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Sample records for loss-of-coolant accident loca

  1. Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

    We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

  2. Small break LOCA [loss of coolant accident] mitigation for Bellefonte

    International Nuclear Information System (INIS)

    Bayless, P.D.; Dobbe, C.A.

    1986-01-01

    Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

  3. Cobalt-60 simulation of LOCA [loss of coolant accident] radiation effects

    International Nuclear Information System (INIS)

    Buckalew, W.H.

    1989-07-01

    The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs

  4. Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions

    International Nuclear Information System (INIS)

    Pettersson, Kjell; Chung, Haijung; ); Billone, Michael; Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Hache, George; Papin, Joelle; Heins, Lothar; Hozer, Zoltan; In de Betou, Jan; Kelppe, Seppo; Mayer, Ralph; Scott, Harold; Voglewede, John; Sonnenburg, Heinz; Sunder, Sham; Valach, Mojmir; Vrtilkova, Vera; Waeckel, Nicolas; Wiesenack, Wolfgang; Zimmermann, Martin

    2009-01-01

    The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the current understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burn-up and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including as regards experimental approaches and interpretation and the use of experimental data relevant for safety. In 1986, a working group of the NEA Committee on the Safety of Nuclear Installations (CSNI) issued a state-of-the-art report on water reactor fuel behaviour in design-basis accident (DBA) conditions. The 1986 report was limited to the oxidation, embrittlement and deformation of pressurised water reactor (PWR) fuel in a loss-of-coolant accident (LOCA). Since then, considerable experimental and analytical work has been performed, which has led to a broader and deeper understanding of LOCA-related phenomena. Further, new cladding alloys have been produced, which might behave differently than the previously used Zircaloy-4, both under normal operating conditions and during transients. Compared with 20 years ago, fuel burn-up has been significantly increased, which requires extending the LOCA database in order to cover the high burnup range. There was also a clear need to address LOCA performance for reactor types other than PWRs. The present report has been prepared by the WGFS and covers the following technical aspects: - Description of different LOCA scenarios for major types of reactors: BWRs, PWRs, VVERs and to a lesser extent CANDUs. - LOCA phenomena: ballooning, burst, oxidation, fuel relocation and possible fracture at quench. - Details of high-temperature oxidation behaviour of various cladding materials. - Metallurgical phase change, effect of hydrogen and oxygen on residual cladding ductility. - Methods for LOCA testing, for example two-sided oxidation and ring compression for ductility, and integral quench test for

  5. Fast instrumentation for loss of coolant accident (LOCA) experimental studies pertaining to nuclear reactors

    International Nuclear Information System (INIS)

    Venkat Raj, V.; Sreenivas Rao, G.; Belokar, D.G.; Dolas, P.K.

    1989-01-01

    The loss of coolant accident (LOCA) which involves a breach in the pressure boundary of the primary coolant system (PCS) is one of the postulated accident conditions against which the safety of the reactor system is to be ensured. Mathematical models have been developed to analyse this kind of transients. However, because of the extremely complicated nature of the phenomena involved, it is necessary to validate the analytical models with appropriate experimental data. Many parameters are to be measured during the experiments, out of which temperature, pressure, void fraction and two-phase mass flow rate are the most important parameters. Since the phenomenon is very fast, special fast response instruments are required. This paper deals with the considerations that govern the selection of appropriate instruments and the development of suitable instruments for transient two-phase flow and void fraction measurements. The requirements of the associated fast data acquisition system are also discussed. (author). 4 figs

  6. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    International Nuclear Information System (INIS)

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-01-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  7. Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

    International Nuclear Information System (INIS)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

  8. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    International Nuclear Information System (INIS)

    Nelson, C.F.; Gauthier, G.; Carlin, F.

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40 degrees C or 70 degrees C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased

  9. Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

    International Nuclear Information System (INIS)

    Vigil, R.A.; Jacobus, M.J.

    1994-04-01

    Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent

  10. Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

    Energy Technology Data Exchange (ETDEWEB)

    Fleming, K.N., E-mail: KarlFleming@comcast.net [KNF Consulting LLC, Spokane, WA (United States); Lydell, B.O.Y. [SIGMA-PHASE INC., Vail, AZ (United States)

    2016-08-15

    Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

  11. Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

    International Nuclear Information System (INIS)

    Fleming, K.N.; Lydell, B.O.Y.

    2016-01-01

    Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

  12. Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

    International Nuclear Information System (INIS)

    Nguyen Van Thai; Kieu Ngoc Dung

    2015-01-01

    This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

  13. Compendium of ECCS [Emergency Core Cooling Systems] research for realistic LOCA [loss-of-coolant accidents] analysis: Final report

    International Nuclear Information System (INIS)

    1988-12-01

    In the United States, Emergency Core Cooling Systems (ECCS) are required for light water reactors (LWRs) to provide cooling of the reactor core in the event of a break or leak in the reactor piping or an inadvertent opening of a valve. These accidents are called loss-of-coolant accidents (LOCA), and they range from small leaks up to a postulated full break of the largest pipe in the reactor cooling system. Federal government regulations provide that LOCA analysis be performed to show that the ECCS will maintain fuel rod cladding temperatures, cladding oxidation, and hydrogen production within certain limits. The NRC and others have completed a large body of research which investigated fuel rod behavior and LOCA/ECCS performance. It is now possible to make a realistic estimate of the ECCS performance during a LOCA and to quantify the uncertainty of this calculation. The purpose of this report is to summarize this research and to serve as a general reference for the extensive research effort that has been performed. The report: (1) summarizes the understanding of LOCA phenomena in 1974; (2) reviews experimental and analytical programs developed to address the phenomena; (3) describes the best-estimate computer codes developed by the NRC; (4) discusses the salient technical aspects of the physical phenomena and our current understanding of them; (5) discusses probabilistic risk assessment results and perspectives, and (6) evaluates the impact of research results on the ECCS regulations. 736 refs., 412 figs., 66 tabs

  14. Analysis of risk reduction methods for interfacing system LOCAs [loss-of-coolant accidents] at PWRs

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1988-01-01

    The Reactor Safety Study (WASH-1400) predicted that Interfacing System Loss-of-Coolant Accidents (ISL) events were significant contributors to risk even though they were calculated to be relatively low frequency events. However, there are substantial uncertainties involved in determining the probability and consequences of the ISL sequences. For example, the assumed valve failure modes, common cause contributions and the location of the break/leak are all uncertain and can significantly influence the predicted risk from ISL events. In order to provide more realistic estimates for the core damage frequencies (CDFs) and a reduction in the magnitude of the uncertainties, a reexamination of ISL scenarios at PWRs has been performed by Brookhaven National Laboratory. The objective of this study was to investigate the vulnerability of pressurized water reactor designs to ISLs and identify any improvements that could significantly reduce the frequency/risk of these events

  15. Sensitivity analysis of local uncertainties in large break loss-of-coolant accident (LB-LOCA) thermo-mechanical simulations

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Ikonen, Timo

    2016-08-15

    Highlights: • A sensitivity analysis using the data from EPR LB-LOCA simulations is done. • A procedure to analyze such complex data is outlined. • Both visual and quantitative methods are used. • Input factors related to core design are identified as most significant. - Abstract: In this paper, a sensitivity analysis for the data originating from a large break loss-of-coolant accident (LB-LOCA) analysis of an EPR-type nuclear power plant is presented. In the preceding LOCA analysis, the number of failing fuel rods in the accident was established (Arkoma et al., 2015). However, the underlying causes for rod failures were not addressed. It is essential to bring out which input parameters and boundary conditions have significance to the outcome of the analysis, i.e. the ballooning and burst of the rods. Due to complexity of the existing data, the first part of the analysis consists of defining the relevant input parameters for the sensitivity analysis. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The ultimate goal is to develop a systematic procedure for the sensitivity analysis of statistical LOCA simulation that takes into account the various sources of uncertainties in the calculation chain. In the current analysis, the most relevant parameters with respect to the cladding integrity are the decay heat power during the transient, the thermal hydraulic conditions in the rod’s location in reactor, and the steady-state irradiation history of the rod. Meanwhile, the tolerances in fuel manufacturing parameters were found to have negligible effect on cladding deformation.

  16. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  17. Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

    Energy Technology Data Exchange (ETDEWEB)

    Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

    2006-07-01

    A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

  18. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

  19. Reactor elements properties response during a postulated loss-of-coolant accident (LOCA)

    International Nuclear Information System (INIS)

    Ahmed, E.E.; Rahman, F.A.

    1985-01-01

    Four computer algorithms have been introduced to solve for the reactor different materials response subjected to LOCA conditions, they were developed with the intent of producing a simple, accurate and efficient prediction schemes. A general overview of the solution procedures design and working of each of four algorithms are presented, followed by short description of the nature of solution and calculated results. These algorithms are: 1. ZIRCP to give the cladding material properties response under normal and transient conditions. 2. FCGAPP to give the fuel- cladding gas-gap conductivity. 3. NFUEIP to solve the temperature dependent of nuclear fuel properties during normal and transient conditions. 4. TSDATP has been developed to solve for the thermodynamic and transport properties of water and steam over a large range of temperature and pressure. 14 fig

  20. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

    2013-10-01

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  1. LOFA [loss of flow accident] and LOCA [loss of coolant accident] in the TIBER-II engineering test reactor: Appendix A-4

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510 0 C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs

  2. 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)

  3. Evaluation of the radiative transfer in the core of a Pressurized Water Reactor (PWR) during the reflooding step of a Loss Of Coolant Accident (LOCA)

    International Nuclear Information System (INIS)

    Gerardin, J.

