate.

    Directory of Open Access Journals (Sweden)

    Zhi Yang

    Full Text Available High hydrostatic pressure (HHP has been employed to gelatinize or physically modify starch dispersions. In this study, waxy maize starch, normal maize starch, and two high amylose content starch were processed by a HHP of the order of 600 MPa, at 25°C for 15min. The effect of HHP processing on the crystallization of maize starches with various amylose content during storage at 4°C was investigated. Crystallization kinetics of HHP treated starch gels were investigated using rheology and FTIR. The effect of crystallization on the mechanical properties of starch gel network were evaluated in terms of dynamic complex modulus (G*. The crystallization induced increase of short-range helices structures were investigated using FTIR. The pressure releasing rate does not affect the starch retrogradation behaviour. The rate and extent of retrogradation depends on the amylose content of amylose starch. The least retrogradation was observed in HHP treated waxy maize starch. The rate of retrogradation is higher for HHP treated high amylose maize starch than that of normal maize starch. A linear relationship between the extent of retrogradation (phase distribution measured by FTIR and G* is proposed.

  1. Is activation analysis still active?

    International Nuclear Information System (INIS)

    Chai Zhifang

    2001-01-01

    This paper reviews some aspects of neutron activation analysis (NAA), covering instrumental neutron activation analysis (INAA), k 0 method, prompt gamma-ray neutron activation analysis (PGNAA), radiochemical neutron activation analysis (RNAA) and molecular activation analysis (MAA). The comparison of neutron activation analysis with other analytical techniques are also made. (author)

  2. Performance behavior of the passive containment cooling system of a natural circulation BWR during postulated accident condition

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Nayak, A.K.; Jain, Vikas; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Passive systems are playing prominent role in the development of innovative nuclear reactor systems due to their simplicity, enhanced safety, reliability and economy. These systems are being considered for normal operation as well as accidental conditions of reactor following a postulated accident scenario to preclude the scenarios arising out of failure of active systems as well as to minimize the operator intervention. Indian innovative reactor AHWR being designed for thorium utilization employs various passive safety concepts. As containment is the ultimate barrier to the release of radioactivity, passive concepts are being employed in BWRs for minimize peak containment pressure in the containment during a postulated accident condition like LOCA. The concept of passive containment cooling system (PCCS) in the AHWR comprises of inclined tube heat exchangers located underneath an elevated pool that removes the heat from the steam-air atmosphere of containment following a LOCA by natural circulation of water inside the tubes. The steam condenses on the external surface of tubes of PCCS in addition to the wall of the containment which in turn depressurizes the containment. This paper deals with the performance assessment of PCCS of AHWR during a postulated design basis LOCA by using the best estimate code RELAP5/Mod3.2. (author)

  3. Investigation on accident management measures for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

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

    International Nuclear Information System (INIS)

    1987-01-01

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

  5. Report on the Sunshine Project in fiscal 1990. Studies on liquefaction characteristics by coal types and on engineering properties; 1990 nendo tanshu ni yoru ekika tokusei to kogakuteki busseichi ni kansuru kenkyu hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-07-01

    This paper describes the achievements at the Government Industrial Development Laboratory, Hokkaido in the Sunshine Project in fiscal 1990. Surveys were carried on different catalytic activities for co-liquefaction of various types of coals. The Ru system was found preferable in acquiring fatty family products of low boiling points. Reactions of co-liquefaction constituents do not progress independently, but have positive interaction. Adding Ru is effective in accelerating hydrogenation of condensed aromatic rings and hydrogenation of cleaving nucleus, but in view of the hydrogenating decomposition capability, the Mo-system may be better. In a liquefaction plant test, diesel oil was manufactured by reforming and improving the Yallourn coal light oil. However, the oxygen content was too high, aggravating the temperature distribution control in catalyst bed, and not having achieved the intended denitrification rate. Discussions were given on co-liquefaction of the depressurization residues of Cold Lake (CLVB[vacuum bottom])/Athabasca (ATVB) and Battle River coal. The CLCVB and ATVB were turned into lower molecules easily through hydrogenation treatment. The reaction characteristics reflected the characteristics of medium oil. Depolymerization of coal has progressed faster in ATVB. In the study on the engineering property values, discussions were given on behavior of solid particles in a bubble column, and gas hold-up of slurry. (NEDO)

  6. Preliminary risk assessments of the small HTGR

    International Nuclear Information System (INIS)

    Everline, C.J.; Bellis, E.A.