    2012-01-01

    We developed a method of resolution of radiative transfer inside a medium of vapor-droplets surrounded by hot walls, in order to couple it with a simulation of the flow at the CFD scale. The scope is the study of the cooling of the core of nuclear reactor following a Loss Of Coolant Accident (LOCA). The problem of radiative transfer can be cut into two sub problems, one concerning the evaluation of the radiative properties of the medium and a second concerning the solution of the radiative transfer equation. The radiative properties of the droplets have been computed with the use of the Mie Theory and those of the vapor have been computed with a Ck model. The medium made of vapor and droplets is an absorbing, anisotropically scattering, emissive, non grey, non homogeneous medium. Hence, owing to the possible variations of the flow properties (diameter and volumetric fraction of the droplets, temperature and pressure of the vapor), the medium can be optically thin or thick. Consequently, a method is required which solves the radiative transfer accurately, with a moderate calculation time for all of these prerequisites. The IDA has been chosen, derived from the well-known P1-approximation. Its accuracy has been checked on academical cases found in the literature and by comparison with experimental data. Simulations of LOCA flows have been conducted taking account of the radiative transfer, evaluating the radiative fluxes and showing that radiative transfer influence cannot be neglected. (author)

  4. Hydrogen radiolytic production in light and heavy water mixtures under conditions similar to LOCA (loss of coolant accidents)

    International Nuclear Information System (INIS)

    Garcia Rodenas, L.; Ali, S.P.; Liberman, S.J.

    1987-01-01

    H 2 , HD and D 2 radiolytic yield in heavy and light water mixtures has been determined to supply the necessary data which will allow to make a realistic estimation of the solution of such gas under LOCA conditions as a function of time. (Author)

  5. Definition of loss-of-coolant accident radiation source

    International Nuclear Information System (INIS)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist

  6. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of class 1E electrical cables

    International Nuclear Information System (INIS)

    Jacobus, M.J.

    1992-11-01

    This report describes the results of aging, condition monitoring, and accident testing of miscellaneous cable types. Three sets of cables were aged for up to 9 months under simultaneous thermal (≅100 degrees C) and radiation (≅0.10 kGy/hr) conditions. A sequential accident consisting of high dose rate irradiation (≅6 kGy/hr) and high temperature steam followed the aging. Also exposed to the accident conditions was a fourth set of cables, which were unaged. The test results indicate that, properly installed, most of the various miscellaneous cable products tested should be able to survive an accident after 60 years for total aging doses of at least 150 kGy or higher (depending on the material) and for moderate ambient temperatures on the order of 45--55 degrees C (potentially higher or lower, depending on material specific activtion energies and total radiation doses). Mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation

  7. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  8. An experimental study of the corrosion and precipitation of aluminum in the presence of trisodium phosphate buffer following a loss of coolant accident (LOCA) scenario

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Jun [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry J. [Department of Civil Engineering, University of New Mexico (United States); Leavitt, Janet J. [Department of Civil Engineering, University of New Mexico (United States); Alion Science and Technology (United States); Hammond, Kyle; Mitchell, Lana [Department of Civil Engineering, University of New Mexico (United States); Kee, Ernie [South Texas Project Nuclear Operating Company (STPNOC) (United States); Blandford, Edward D., E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States)

    2015-02-15

    Highlights: • Experimental head loss testing was conducted by aggressively promoting corrosion in loss of coolant accidents. • Blender-processed debris beds have higher head loss but tend to be less reproducible than NEI-processed debris beds. • Precipitation was observed from aluminum concentration and turbidity measurements. • Precipitation results were compared to predictions from Visual MINTEQ. - Abstract: This paper presents the results of an integrated chemical effects experiment of head loss across the sump pump screen with fibrous debris bed over a non-prototypical 10-day post-LOCA incident window. The corrosion head loss experiments (CHLE) is a reduced scaled integral effects testing facility built at the University of New Mexico (UNM) to investigate potential chemical effects on head loss across prepared fibrous debris beds. The results in this paper come from two integral effect tests performed at UNM in order to determine the chemical effects on head loss induced by a zinc source effect and an aluminum precipitation effect (T3: without Zn source case, T4: with Zn source case in containment). The tests were performed with a large surface area of aluminum coupons in the testing facility for an extended period of elevated temperature to accelerate corrosion above that expected under prototypical conditions. These conditions were sufficient to force aluminum precipitation to occur and induce the onset of chemical effects on debris bed head loss. The head loss behavior on two different types of fiber debris beds (blender-processed and NEI-processed debris bed) was evaluated in this study. It was found that the blender-processed bed is much more sensitive in filtering than the NEI-processed bed and consequently had a much higher head loss value across the beds. Aluminum precipitation was observed, with aluminum concentration and turbidity measurements, to form starting on day 7 in Test T3 and on day 6 in Test T4. The onset of aluminum precipitation

  9. An experimental study of the corrosion and precipitation of aluminum in the presence of trisodium phosphate buffer following a loss of coolant accident (LOCA) scenario

    International Nuclear Information System (INIS)

    Kim, Seung Jun; Howe, Kerry J.; Leavitt, Janet J.; Hammond, Kyle; Mitchell, Lana; Kee, Ernie; Blandford, Edward D.

    2015-01-01

    Highlights: • Experimental head loss testing was conducted by aggressively promoting corrosion in loss of coolant accidents. • Blender-processed debris beds have higher head loss but tend to be less reproducible than NEI-processed debris beds. • Precipitation was observed from aluminum concentration and turbidity measurements. • Precipitation results were compared to predictions from Visual MINTEQ. - Abstract: This paper presents the results of an integrated chemical effects experiment of head loss across the sump pump screen with fibrous debris bed over a non-prototypical 10-day post-LOCA incident window. The corrosion head loss experiments (CHLE) is a reduced scaled integral effects testing facility built at the University of New Mexico (UNM) to investigate potential chemical effects on head loss across prepared fibrous debris beds. The results in this paper come from two integral effect tests performed at UNM in order to determine the chemical effects on head loss induced by a zinc source effect and an aluminum precipitation effect (T3: without Zn source case, T4: with Zn source case in containment). The tests were performed with a large surface area of aluminum coupons in the testing facility for an extended period of elevated temperature to accelerate corrosion above that expected under prototypical conditions. These conditions were sufficient to force aluminum precipitation to occur and induce the onset of chemical effects on debris bed head loss. The head loss behavior on two different types of fiber debris beds (blender-processed and NEI-processed debris bed) was evaluated in this study. It was found that the blender-processed bed is much more sensitive in filtering than the NEI-processed bed and consequently had a much higher head loss value across the beds. Aluminum precipitation was observed, with aluminum concentration and turbidity measurements, to form starting on day 7 in Test T3 and on day 6 in Test T4. The onset of aluminum precipitation

  10. RELAP5 simulation of a large break Loss of Coolant Accident (LOCA) in the hot leg of the primary system in Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Andrade, Delvonei Alves de; Sabundjian, Gaiane

    2004-01-01

    The objective of this work is to present the simulation of a large break loss of coolant accident - LBLOCA in the hot leg of the primary loop in Angra 2, with RELAP5/MOD3.2.2g code. This accident is described in the Final Safety Report Analysis of Angra 2 - FSAR and consists basically of the hot leg total break, in loop 20 of the plant. The area considered for the rupture is 4480 cm 2 , which corresponds to 100% of the pipe flow area. Besides, this work also has the objective of verifying the efficiency of the emergency core coolant system - ECCS in case of accidents and transients. The thermal-hydraulic processes inherent to the accident phenomenology, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the liquid level, until the ECCS is capable to reflood it

  11. ESBWR long term containment response to loss of coolant accidents

    International Nuclear Information System (INIS)

    Alamgir, M. D.; Marquino, W.; Diaz-Quiroz, J.; Tucker, L.

    2010-01-01

    ESBWR is a 4500 MWt generation III+ natural circulation reactor with an array of robust passive safety systems to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US certification process. This paper focuses on the bounding licensing analysis of the long-term (30-day) response of the ESBWR containment to limiting Loss of Coolant Accident (LOCA) performed with the TRACG code. It is shown that using only passive systems available during the first 72 hours after the limiting Main Steam Line Break LOCA, the predicted peak containment pressure in the ESBWR containment remain well below the design limits with good margin. After 72 hours of LOCA initiation, PCCS Vent Fans (non-safety system) become available that remove non-condensable gases from, and further enhance the effectiveness of, PCCS heat exchangers to reduce the containment pressure and temperature to values substantially below the design limits. During the post- 72 hour period, the beneficial effects of the Vent Fan operation, combined with the available operator action to refill of PCCS pools, continue to maintain the containment pressure to about 30% below the design limit at 30 days after a limiting ESBWR LOCA. (authors)

  12. Evaluation of a postulated loss of coolant accident (LOCA) due to a 160 cm2 break in a cold leg of Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

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

    2002-01-01

    The development of a qualified full nodalization of Angra2 NPP for RELAP5/Mod 3.2.2 gamma, aiming at the evaluation of a comprehensive number of accidents and transients, thus providing suitable safety analysis support for licensing purposes, is being carried out within the framework of CNEN internal technical cooperation, involving some of its institutes (CDTN, IPEN and IEN) and the Reactors Coordination (CODRE). This work presents a simulation of a postulated Angra2 small cold leg break loss of coolant accident (SBLOCA). A 160 cm 2 break is supposed to occur at one cold leg between the main coolant pump and the reactor vessel and is described in the Angra2 Final Safety Analysis Report, section 15.6.4.1.3.4. The simulation of several types of transients and accidents is necessary to verify the adequate performance of the modeled logic and systems. In general, the analysis of such and accident allows to demonstrate the safety Injection System performance and the reliable transition between the high pressure safety injection, the accumulator injection and the residual heat removal phases. Furthermore, it is assumed that some components are out of service due to fail or repair in order to make a conservative analysis. The results showed a compatible behavior of the molded systems and that the simulated Emergency Core Cooling System was able to provide sufficient cooling to avoid any damage to the core. (author)

  13. Simulation of a loss of coolant accident

    International Nuclear Information System (INIS)

    1987-06-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

  14. Research on loss of coolant accident of pressurized-water reactor based on PSO algorithm

    International Nuclear Information System (INIS)

    Ma Jie; Guo Lifeng; Peng Qiao

    2012-01-01

    In order to improve the diagnosis performance of Loss of Coolant Accident (LOCA), based on Back Propagation (BP) algorithm study, a fault diagnosis network is established based on Particle Swarm Optimization (PSO) algorithm in this paper. The PSO algorithm is used to train the weights and the thresholds of neural network, which can conquer part convergence problem of BP algorithm. The test results show that the diagnosis network has higher accuracy of LOCA. (authors)

  15. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

  16. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW) [de

  17. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  18. Tools evaluation and development for loss of coolant accidents analysis in research reactors

    International Nuclear Information System (INIS)

    Maprelian, Eduardo; Cabral, Eduardo L.L.; Silva, Antonio T. e

    1999-01-01

    The loss of coolant accidents (LOCA) in pool type research reactors are normally considered as limiting in the licensing process. This paper verifies the viability of the computer code 3D-AIRLOCA to analyze LOCA in a pool type research reactor, and also develops two computer codes LOSS and TEMPLOCA. The computer code LOSS determines the time tom drawn the pool down to the level of the bottom of the core, and the computer code TEMPLOCA calculates the peak fuel element temperature during the transient. These two coders substitutes the 3D-AIRLOCA in the LOCA analysis for pool type research reactors. (author)

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  20. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde; Simulacion de escenarios de accidente severo tipo perdida de refrigerante (Loca) pequeno, con el codigo MELCOR version 2.1 para la central nucleo-electrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft{sup 2} (4.6 cm{sup 2}), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

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

    International Nuclear Information System (INIS)

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

    1991-08-01

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

  2. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  3. Review on research of small break loss of coolant accident

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

    The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

  4. Loss-of-coolant accident analysis of the Savannah River new production reactor design

    International Nuclear Information System (INIS)

    Maloney, K.J.; Pryor, R.J.

    1990-11-01

    This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

  5. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  6. Transient Analysis for Evaluating the Potential Boiling in the High Elevation Emergency Cooling Units of PWR Following a Hypothetical Loss of Coolant Accident (LOCA) and Subsequent Water Hammer Due to Pump Restart

    International Nuclear Information System (INIS)

    Husaini, S. Mahmood; Qashu, Riyad K.

    2004-01-01

    The Generic Letter GL-96-06 issued by the U.S. Nuclear Regulatory Commission (NRC) required the utilities to evaluate the potential for voiding in their Containment Emergency Cooling Units (ECUs) due to a hypothetical Loss Of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) accompanied by the Loss Of Offsite Power (LOOP). When the offsite power is restored, the Component Cooling Water (CCW) pumps restart causing water hammer to occur due to cavity closure. Recently EPRI (Electric Power Research Institute) performed a research study that recommended a methodology to mitigate the water hammer due to cavity closure. The EPRI methodology allows for the cushioning effects of hot steam and released air, which is not considered in the conventional water column separation analysis. The EPRI study was limited in scope to the evaluation of water hammer only and did not provide any guidance for evaluating the occurrence of boiling and the extent of voiding in the ECU piping. This paper presents a complete methodology based on first principles to evaluate the onset of boiling. Also, presented is a methodology for evaluating the extent of voiding and the water hammer resulting from cavity closure by using an existing generalized computer program that is based on the Method of Characteristics. The EPRI methodology is then used to mitigate the predicted water hammer. Thus it overcomes the inherent complications and difficulties involved in performing hand calculations for water hammer. The heat transfer analysis provides an alternative to the use of very cumbersome modeling in using CFD (computational fluid dynamics) based computer programs. (authors)

  7. Analysis of small break loss of coolant accident for Chinese CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

    2016-05-15

    This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

  8. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  9. Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

    In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

  10. Loss-of-coolant accident for large pipe breaks in light water reactor plants

    International Nuclear Information System (INIS)

    Keusenhoff, J.

    1980-01-01

    The importance of loss-of-coolant accidents (LOCA) and their control for nuclear reactor safety is explained. Showing the cooling circuits and emergency core cooling systems (ECCS) of both, PWR and BWR, the possible break spectrum and the general sequence of events is discussed. The governing physical phenomena for the different LOCA phases are pointed out in more detail. Special emphasis is taken on rules, regulations and failure criteria for licensing purposes. Analysis methods and codes for both, evaluation and best-estimate model are compared under deterministic and probabilistic approach, respectively. Some insight in present integral and separate effect tests demonstrates the interdependency of analysis and experiment. Results of LOCA analysis and experiments show the present state of the art. (orig.)

  11. Break spectrum analyses for small break loss of coolant accidents in a RESAR-3S Plant

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Kullberg, C.M.

    1986-03-01

    A series of thermal-hydraulic analyses were performed to investigate phenomena occurring during small break loss-of-coolant-accident (LOCA) sequences in a RESAR-3S pressurized water reactor. The analysis included simulations of plant behavior using the TRAC-PF1 and RELAP5/MOD2 computer codes. Series of calculations were performed using both codes for different break sizes. The analyses presented here also served an audit function in that the results shown here were used by the US Nuclear Regulatory Commission (NRC) as an independent confirmation of similar analyses performed by Westinghouse Electric Company using another computer code. 10 refs., 62 figs., 14 tabs

  12. Analysis of forces on core structures during a loss-of-coolant accident. Final report

    International Nuclear Information System (INIS)

    Griggs, D.P.; Vilim, R.B.; Wang, C.H.; Meyer, J.E.

    1980-08-01

    There are several design requirements related to the emergency core cooling which would follow a hypothetical loss-of-coolant accident (LOCA). One of these requirements is that the core must retain a coolable geometry throughout the accident. A possible cause of core damage leading to an uncoolable geometry is the action of forces on the core and associated support structures during the very early (blowdown) stage of the LOCA. An equally unsatisfactory design result would occur if calculated deformations and failures were so extensive that the geometry used for calculating the next stages of the LOCA (refill and reflood) could not be known reasonably well. Subsidiary questions involve damage preventing the operation of control assemblies and loss of integrity of other needed safety systems. A reliable method of calculating these forces is therefore an important part of LOCA analysis. These concerns provided the motivation for the study. The general objective of the study was to review the state-of-the-art in LOCA force determination. Specific objectives were: (1) determine state-of-the-art by reviewing current (and projected near future) techniques for LOCA force determination, and (2) consider each of the major assumptions involved in force determination and make a qualitative assessment of their validity

  13. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    International Nuclear Information System (INIS)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul

    2015-01-01

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree

  14. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

  15. Condensing heat transfer following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Krotiuk, W.J.; Rubin, M.B.

    1978-01-01

    A new method for calculating the steam mass condensation energy removal rates on cold surfaces in contact with an air-steam mixture has been developed. This method is based on the principles of mass diffusion of steam from an area of high concentration to the condensing surface, which is an area of low steam concentration. This new method of calculating mass condensation has been programmed into the CONTEMPT-LT Mod 26 computer code, which calculates the pressure and temperature transients inside a light water reactor containment following a loss-of-coolant accident. The condensing heat transfer coefficient predicted by the mass diffusion method is compared to existing semi-empirical correlations and to the experimental results of the Carolinas Virginia Tube Reactor Containment natural decay test. Closer agreement with test results is shown in the calculation of containment pressure, temperature, and heat sink surface temperature using the mass diffusion condensation method than when using any existing semi-empirical correlation

  16. Analysis of loss of coolant accident and emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

    1977-01-01

    In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

  17. Responses to Small Break Loss of Coolant Accidents for SMART

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

    2004-01-01

    The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  19. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

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

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  20. Estimation of break location and size for loss of coolant accidents using neural networks

    International Nuclear Information System (INIS)

    Na, Man Gyun; Shin, Sun Ho; Jung, Dong Won; Kim, Soong Pyung; Jeong, Ji Hwan; Lee, Byung Chul

    2004-01-01

    In this work, a probabilistic neural network (PNN) that has been applied well to the classification problems is used in order to identify the break locations of loss of coolant accidents (LOCA) such as hot-leg, cold-leg and steam generator tubes. Also, a fuzzy neural network (FNN) is designed to estimate the break size. The inputs to PNN and FNN are time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. An automatic structure constructor for the fuzzy neural network automatically selects the input variables from the time-integrated values of many measured signals, and optimizes the number of rules and its related parameters. It is verified that the proposed algorithm identifies very well the break locations of LOCAs and also, estimate their break size accurately

  1. Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

    International Nuclear Information System (INIS)

    Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

    1995-01-01

    The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

  2. Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

    1995-01-01

    Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

  3. Loss of coolant accident. Past, present and future

    International Nuclear Information System (INIS)

    Cermak, J.O.

    1978-01-01

    The history of LOCA is covered from the original design basis, failure of the largest connecting pipe with no peak clad temperature limit through to the current design basis, double ended failure of the largest pipe in the primary system with a design peak clad temperature limit of 2200 0 F. Various obstacles along the way are addressed such as, degree of analytical sophisticaton, perplexing experimental results, the infamous 1971 semiscale tests, fuel densification and changes in USNRC evaluation models. In the future, it is projected that more reliance will be put on probabilistic evaluation of the LOCA with respect to both the accident analysis, reliability of the system and the probability of the accident itself. (author)

  4. Analysis of loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Moldaschl, H.

    1982-01-01

    Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

  5. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

    International Nuclear Information System (INIS)

    1980-01-01

    The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

  6. Source term and behavioural parameters for a postulated HIFAR loss-of-coolant accident

    International Nuclear Information System (INIS)

    May, F.G.

    1987-01-01

    The fraction of the fission product inventory which might be released into the atmosphere of the HIFAR reactor containment building (RCB) during a postulated loss-of-coolant accident (LOCA) has been evaluated as a function of time, for each classification of airborne radioactivity. This appraisal will be used as the source term for a computer program, which uses realistic attenuation of the fission product aerosol in a single compartment model with a defined leakrate to predict possible radioactive releases into the environment in a hypothetical bounding case reactor accident which is rather more severe in all major aspects than any single LOCA. Also given are the parameters governing the attenuation of the aerosol and vapours in the atmosphere of the RCB so that their behaviour may be accurately modelled. The source terms for several other types of accident involving the meltdown of fuel elements have also been considered but in less detail than the LOCA case. In some of the cases, the fission products are released directly to atmosphere, so there is no attenuation of the release by deposition within the RCB

  7. Thermal-hydraulic analysis of loss-of-coolant accident in the JMTR

    International Nuclear Information System (INIS)

    Sakurai, Fumio; Oyamada, Rokuro

    1985-02-01

    The reevaluation of the Loss-of-Coolant Accident (LOCA) was required through the process of a safety review for the Japan Materials Testing Reactor (JMTR) core conversion from the high-enriched uranium fuel (Enrichment : 93%) to the medium-enriched uranium fuel (Enrichment : 45%). The following were concluded by thermal-hydraulic analysis of a LOCA caused by a double-ended pipe break in the JMTR primary cooling system. (1) The fuel in the core does not burn-out as long as it is covered with water. (2) A larger siphon break valve (larger than phi60mm) should be installed instead of the present one (phi25mm) on the primary cooling system in order to prevent the core from being uncovered with water in case of a LOCA caused by a double-ended pipe break. The present siphon break valve was installed to keep the core covered with water in case of a LOCA caused by a small pipe rupture. In this analysis, the Siphon Breaker Analysis Code (SBAC) was written in order to analyse the size of the siphon break valve and its accuracy was confirmed to be within 5% through a verification experiment. (author)

  8. Pressure behaviour in a nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    KHattab, M.S.; Ibrahim, N.A.; Bedrose, S.D.