    1985-05-01

    Preliminary investment and safety risk assessments were performed for a preconceptual design of a four-module 250-MW(t) side-by-side steel-vessel pebble bed HTGR plant. Broad event spectra were analyzed involving component damage resulting in unscheduled plant outages and fission product releases resulting in offsite doses. The preliminary assessment indicates at this stage of the design that two categories of events govern the investment risk envelope: primary coolant leaks which release some circulating and plate-out activity that contaminates the confinement and turbogenerator damage which involves extensive turbine blade failure. Primary coolant leaks are important contributors because associated cleanup and decontamination requirements result in longer outages that arise from other events with comparable frequencies. Turbogenerator damage is the salient low-frequency investment risk accident due to the relatively long outages being experienced in the industry. Thermal transients are unimportant investment risk contributors because pressurized core heatups cause little damage, and depressurized core heatups occur at negligible frequencies relative to the forced outage goal. These preliminary results demonstrate investment and safety risk goal compliance at this stage in the design process. Studies are continuing in order to provide valuable insights into risk-significant events to assure a balanced approach to meeting user and regulatory requirements

  7. Post-test analysis of the ROSA/LSTF and PKL counterpart test

    Energy Technology Data Exchange (ETDEWEB)

    Carlos, S., E-mail: scarlos@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Querol, A., E-mail: anquevi@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Instituto de Seguridad Industrial, Radiofísica y Medioambiental, Universitat Politècnica de València, Camí de Vera, 14, València (Spain); Gallardo, S., E-mail: sergalbe@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Instituto de Seguridad Industrial, Radiofísica y Medioambiental, Universitat Politècnica de València, Camí de Vera, 14, València (Spain); Sanchez-Saez, F., E-mail: frasansa@etsii.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); and others

    2016-02-15

    Highlights: • TRACE modelization for PKL and ROSA/LSTF installations. • Secondary-side depressurization as accident management action. • CET vs PCT relation. • Analysis of differences in the vessel models. - Abstract: Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety. Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper

  8. Post-test analysis of the ROSA/LSTF and PKL counterpart test

    International Nuclear Information System (INIS)

    Carlos, S.; Querol, A.; Gallardo, S.; Sanchez-Saez, F.

    2016-01-01

    Highlights: • TRACE modelization for PKL and ROSA/LSTF installations. • Secondary-side depressurization as accident management action. • CET vs PCT relation. • Analysis of differences in the vessel models. - Abstract: Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety. Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper

  9. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  10. Absorption cycle commercial refrigerator using wood burning cook stove; Geladeira de absorcao acionada por fogao a lenha

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Jose Tomaz Vieira; Martins, Gilberto [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Energia

    1991-12-31

    The current utilization of wood burning cook stoves in Brazil and the socio-economical profile of their users were surveyed. A traditional heavy-mass wood-burning cook stove was studied as a thermal equipment. Simple changes in the geometry of the combustion chamber were suggested to improve the cooking efficiency. A closed two-phase thermosyphon using water as working fluid was designed, built and connected between the combustion chamber of the cook stove and a depressurized absorption refrigeration system to determine the heat flux and the temperature level. A commercial refrigerator unit, using the absorption cycle, was coupled with the wood stove through the thermosyphon. The overall results of the coupling point to successful country-side applications. (author) 12 refs., 9 figs., 4 tabs.

  11. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  12. Analysis of the ANO-2 turbine trip test

    International Nuclear Information System (INIS)

    McDonald, T.A.; Tessier, J.H.; Senda, Y.; Waterman, M.D.

    1983-01-01

    The start-up tests performed with the Arkansas Nuclear One-Unit Two (ANO-2) plant provided an opportunity for studying the validity of certain integral systems codes. In particular, the turbine trip from 98.2 percent full power test was investigated with the RELAP5/MOD1 (cycle 18) ode. A detailed plant model was developed and used to understand the test reports. The early depressurization portion of the transient was reproduced; however, the resultant repressurization was not well represented due to uncertainty in the data and plant response. As a result of these computations and detailed analyses of the test data considerable insight was drawn as to the best way to perform and gather data from such integral systems tests for use in code verification studies

  13. Sensor for measurement of fuel rod gas pressure during loss-of-fluid-tests

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-05-01

    Qualification tests have been conducted of a measurement system for determining the pressure of certain fuel rods in the loss-of-fluid-test (LOFT) reactor. Because of physical size (0.35-in. OD by 5.5-in length) and operational characteristics, an eddy current device was selected as the most promising measurement transducer for the application. The sensor must operate at pressure up to 17.2 MPa (2500 psig) and at temperatures up to 800 0 F. During the reactor transient caused by loss of coolant flow, sensor temperature and applied pressure will vary rapidly and significantly. Consequently, qualification tests included subjection of the sensor to rapid depressurization, temperature transients, and blowdowns in an autoclave, as well as to calibrations and various slow temperature cycles