    1994-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break, is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The scenarios of small, medium and large LOCA at 2%, 15% and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The results of large LOCA showed good agreement with westinghouse calculations of the same design. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs., 1 tab

  9. 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)

  10. Simulation of a loss of coolant accident with hydroaccumulator injection

    International Nuclear Information System (INIS)

    1988-10-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. The IAEA, having identified the need for experimental data due to the difficulties of building integral test facilities and the high costs of these experiments, has accepted the offer of the Hungarian Academy of Sciences and organized two standard problem exercises. In these exercises, experimental data from the simulation of a 7.4% break loss of coolant accident was compared with analytical prediction of the behaviour of the facility calculated with computer codes. The second standard problem exercise involved a similar test, with the exception that in this case hydroaccumulator of the safety injection system were allowed to inject water in the system as anticipated in the design of the plant. This document presents a complete overview of the Second Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation. 22 refs, figs and tabs

  11. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs

  12. Description of steam-condensation phenomena during the loss-of-coolant accident

    International Nuclear Information System (INIS)

    McCauley, E.W.; Holman, G.S.; Aust, E.; Schwan, H.; Vollbrandt, J.; Fuerst, H.

    1980-01-01

    The development and verification of advanced computer models which describe the boiling water reactor (BWR) pressure suppression process for a hypothetical loss-of-coolant accident (LOCA) require a clear description of basic steam condensation phenomena. The GKSS Research Center, in coordination with interested institutions of West Germany and the United States, is currently conducting a test program for such basic research on a multivent BWR-related pressure suppression system. The Lawrence Livermore National Laboratory (LLNL) acts as the principal US NRC liaison for this test program, with particular emphasis on development of GKSS data for confirmatory use regarding US Mark II nuclear power plants as well as to advanced code development. The multivent test facility, placed in operation in February 1979, is a three-pipe full-scale vent system modelling main features of both the West German KWU and United States G.E. Mk II BWR pressure suppression systems. The test facility and testing programs are described

  13. Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Raza, S.H.

    1993-12-01

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs

  14. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kao, S.P.; Chang, S.K.; Huang, H.C. [Nuclear Training Branch, Northeast Utilities, Waterford, CT (United States)

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

  15. Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-01-01

    A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios

  16. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Khattab, M; Ibrahim, N A; Bedrose, C D [Reactors department, nuclear research center, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs.

  17. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

    1995-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

  18. Prediction of thermal hydraulic parameters in the loss of coolant accident by using artificial neural networks

    International Nuclear Information System (INIS)

    Vaziri, N.; Erfani, A.; Monsefi, M.; Hajabri, A.

    2008-01-01

    In a reactor accident like loss of coolant accident , one or more signals may not be monitored by control panel for some reasons such as interruptions and so on. Therefore a fast alternative method could guarantee the safe and reliable exploration of nuclear power planets. In this study, we used artificial neural network with Elman recurrent structure to predict six thermal hydraulic signals in a loss of coolant accident after upper plenum break. In the prediction procedure, a few previous samples are fed to the artificial neural network and the output value or next time step is estimated by the network output. The Elman recurrent network is trained with the data obtained from the benchmark simulation of loss of coolant accident in VVER. The results reveal that the predicted values follow the real trends well and artificial neural network can be used as a fast alternative prediction tool in loss of coolant accident

  19. Identification of Loss-of-Coolant Accidents in LWRs by Inverse Models

    International Nuclear Information System (INIS)

    Cholewa, Wojciech; Frid, Wiktor; Bednarski, Marcin

    2004-01-01

    This paper describes a novel diagnostic method based on inverse models that could be applied to identification of transients and accidents in nuclear power plants. In particular, it is shown that such models could be successfully applied to identification of loss-of-coolant accidents (LOCAs). This is demonstrated for LOCA scenarios for a boiling water reactor. Two classes of inverse models are discussed: local models valid only in a selected neighborhood of an unknown element in the data set, representing a state of a considered object, and global models, in the form of partially unilateral models, valid over the whole learning data set. An interesting and useful property of local inverse models is that they can be considered as example-based models, i.e., models that are spanned on particular sets of pattern data. It is concluded that the optimal diagnostic method should combine the advantages of both models, i.e., the high quality of results obtained from a local inverse model and the information about the confidence interval for the expected output provided by a partially unilateral model

  20. Loss of coolant accident mitigation for liquid metal cooled space reactors

    International Nuclear Information System (INIS)

    Georgevich, Vladimir; Best, Frederick; Erdman, Carl

    1989-01-01

    A loss of coolant accident (LOCA) in a liquid metal-cooled space reactor system has been considered as a possible accident scenario. Development of new concepts that will prevent core damage by LOCA caused elevated temperatures is the primary motivation of this work. Decay heat generated by the fission products in the reactor core following shutdown is sufficiently high to melt the fuel unless energy can be removed from the pins at a sufficiently rapid rate. There are two major reasons that prevent utilization of traditional emergency cooling methods. One is the absence of gravity and the other is the vacuum condition outside the reactor vessel. A concept that overcomes both problems is the Saturated Wick Evaporation Method (SWEM). This method involves placing wicking structures at specific locations in the core to act as energy sinks. One of its properties is the isothermal behaviour of the liquid in the wick. The absorption of energy by the surface at the isothermal temperature will direct the energy into an evaporation process and not in sensible heat addition. The use of this concept enables establishment of isothermal positions within the core. A computer code that evaluates the temperature distribution of the core has been developed and the results show that this design will prevent fuel meltdown. (author)

  1. Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

    International Nuclear Information System (INIS)

    Megahed, M.M.

    1992-01-01

    An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

  2. Confinement barriers for loss of coolant accidents in the SEAFP reactor plant models

    International Nuclear Information System (INIS)

    Blomquist, R.; Ebert, E.; Gay, J.M.; Mazille, F.; Natalizio, A.; Rolandsson, S.; Ross, W.E.; Shen, K.; Sjoeberg, A.

    1995-01-01

    Loss of coolant accidents may mobilise radioactivity and pressurise confinement barriers thereby making a release to the environment possible. The paper defines the radioactivity confinements and presents principal results from the underlying thermal-hydraulic analyses. (orig.)

  3. Generic evaluation of small break loss-of-coolant accident behavior in Babcock and Wilcox designed 177-FA operating plants

    International Nuclear Information System (INIS)

    1980-01-01

    Slow system depressurization resulting from small break loss-of-coolant accidents (LOCAs) in the reactor coolant system have not, until recently, received detailed analytical study comparable to that devoted to large breaks. Following the TMI-2 accident, the staff had a series of meetings with Babcock and Wilcox (B and W) and the B and W licensees. The staff requested that B and W and the licensees: (1) systematically evaluate plant response for small break loss-of-coolant accidents; (2) address each of the concerns documented in the Michelson report; (3) validate the computer codes used against the TMI-2 accident; (4) extend the break spectrum analysis to very small breaks, giving special consideration to failure of pressurizer valves to close; (5) analyze degraded conditions where AFW is not available; (6) prepare design changes aimed at reducing the probability of loss-of-coolant accidents produced by the failure of a PORV to close; and (7) develop revised emergency procedures for small breaks. This report describes the review of the generic analyses performed by B and W based on the requests stated above

  4. Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-01-01

    This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

  5. Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1994-07-01

    This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

  6. Containment parametric analysis for loss of coolant accident

    International Nuclear Information System (INIS)

    Fabjan, L.

    1985-01-01

    Full text: This paper presents parametric analysis of double containment response to LOCA using CONTEMPT-LT/28 code. The influence of the active and passive heat sinks on thermodynamic parameters in the containment after big and small LOCA was considered. (author)

  7. Nonstationary pressure build up in full-pressure containments after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1977-01-01

    The time histories of pressure, temperature and pressure difference during the pressure build up phase of a loss-of-coolant accident (LOCA) in the primary system in full-pressure containments of water cooled nuclear power reactors are treated. These are important for the design of such containments. The experiments within the German research program RS 50 ''Druckverteilung im Containment'' offered, for the first time, the opportunity to observe experimentally fluid-dynamic processes in a multiple divided full-pressure containment, and to test at the same time, computer codes which serve to describe the physical processes during the LOCA. The comparison of the results calculated by the computer codes ZOCO VI and DDIFF with the experimental results showed apparent deviations by special arrangements of the compartments and the vent flow paths of a model containment for the calculation of time dependent pressure-, temperature- and pressure difference-histories. The deviations lead to the development of the analytical model and computer code COFLOW. This new model was primarily designed to deal with the fluid-dynamic processes in the beginning phase of the blowdown as maximal pressure differences appear. Furthermore, it can be used to determine the maximum containment pressure, as well as for long term calculations. The analytical model and computer code COFLOW shows a better correlation between theory and experiment than previous codes

  8. Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Joo S.; Diamond, David

    2016-12-06

    A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in the analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.

  9. CNE (central nuclear en Embalse): probabilistic safety study. Loss-of-coolant accidents. Analysis through events sequence

    International Nuclear Information System (INIS)

    Layral, S.I.

    1987-01-01

    The aim of this study was to perform for the Embalse nuclear power plant, a probabilistic evaluation of loss-of-coolant accidents (LOCA) to identify the risks associated with them and to determine their acceptability in accordance with norms. This study includes all ruptures in the primary system that produce the automatic activation of 'emergency core cooling system'. Three starting events were selected for the probabilistic evaluation: 100% rupture of an input collector; 5% rupture of an input collector; 1.2% rupture of an input collector. At this stage the evaluation is focussed on the identification and quantization of the main failure sequences that follow a LOCA and lead to an uncontrolled reactor state or 'core meltdown'. The most important contribution to the core meltdown due to LOCA is the failure of supplies that are required for the emergency core cooling system. (Author)

  10. CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1978-04-01

    The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

  11. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  12. A review of Zircaloy fuel cladding behavior in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Leistikow, S.

    1985-09-01

    The paper reviews the state-of-the-art experimental work performed in several countries with respect to the acceptance criteria established for emergency core cooling (ECC) in a loss-of-coolant accident (LOGA) of light water reactors (LWRs). It covers in detail oxidation, embrittlement, plastic deformation and coolability of deformed rod bundles. The main test results are discussed on the basis of research work performed at the Karlsruhe Nuclear Research Center (KfK) within the framework of the Nuclear Safety Project (PNS) and reference is made to test data obtained in other countries. The conclusion reached in the paper is that the major mechanisms and consequences of oxidation, deformation and emergency core cooling are sufficiently investigated in order to provide a reliable data base for safety assessments and licensing of LWRs. All test data prove that the ECC-criteria are conservative and that the coolability of an LWR and the public safety can be maintained in a LOCA. (orig.) [de

  13. A study of the loss of coolant accident

    International Nuclear Information System (INIS)

    Lee, Y.W.; Chung, M.K.; Kim, S.H.; Park, J.S.; Lee, C.B.; Kim, S.B.; Won, S.Y.; Cho, Y.R.

    1983-01-01

    The primary objectives of this project are: (1) To review the published information on LOCA/ECCS study (2) To investigate reflood phenomena and to provide necessary information for analytical model development (3) To modyfy and develop a reflood analysis code. To review the published information on LOCA/ECCS, heat transfer phenomena are divided into 4 regions. Heat transfer correlations published in the references are reviewed and classified according to the regions. To investigate reflood phenomena and to provide better modeling of reflood phenomena, experments have been carried out with an electrically heated 3x3 rod bundle. Heat flux and heat transfer coefficients at the hot surface have been determined from the experimental data by HTC program. The influences of the parameters such as flooding rate, coolant subcooling and power generation on the propagation of rewetting front were also investigated. Calculations obtained from REFLUX code were compared with the experimental data to help an understanding of the reflood heat transfer mechanisms, and then some modifications of the code were provided. Improvements in heat transfer correlations of transition and inverted annular film boiling region, and the logic for the selection of heat transfer regime allowed better estimate for rod temperature behavior. (Author)

  14. CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

    International Nuclear Information System (INIS)

    Haste, T.J.

    1982-07-01

    The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

  15. Small break loss of coolant accidents: Bottom and side break

    International Nuclear Information System (INIS)

    Hardy, P.G.; Richter, H.J.

    1987-01-01

    A LOCA can be caused, e.g. by a small break in the primary cooling system. The rate of fluid escaping through such a break will define the time until the core will be uncovered. Therefore the prediction of fluid loss and pressure transient is of major importance to plan for timely action in response to such an event. Stratification of the two phases might be present upstream of the break, thus, the location of the break relative to the vapor-liquid interface and the overall upstream fluid conditions are relevant for the calculation of fluid loss. Experimental results and analyses are presented here for small breaks at the bottom or at the side of a small pressure vessel. It was found that in such a case the onset of the so-called ''vapor pull through'' is important but swelling at sufficient depressurization rates of the liquid due to flashing is also of significance. It was also discovered that in the bottom break the flow rate is strongly dependent on the break entrance quality of the vapour-liquid mixture. The side break can be treated similarly to the bottom break if the interface level is above the break. The analyses developed on the basis of experimental observations showed reasonable agreement of predicted and measured pressure transients. It was possible to calculate the changing interface level and mixture void fraction history in a way compatible with the behavior observed during the experiments. Even though the experiments were performed at low pressures, this work should help to get a better understanding of physical phenomena occurring in a full scale small break LOCA. (orig./HP)

  16. Environmental radiological consequences of a loss of coolant accident

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1981-01-01

    The elaboration of a calculation model to determine safety areas, named Exclusion Zone and Low Population Zone for nuclear power plants, is dealt with. These areas are determined from a radioactive doses calculation for the population living around the NPP after occurence of a postulated ' Maximum Credible Accident' (MCA). The MCA is defined as an accident with complete loss of primary coolant and consequent fusion of a substantial portion of the reactor core. In the calculations carried out, data from NPP Angra I were used and the assumptions made were conservative, to be compatible with licensing requirements. Under the most pessimistic assumption (no filters) the values of 410m and 1000m were obtained for the Exclusion Zone and Low Population Zone radii, respectivily. (Author) [pt

  17. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 – Aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Howe, Kerry J., E-mail: howe@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Mitchell, Lana, E-mail: lmitchell@alionscience.com [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kim, Seung-Jun, E-mail: skim@lanl.gov [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Blandford, Edward D., E-mail: edb@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kee, Ernest J., E-mail: erniekee@gmail.com [South Texas Project Nuclear Operating Company, P.O. Box 270, Wadsworth, TX 77483 (United States)

    2015-10-15

    Highlights: • Trisodium phosphate (TSP) causes aluminum corrosion to cease after 24 h of exposure. • Chloride, iron, and copper have a minimal effect on the rate of aluminum corrosion when TSP is present. • Zinc can reduce the rate of aluminum corrosion when TSP is present. • Aluminum occasionally precipitates at concentrations lower than the calculated solubility for Al(OH){sub 3}. • Corrosion and solubility equations can be used to calculate the solids generated during a LOCA. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum from metallic aluminum surfaces under conditions representative of the containment pool following a postulated loss of coolant accident at a nuclear power generating facility. The experiments showed that TSP is capable of passivating the aluminum surface and preventing continued corrosion after about 24 h at the conditions tested. A correlation that describes the rate of corrosion including the passivation effect was developed from the bench experiments and validated with a separate set of experiments from a different test system. The saturation concentration of aluminum was shown to be well described by the solubility of amorphous aluminum hydroxide for the majority of cases, but instances have been observed when aluminum precipitates at concentrations lower than the calculated aluminum hydroxide solubility. Based on the experimental data and previous literature, an equation was developed to calculate the saturation concentration of aluminum as a function of pH and temperature under conditions representative of a loss of coolant accident (LOCA) in a TSP-buffered pressurized water reactor (PWR) containment. The corrosion equation and precipitation equation can be used in concert with each other to calculate the quantity of solids that would form as a function of time during a LOCA if the temperature and pH profiles were known.

  18. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 – Aluminum

    International Nuclear Information System (INIS)

    Howe, Kerry J.; Mitchell, Lana; Kim, Seung-Jun; Blandford, Edward D.; Kee, Ernest J.

    2015-01-01

    Highlights: • Trisodium phosphate (TSP) causes aluminum corrosion to cease after 24 h of exposure. • Chloride, iron, and copper have a minimal effect on the rate of aluminum corrosion when TSP is present. • Zinc can reduce the rate of aluminum corrosion when TSP is present. • Aluminum occasionally precipitates at concentrations lower than the calculated solubility for Al(OH) 3 . • Corrosion and solubility equations can be used to calculate the solids generated during a LOCA. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum from metallic aluminum surfaces under conditions representative of the containment pool following a postulated loss of coolant accident at a nuclear power generating facility. The experiments showed that TSP is capable of passivating the aluminum surface and preventing continued corrosion after about 24 h at the conditions tested. A correlation that describes the rate of corrosion including the passivation effect was developed from the bench experiments and validated with a separate set of experiments from a different test system. The saturation concentration of aluminum was shown to be well described by the solubility of amorphous aluminum hydroxide for the majority of cases, but instances have been observed when aluminum precipitates at concentrations lower than the calculated aluminum hydroxide solubility. Based on the experimental data and previous literature, an equation was developed to calculate the saturation concentration of aluminum as a function of pH and temperature under conditions representative of a loss of coolant accident (LOCA) in a TSP-buffered pressurized water reactor (PWR) containment. The corrosion equation and precipitation equation can be used in concert with each other to calculate the quantity of solids that would form as a function of time during a LOCA if the temperature and pH profiles were known

  19. Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Carl, H.; Kubis, K.

    1979-01-01

    Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

  20. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  1. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  2. On the air coolability of TRIGA reactors following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    El-Genk, Mohamed S.; Kim, Sung-Ho; Zaki, Galal M.; Foushee, Fabian; Philbin, Jeffrey S.; Schulze, James

    1986-01-01

    This paper describes the experiments on the air-coolability of a heated rod in a vertical open annulus at near atmospheric pressure. This data can be applied to the coolability of reactor fuel rods that are totally uncovered in a Loss-of-Coolant Accident (LOCA). As a prelude to measuring air coolability of specific core geometries (bundles), heat transfer data was collected for natural convection of atmospheric air in open vertical annuli with an isoflux inner wall and an insulated outer wall (diameter ratios, annulus ratio, of 1.155, 1.33, 1.63, and 12). Although the inner heated tube had the same overall dimensions as the fuel rod in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (3.81 cm o.d. and 55.5 cm long), the heated length was only 36.0 cm rather than the entire 50.5 cm for the ACRR's rods. The test assembly was operated at heat fluxes up to 1.38 W/cm 2 with a corresponding surface temperature of 852 K. The annulus data was extrapolated to an equilibrium surface temperature of 1200 K (as a coolability limit of TRIGA reactors) to provide a qualitative estimate of the coolability of multirod bundles by free convection of atmospheric air. The results suggest that for a typical pitch-to-diameter ratio of 1.12 in the ACRR the decay heat removal level is about 1.0 kW/m. This corresponds to an initial decay power following sustained operations at about 12.5 kW/m in the ACRR. However, because of the uncertainties in duplicating the actual thermal-hydraulic conditions in a multirod bundle using a single rod annulus, the actual coolability of open pool reactors could be different from those suggested in this paper. (author)

  3. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    International Nuclear Information System (INIS)

    Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

    2014-01-01

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

  4. Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

  5. The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    International Nuclear Information System (INIS)

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

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

  6. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

  7. MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.

    1982-04-01

    MABEL-2 has been developed to predict the extent of cladding deformation in PWR fuel rods during a loss of coolant accident. The user notes describe how to run MABEL. They include case preparation and input data, the job control language, a description of the output tables available, and aids to debugging. The input data and results for two sample cases are given. (U.K.)