  14. Experiment on performance of upper head injection system with ROSA-II

    International Nuclear Information System (INIS)

    1976-09-01

    Thermo-hydraulic behavior in the primary cooling system of a pressurized water reactor with an upper head injection system (UHI) in a postulated loss-of-coolant accident (LOCA) has been studied with ROSA-II test facility. Simulated UHI and internal structures of the pressure vessel were installed to the facility for the experiment. Nine maximum-sized double-ended break tests and one medium-sized split break test were performed for the cold-leg break condition. The results are as follows: (1) Fluid mixing in the upper head is not perfect. (2) Cold water injection into the steam or two-phase fluid causes violent depressurization due to the condensation. Flow pattern in the primary cooling system is largely influenced by the above two. (auth.)

  15. Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

    1977-02-01

    Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

  16. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P [Nuclear Research Inst. Rez (Switzerland)

    1996-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  17. Separate-effects experiments on the hydrodynamics of air ingress phenomena for the very high temperature reactor

    International Nuclear Information System (INIS)

    Kim, S.; Talley, J.; Yadav, M.; Ireland, A.; Bajorek, S.

    2011-01-01

    The present study performs scaled separate-effects experiments to investigate the hydrodynamics in the air-ingress phenomena following a Depressurized Condition Cooldown in the Very High Temperature Gas-Cooled Reactor. First, a scoping experiment using water and brine is performed. The volumetric exchange rate is measured using a hydrometer, and flow visualizations are performed. Next, Helium-air experiments are performed to obtain three-dimensional oxygen concentration transient data using an oxygen analyzer. It is found that there exists a critical density difference ratio, before which the ingress rate increases linearly with time and after which the ingress rate slows down significantly. In both the water-brine and Helium-air experiments, this critical ratio is found to be approximately 0.7. (author)

  18. Separate-effects experiments on the hydrodynamics of air ingress phenomena for the very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.; Talley, J.; Yadav, M., E-mail: skim@psu.edu [The Pennsylvania State Univ., University Park, Pennsylvania (United States); Ireland, A.; Bajorek, S. [The United States Nuclear Regulatory Commission, Washington DC (United States)

    2011-07-01

    The present study performs scaled separate-effects experiments to investigate the hydrodynamics in the air-ingress phenomena following a Depressurized Condition Cooldown in the Very High Temperature Gas-Cooled Reactor. First, a scoping experiment using water and brine is performed. The volumetric exchange rate is measured using a hydrometer, and flow visualizations are performed. Next, Helium-air experiments are performed to obtain three-dimensional oxygen concentration transient data using an oxygen analyzer. It is found that there exists a critical density difference ratio, before which the ingress rate increases linearly with time and after which the ingress rate slows down significantly. In both the water-brine and Helium-air experiments, this critical ratio is found to be approximately 0.7. (author)

  19. Analysis of some accident conditions in confirmation of the HTGR safety

    Energy Technology Data Exchange (ETDEWEB)

    Grebennik, V. N.; Grishanin, E. I.; Kukharkin, N. E.; Mikhailov, P. V.; Pinchuk, V. V.; Ponomarev-Stepnoy, N. N.; Fedin, G. I.; Shilov, V. N.; Yanushevich, I. V. [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

    1981-01-15

    This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved.

  20. Risk evaluation of the alternate-3A modification to the ATWS prevention/mitigation system in a BWR-4, MARK-II power plant

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Karol, R.; Shiu, K.

    1983-01-01

    The authors present a risk evaluation of the ATWS Alternate 3A modification proposed by NRC staff in NUREG-0460 to the ATWS prevention/mitigation system in a BWR nuclear power plant. The evaluation is done relative to three risk indices: the frequency of core damage, the expected early fatalities, and the expected latent fatalities. The ATWS prevention tree includes: the mechanical subsystem of the reactor protection system, the electrical subsystem of the reactor protection system, the recirculation pump trip and the Alternate Rod Insertion System. The mitigation tree includes: standby liquid control system, opening of the relief valves, reclosing the relief valves, failure of coolant injection, inadvertent actuation of the automatic depressurization system, inadvertent operation of high-pressure injection system and containment heat removal