  8. Description of steam condensation phenomena during the loss-of-coolant accident

    International Nuclear Information System (INIS)

    McCauley, E.W.; Holman, G.S.; Aust, E.; Furst, H.; Schwan, H.; Vollbrandt, J.

    1981-01-01

    Study of results from the full scale multivent pressure suppression experiment conducted by the GKSS Laboratory has developed an improved understanding of the dynamic, oscillatory steam condensation events and related loading functions which occur during the hypothetical loss-of-coolant accident in a boiling water nuclear reactor. Due to the unique measurements systems which combines both cinematic and digital data, qualified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena has been obtained

  9. Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Suman, Siddharth [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Khan, Mohd Kaleem, E-mail: mkkhan@iitp.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Pathak, Manabendra [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Singh, R.N.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2016-10-15

    Highlights: • Modelling of nuclear fuel cladding during loss-of-coolant accident transient. • Phase transformation, corrosion, and creep combined to evaluate burst criterion. • Effect of oxygen concentration on burst stress and burst strain. • Effect of heating rate, internal pressure fluctuation, shear modulus incorporated. - Abstract: A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

  10. Study and development of a pyrometric in-core measurement technique to follow the temperature of the fuel rod cladding; applied to the study of Loss of Coolant Accident (LOCA) during trial simulations in the Jules Horowitz Reactor (Material Testing Reactor)

    International Nuclear Information System (INIS)

    Ramiandrisoa, Liana

    2014-01-01

    In both research and industry, temperature is a key parameter for understanding and characterizing the behavior of materials. To study the thermomechanical behavior of a fuel rod, a test device is designed for the Jules Horowitz Material Testing Reactor (currently under construction in the CEA Cadarache). The device will be placed under accidental conditions (Loss Of Coolant Accident, LOCA) causing rapid overheating. The temperature tracking, between 700 and 1200 C, will be measured by a fiber optic sensor. The aim of the project is to optimize temperature measurement by comparing different pyrometry techniques. This study covers the management of the main difficulties inherent to the design of the sensor.The first challenge consists of predicting optical fiber behavior in such complex environments where irradiation and high temperature are combined. The fiber will be exposed to a neutron dose rate about 10 12 nfast/cm 2 /s and a dose rate of about 1 kGy/s. Moreover its extremity is heated to approximately 800 C. It is shown that under these conditions, light interferences, absorption bands and fluctuating attenuation are obstacles to overcome or to mitigate.The second challenge, concerning pyrometric measurement, comes from spectral variations expected for the rod emissivity. The material of study is chosen for its widespread use in France: Zircaloy-4. Under oxidizing conditions the spectral emissivity of this Zirconium alloy evolves. This thesis proves that between 700 and 800 C pyrometric measurement is possible from experimental point of view in laboratory without irradiation.In conclusion rod temperature tracking in JHR conditions may be possible providing that interferences are mastered and wavelengths are chosen. This work makes the use of optical pyrometry under civil nuclear extreme conditions more promising. (author) [fr

  11. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    International Nuclear Information System (INIS)

    Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-01-01

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  12. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-04-15

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  13. Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

    International Nuclear Information System (INIS)

    Albrecht, H.; Malinauskas, A.P.

    1978-01-01

    A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

  14. Radiological impact of a loss of coolant accident at Angra 2 reactor

    International Nuclear Information System (INIS)

    Dias, W.

    1992-01-01

    A loss of coolant accident is analyzed which comprises a double ended rupture of a main primary system line. The accident sequence is described and the main assumptions as to the activity release are presented. On the basis of site specific meteorological data, the atmospheric dispersion factors are calculated using the Gaussian plume diffusion model and the doses are then determined at the boundary of the low population zone. The resulting values for the effective dose equivalent are more than one order of magnitude below that due to the average background radiation received in one year. (author)

  15. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  16. Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

    2010-01-01

    For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

  17. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in GE-designed operating plants and near-term operating license applications

    International Nuclear Information System (INIS)

    1980-01-01

    The results are presented of a generic evaluation of feedwater transients, small-break loss-of-coolant accidents (LOCAs), and other TMI-2-related events for General Electric Company (GE)-designed operating plants and near-term operating license applications to confirm or establish the bases for the continued safe operation of the operating plants. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas. Additional review of the accident is continuing and further information is being obtained and evaluated. Any new information will be reviewed and modifications will be made as appropriate

  18. Analysis of fuel behaviour after loss-of-coolant accident with the TESPA-code

    International Nuclear Information System (INIS)

    Keusenhoff, J.

    1981-01-01

    After a loss-of-coolant accident fuel rods go through a phase of high temperature and differential pressure before quenching and initiation of long term cooling. For licensing purpose the highest cladding temperature and the coolability of the core is of interest. The highest temperature is evaluated by a hot channel calculation with conservative assumptions. It gives little information about the status of the entire core. Therefore more detailed information is necessary. TESPA is a fast running code, which uses best-estimate assumptions, considers statistical uncertainties in the input parameters and calculates clad ballooning and rupture. The code is a usefull tool for calculation of channel blockage and cladding rupture

  19. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  20. Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kanzleiter, T.F.

    1976-01-01

    For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1 : 64 compared to the 1200 MW PWR plant Biblis A. (Auth.)

  1. Microstructural examination of fuel rods subjected to a simulated large-break loss of coolant accident in reactor

    International Nuclear Information System (INIS)

    Garlick, A.

    1985-01-01

    A series of tests has been conducted in the National Research Universal (NRU) reactor, Chalk River, Canada, to investigate the behaviour of full-length 32-rod PWR fuel bundles during a simulated large-break loss of coolant accident (LOCA). In one of these tests (MT-3), 12 central rods were pre-pressurized in order to evaluate the ballooning and rupture of cladding in the Zircaloy high-α/α+β temperature region. All 12 rods ruptured after experiencing < 90% diametral strain but there was no suggestion of coplanar blockage. Post-irradiation examination was carried out on cross-sections of cladding from selected rods to determine the aximuthal distribution of wall thinning along the ballooned regions. These data are assessed to check whether they are consistent with a mechanism in which fuel stack eccentricity generates temperature gradients around the ballooning cladding and leads to premature rupture during a LOCA. After anodizing, the cladding microstructures were examined for the presence of prior-beta phase that would indicate the α/α+β transformation temperature (1078K) had been exceeded. These results were compared with isothermal annealing test data on unirradiated cladding from the same manufacturing batch

  2. Theoretical study on loss of coolant accident of a research reactor

    International Nuclear Information System (INIS)

    Lee, Kwon-Yeong; Kim, Wan-Soo

    2016-01-01

    Highlights: • A theoretical model of siphon breaking phenomena was developed. • A general formula using Chisholm coefficient B was proposed. • The safety requirements regarding a loss of coolant accident of research reactors could be found out. - Abstract: Under the design conditions of a research reactor, the siphon phenomenon induced by pipe rupture can cause continuous efflux of water. In order to prevent water efflux, an additional facility is necessary. A siphon breaker is a type of safety facility that can resist the loss of coolant effectively. However, analysis of siphon breaking is complex since it comprises two-phase flow and there are many inputs to be considered. For this reason, we analyzed the experimental results to develop a theoretical model of siphon breaking phenomena. Developed model is based on fluid mechanics and Chisholm model. From Bernoulli’s equation, the velocity and quantity as well as undershooting height, water level, pressure, friction coefficient, and factors related to the two-phase flow could be calculated. The Chisholm model, which is able to analyze the two-phase flow, can predict the results in a manner similar to those obtained from a real-scale experiment, and a general formula using Chisholm coefficient B was proposed in this study. Also, we verified the theoretical model and concluded that it is possible to analyze the siphon breaking. Moreover, the design conditions that can satisfy the safety requirements regarding a loss of coolant accident of research reactors could be found out by using the theoretical model. In conclusion, we propose the theoretical model which can analyze the siphon breaking as real, and it is helpful not only to analyze but also to design the siphon breaker.

  3. The condensation of steam on the external surfaces of the shells of HIFAR heavy water heat exchangers during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Chapman, A.G.

    1987-03-01

    A study of steam condensation rates on the HIFAR heavy water heat exchangers was undertaken to predict thermohydraulic conditions in the HIFAR containment during a postulated loss-of-coolant accident (LOCA). The process of surface condensation from a mixture of air and steam, and methods for calculating the rate of condensation, are briefly reviewed. Suitable experimental data are used to estimate coefficients of condensation heat transfer to cool surfaces in a reactor containment during a LOCA. The relevance of the available data to a LOCA in the HIFAR materials testing reactor is examined, and two sets of data are compared. The differences between air/H 2 O and air/D 2 O mixtures are discussed. Formulae are derived for the estimation of the coefficient of heat transfer from the heat exchanger shells to the cooling water, and a method of calculating the rate of condensation per unit area of surface is developed

  4. Loss of Coolant Accident Simulation for the Top-Slot break at Cold Leg Focusing on the Loop Seal Reformation under Long Term Cooling with the ATLAS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Rok; Park, Yu Sun; Bae, Byoung Uhn; Choi, Nam Hyun; Kang, Kyoung Ho; Choi, Ki Yong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In the present paper, loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS, which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors (PWRs) of an advanced power reactor of 1400 MWe (APR1400). The simulation was focused on the loop seal reformation under long term cooling condition. During a certain class of Loss of Coolant Accident (LOCA) in a PWR like an advanced power reactor of 1400 MWe (APR1400), the steam volume in the reactor vessel upper plenum and/or upper head may continue expanding until steam blows liquid out of the intermediate leg (U-shaped pump suction cold leg), called loop seal clearing (LSC), opening a path for the steam to be relieved from the break. Prediction of the LSC phenomena is difficult because they are varies for many parameters, which are break location, type, size, etc. This LSC is the major factor that affects the coolant inventory in the small break LOCA (SBLOCA) or intermediate break LOCA (IBLOCA). There is an issue about the loop seal reformation that liquid refills intermediate leg and blocks the steam path after LSC. During the SBLOCA or IBLOCA, the Emergency Core Cooling System (ECCS) is operated. For long term of the top slot small or intermediate break at cold leg, the primary steam condensation by SG heat transfer or SIP, SIT water flooding (reverse flow to loop seal) make loop seal reformation possibly. The primary pressure increase at the top core region due to the steam release blockage by loop seal reformation. And then core level decreases and partial core uncover may occur. The loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS. The loop seal clearing and loop seal reformation were occurred repeatedly.

  5. Fission product source from Ignalina NPP in case of loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Ubonavicius, E.; Rimkevicius, S.

    2001-01-01

    The release of radioactive materials to the environment is of special importance in the case of any accident at Nuclear Power Plants (NPP). The integrated analysis of thermal-hydraulic parameters behavior and radioactive fission products (FP) transport and deposition in the compartments play an important role in the evaluation of FP release to the environment and determines the irradiation dozes of personnel and public. In this report the transport and the deposition of radioactive material in the Ignalina NPP unit 1 compartments as well as the FP source term to the environment in the case of design basis loss-of-coolant accidents are discussed. The calculation models for the evaluation of FP transport and deposition as well as the results of performed calculations of several accidents at Ignalina NPP are presented. (author)

  6. Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

    International Nuclear Information System (INIS)

    Luo Bangqi

    2012-01-01

    The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

  7. Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

    International Nuclear Information System (INIS)

    Rivas, J.C.; Dies, J.; Fajarnés, X.

    2015-01-01

    Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

  8. Touch-sensitive colour graphics enhance monitoring of loss-of-coolant accident tests

    International Nuclear Information System (INIS)

    Snedden, M.D.; Mead, G.L.

    1982-01-01

    A stand-alone computer-based system with an intelligent colour termimal is described for monitoring parameters during loss-of-coolant accident tests. Colour graphic displays and touch-sensitive control have been combined for effective operator interaction. Data collected by the host MODCOMP II minicomputer are dynamically updated on colour pictures generated by the terminal. Experimenters select system functions by touching simulated switches on a transparent touch-sensitive overlay, mounted directly over the face of the colour screen, eliminating the need for a keyboard. Switch labels and colours are changed on the screen by the terminal software as different functions are selected. Interaction is self-prompting and can be learned quickly. System operation for a complete set of 20 tests has demonstrated the convenience of interactive touchsensitive colour graphics

  9. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided

  10. Interfacing systems loss of coolant accident (ISLOCA) pressure capacity methodology and Davis-Besse results

    International Nuclear Information System (INIS)

    Wesley, D.A.

    1991-01-01

    A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA

  11. Frontier between medium and large break loss of coolant accidents of pressurized water reactor

    Science.gov (United States)

    Kim, Taewan

    2017-10-01

    In order to provide the probabilistic safety assessment with more realistic condition to calculate the frequency of the initiating event, a study on the frontier between medium-break and large-break loss-of-coolant-accidents has been performed by using best-estimate thermal hydraulic code, TRACE. A methodology based on the combination of the essential safety features and system parameter has been applied to the Zion nuclear power plant to evaluate the validity of the frontier utilized for the probabilistic safety assessment. The peak cladding temperature has been chosen as a relevant system parameter that represents the system behavior during the transient. The results showed that the frontier should be extended from 6 in. to 10 in. based on the required safety functions and system response.

  12. Deformation of PWR cladding following a loss-of-coolant accident

    International Nuclear Information System (INIS)

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

    1979-07-01

    A review is presented of recent experiments to investigate the deformation behaviour of Zircaloy cladding in simulated loss-of-coolant accidents. The behaviour of Zircaloy cladding is shown to be controlled by a complex interaction of metallurgical and heat transfer variables, with the latter having a major influence. There is a significant increase in both diametral strain and the axial extent of deformation in multi-rod compared with single-rod tests. The extent to which this will occur in nuclear-heated tests is not yet known; however, it is expected that the 'smearing' of the gamma-radiation portion of decay heat in such tests will tend to reduce circumferential temperature variations. Opposing this is the influence of the colder control rods in an assembly. The resolution of this dichotomy will require a series of in-reactor multi-rod tests and attendant code development. (author)

  13. MABEL-1. A code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.

    1978-06-01

    The MABEL-1 code has been written to investigate the deformation, of fuel pin cladding and its effects on fuel pin temperature transients during a loss-of-coolant accident. The code considers a single fuel pin with heated fuel concentric within the cladding. The fuel pin temperature distribution is evaluated using a one-dimensional conduction model with heat transfer to the coolant represented by an input set of heat transfer coefficients. The cladding deformation is calculated using the code CANSWEL, which assumes all strain to be elastic or creep and models the creep under a multi-axial stress system by a spring/dashpot combination undergoing alternate relaxation and elastic strain. (author)

  14. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 2 – Zinc

    International Nuclear Information System (INIS)

    Pease, David; LaBrier, Daniel; Ali, Amir; Blandford, Edward D.; Howe, Kerry J.

    2016-01-01

    Highlights: • Zinc release is limited to less than 1 mg/L in TSP-buffered solution under a variety of conditions (pH, temperature, zinc source). • Zinc release in high-temperature non-TSP-buffered environment is approximately 25 mg/L. • Long-term zinc release is controlled by passivation (without TSP) and zinc solubility (with TSP). • Precipitation and solubility of zinc phosphate limit the release of zinc. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of zinc from metallic zinc-bearing surfaces under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at a nuclear power generating facility. The experiments showed that in non-buffered (acidic) environments, measurable quantities of zinc are released from zinc-bearing surfaces. Precipitation and solubility of phosphate-based corrosion products, such as zinc phosphate, limit the release of zinc from zinc-bearing surfaces. These experiments have found that under a variety of conditions, including variations of temperature, pH, and across different zinc-bearing surfaces, the release of zinc into solution is limited to <1 mg/L when phosphate is present. When phosphate is not present, zinc release is instead bounded by a markedly higher saturation limit which is a strong function of the solution temperature.

  15. Analysis of a simulated small break in the semiscale system under loss-of-coolant accident conditions

    International Nuclear Information System (INIS)

    Cartmill, C.E.

    1978-01-01

    The Semiscale Mod-1 experimental program conducted by EG and G Idaho, Inc., is part of the overall U.S. Nuclear Regulatory Commission (NRC) and Department of Energy (DOE) sponsored research and development program to investigate the behavior of the pressurized water reactor (PWR) system during an hypothesized loss-of-coolant accident (LOCA). The Semiscale Mod-1 program is intended to provide transient thermal-hydraulic data from a simulated LOCA using a small-scale experimental nonnuclear system. The Semiscale Mod-1 program is a major contributor of experimental data that provide a means of evaluating the adequacy of overall system analytical models as well as the models of the individual system components. Selected experimental data produced by this program will also be used to aid other DOE and NRC sponsored experimental programs, such as the Loss-of-Fluid Test (LOFT) program in optimizing test series, selecting test parameters, and evaluating test results. The Semiscale Mod-1 tests are performed with an experimental system which simulates the principal features of a nuclear plant but which is smaller in volume. Nuclear heating is simulated in the tests by a core composed of an array of electrically heated rods. The core is contained in a pressure vessel which also includes a downcomer, lower plenum, and upper plenum. The Semiscale system piping is arranged such that the intact loop represents three loops of a four-loop nuclear plant, and the broken loop represents the fourth loop. In the present configuration the intact loop contains an active steam generator and pump, and the broken loop contains passive simulators for the steam generator and pump

  16. An on-line pressurizer surveillance system design to prevent small-break loss-of-coolant accidents through power-operated relief valves using a microcomputer

    International Nuclear Information System (INIS)

    Lee, J.H.; Chang, S.H.

    1987-01-01

    A small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve is one of the important contributors to nuclear power plant risk. A pressurizer surveillance system was designed to use a microcomputer to prevent the malfunction of the system; the effect of this improvement has been assessed through probabilistic risk assessment. The microcomputer diagnoses the malfunction of the system by a process-checking method and automatically performs the backup action related to each malfunction. This improvement means that we can correctly diagnose ''spurious opening,'' ''failure to reclose,'' and ''small-break LOCA,'' which are difficult for operators to diagnose quickly and correctly, and by taking automatic backup action one can reduce the probability of human error

  17. Analysis of unmitigated large break loss of coolant accidents using MELCOR code

    Science.gov (United States)

    Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

    2017-11-01

    In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

  18. Conservatism of loss-of-coolant accident licensing analysis compared to experimental results and best-estimate calculation

    International Nuclear Information System (INIS)

    Winkler, F.; Friedmann, P.

    1986-01-01

    The paper compares results of loss-of-coolant accident licensing analysis with experimental results and results of best-estimate calculations. The large safety margins resulting from the more realistic best-estimate results are used to show the high conservatism inherent in the licensing process of pressurized water reactors. (orig.) [de

  19. Fuel rod thermal analysis of the Angra-1 reactor during a postulated loss of coolant accident

    International Nuclear Information System (INIS)

    Praes, J.G.L.

    1982-01-01

    A thermal analysis of a fuel element is performed, as subject to the most severe cooling conditions, such as those occurring during a postulated Loss of Coolant Accident in the Angra-I reactor. Our objective was to ascertain whether the cooling of the core is assured according to 10 CRF - 50. According to the stated purpose, sensitivity analyses are necessary, using the swelling and rupture models of the cladding, and at the same time, an updating of the FLECHT heat transfer correlations in the computing program used, which is TOODEE-2 e 1 Version(28), with the purpose of adequating it to the Angra-I core analysis. In addition, we did sensitivity studies on heat transfer coefficient calculations for the steam cooling model. From the results obtained we conclude that the maximum temperature values of the cladding and the oxidation rate due to the Z sub(r) H 2 O reaction were kept well below the maximum allowable limits. Thus, the cooling of the Angra-I core is assured for the assumed accident. (Author) [pt

  20. Analysis of an AP600 intermediate-size loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

    1995-09-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  1. International Standard Problems and Small Break Loss-Of-Coolant Accident (SBLOCA)

    International Nuclear Information System (INIS)

    Aksan, N.

    2008-01-01

    Best-estimate thermalhydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. In this respect, parallel to other national and international programs, OECD/Nea (OECD Nuclear Energy Agency) Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years some forty-eight International Standard Problems (ISPs). These ISPs were performed in different fields as in-vessel thermalhydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermalhydraulic behaviour. 80% of these ISPs were related to the working domain of Principal Working Group no. 2 on Coolant System Behaviour (PWG2). The ISPs have been one of the major PWG2 activities for many years. The individual ISP comparison reports include the analysis and conclusions of the specific ISP exercises. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISP's is given in this paper based on a report prepared by a CSNI-PWG2 writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident, specifically small break LOCA, are shortly summarized. Five small break LOCA related ISP's are considered, since these were used for the assessment of the advanced best-estimate codes. The considered ISP's deal with the phenomenon typical of small break LOCAs in Western design PWRs. The experiments in four integral test facilities, LOBI, SPES, BETHSY

  2. Simulation of the SPE-4 small-break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Cebull, P.; Hassan, Y.A.

    1993-01-01

    A small-break loss of coolant accident (SBLOCA) conducted at the PMK-2 integral test facility was analyzed using RELAP5/MOD3. 1. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). The VVER design differs from pressurized water reactors (PWRS) of western origin, primarily in its use of horizontal steam generators, hot- and cold-leg loop seals, and safety injection tanks. Because of these differences, it will exhibit somewhat different transient behavior than most PWRS. The PMK-2 test facility, located at the KFKI Atomic Energy Research Institute (AEKI), is a scale model of the Paks nuclear power plant in Hungary with scaling factors of 1:2070 in power and volume and 1:1 in elevation. Primarily used to study SBLOCAs and natural circulation behavior of VVER reactors, it has been used in three previous SPEs

  3. An investigation of core liquid level depression in small break loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Schultz, R.R.; Watkins, J.C.; Motley, F.E.; Stumpf, H.; Chen, Y.S.

    1991-08-01

    Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs

  4. Reactor hydrodynamics during the reflood phase of a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gay, R.R.

    1977-01-01

    The thermohydraulics of a nuclear reactor during the reflood phase of a hypothetical loss-of-coolant accident can be represented by moving control volume methodology in which six control volumes are used to represent the downcomer, lower plenum, and reactor core. The one-dimensional, homogeneous, equilibrium constitutive equations for two-phase steam/water flow are solved in each control volume and connecting junctions. One of the three core control volumes represents the quench region; it changes size and position based on the axial location of the clad quench temperature and the condensed liquid level in the flow channel. The lengths of the remaining two core control volumes are determined by the position of the quench region. Simulation of actual reflood experiments demonstrates that the methodology predicts reflood-like flow oscillations and reproduces the correct trends in experimental data. The moving control volume methodology has proven itself as a valid concept for reflood hydrodynamics, but further development of the existing EFLOD code is required for simulation of actual reflood experiments

  5. Prediction of loop seal formation and clearing during small break loss of coolant accident

    International Nuclear Information System (INIS)

    Lee, Suk Ho; Kim, Hho Jung

    1992-01-01

    Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD2 and /MOD3 codes with the test of SB-CL-18 of the LSTF(Large Scale Test Facility). The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery includeing the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/ MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in the base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs with the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing. (Author)

  6. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    International Nuclear Information System (INIS)

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

  7. Contempt-LT: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Wheat, L.L.; Wagner, R.J.; Niederauer, G.F.; Obenchain, C.F.

    1975-06-01

    CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided to describe fan cooler and cooling spray engineered safety systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the reactor system may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. CONTEMPT-LT can be used to model all current boiling water reactor pressure suppression systems, including containments with either vertical or horizontal vent systems. CONTEMPT-LT can also be used to model pressurized water reactor dry containments, subatmospheric containments, and dual volume containments with an annulus region, and can be used to describe containment responses in experimental containment systems. The program user defines which compartments are used, specifies input mass and energy additions, defines heat structure and leakage systems, and describes the time advancement and output control. CONTEMPT-LT source decks are available in double precision extended-binary-coded-decimal-interchange-code (EBCDIC) versions. Sample problems have been run on the IBM360/75 computer. (U.S.)

  8. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  9. Simulation of a loss of coolant accident with rupture in the steam generator hot collector

    International Nuclear Information System (INIS)

    1991-03-01

    The Central Research Institute for Physics of the Hungarian Academy of Sciences designed and constructed the PMK-NVH test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary made the PMK-NVH facility available to the IAEA. The IAEA, having identified the need for experimental data due to the difficulties of building integral test facilities and the high costs of these experiments, has accepted the offer of the Hungarian Academy of Sciences and has organized three standard problem exercises. In these exercises, experimental data from the simulation of loss of coolant accidents were compared with analytical predictions of the behaviour of the facility, calculated with computer codes. The third standard problem exercise involved a test, in which the rupture was simulated to occur at the top of the hot collector of the steam generator, therefore creating a leak from primary to secondary side. Both hydroaccumulators and high pressure injection were allowed to actuate as prescribed in the actual plant. Eighteen organizations from 15 Member States took part in the exercise presenting pre-test and some post-test analyses which were discussed in a final meeting in Vienna in August, 1990. This document presents a complete overview of the third standard problem exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many interrelated steps; therefore, no general conclusion or optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation. 42 refs, figs and tabs

  10. Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

    International Nuclear Information System (INIS)

    Soares, Alexandre de Souza

    2014-01-01

    The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

  11. Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Tasaka, Kanji

    1986-01-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

  12. CONTEMPT-LT/028: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hargroves, D.W.; Metcalfe, L.J.; Wheat, L.L.; Niederauer, G.F.; Obenchain, C.F.

    1979-03-01

    CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided to describe fan cooler and cooling spray engineered safety systems. An annular fan model is also provided to model pressure control in the annular region of dual containment systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the reactor system may have both a liquid pool region and an air--vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different

  13. Study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS using RELAP5 code

    International Nuclear Information System (INIS)

    Hoang Minh Giang; Ha Thi Anh Dao; Hoang Tan Hung; Bui Thi Hoa; Nguyen Thi Tu Oanh; Dinh Anh Tuan; Pham Tuan Nam

    2017-01-01

    The advanced VVER-1200/V491 reactor designed with passive safety systems to deal with design extension conditions is primarily selected as priority candidate for Ninh Thuan 1 nuclear power plant project. So that, in order to enhance competence of nuclear safety and toward participation on review Safety Analysis Report (SAR) of Ninh Thuan nuclear Power project the study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS is implemented. As requirement of the study, the input deck file of VVER-1200/V491 for RELAP5 and analysis report for some special case of LOCAs along with partly failure of ECCS are issued. (author)

  14. Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

    The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

  15. A study of entrainment at a break and in the core during small break loss-of-coolant accidents in PWRs

    International Nuclear Information System (INIS)

    Yonomoto, Taisuke

    1996-05-01

    Objectives of the present study are to obtain a better understanding of entrainment at a break and in the core during small break loss-of-coolant-accidents (SBLOCAs) in PWRs, and to develop a means for the best evaluation of the phenomena. For the study of entrainment at a break, a theoretical model was developed, which was assessed by comparisons with several experimental data bases. By modifying a LOCA analysis code using the present model, experimental results obtained from SBLOCA experiments at a PWR large-scale simulator were reproduced very well. For the study of entrainment in the core, reflooding experiments were conducted at high pressure, from which the onset conditions were obtained. It was confirmed that the cooling behavior for a dry-out core is very simple under typical high pressure reflooding conditions for PWRs, because liquid entrainment does not occur in the core. (author)

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

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

  17. Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

    International Nuclear Information System (INIS)

    Pointner, W.; Broecker, A.

    2012-01-01

    The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

  18. Predicted HIFAR fuel element temperatures for postulated loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Green, W.J.

    1987-04-01

    A two-dimensional theoretical heat transfer model of a HIFAR Mark IV/Va fuel element has been developed and validated by comparing predicted thermal performances with experimental temperature responses obtained from irradiated fuel elements during simulated accident conditions. Full details of the model's development and its verification have been reported elsewhere. In this report, the model has been further used to ascertain acceptable limits of fuel element decay power for the start of two specific LOCAs which have been identified by the Regulatory Bureau of the AAEC. For a single fuel element which is positioned within a fuel load/unload flask and is not subjected to any forced convective air cooling, the model indicates that fission product decay powers must not exceed 1.86 kW if fuel surface temperatures are not to exceed 450 deg C. In the case of a HIFAR core LOCA in which the complete inventory of heavy water is lost, it is calculated that the maximum fission product decay power of a central element must not exceed 1.1 kW if fuel surface temperatures are not to exceed 450 deg C anywhere in the core

  19. The once-through mode of steam generator reflux condensation in loss of coolant accident scenarios

    International Nuclear Information System (INIS)

    Liao, Y.; Guentay, S.; Suckow, D.

    2009-01-01

    The once-through mode of steam generator reflux condensation in the presence of noncondensable gases and/or aerosols for LOCA scenarios is introduced. This phenomenon is planned to be investigated at Paul Scherrer Institute in the ARTIST/RFLX experimental program. The plausible accident scenarios associated with the once-through reflux condensation are analyzed with MELCOR to study the safety significance and the boundary conditions of this phenomenon. This work presents the recent PSI experimental and analytical work on reflux condensation: the progress of modification to the ARTIST test facility for the purpose to study reflux condensation, and the analytical model for the once-through reflux condensation in the presence of noncondensable gas using the heat and mass transfer analogy approach. Future experimental and analytical work on reflux condensation is also outlined. (author)

  20. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  1. Studies of loss-of-coolant and loss-of-regulation accidents

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1979-10-01

    Studies of a CANDU reactor during loss of coolant with delayed emergency core cooling showed that the moderator is an effective heat sink, and that in reactors with moderator dump the calandria sprays provide effective cooling. Fuel channel melting would not occur, and a coolable geometry will be maintained. Studies on film cooling and film stability on calandria tubes and on the analysis of flow reversal in vertical feeder tubes are also reported

  2. Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.

    1978-12-01

    The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author) [pt

  3. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    International Nuclear Information System (INIS)

    Kalkahoran, Omid Noori; Ahangari, Rohollah; Shirani, Amir Saied

    2016-01-01

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results

  4. Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

    Directory of Open Access Journals (Sweden)

    Omid Noori-Kalkhoran

    2016-10-01

    Full Text Available Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model. In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

  5. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Kalkahoran, Omid Noori; Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

  6. A study of the large break loss-of-coolant accident in the Angra-1 nuclear power plant

    International Nuclear Information System (INIS)

    Borges, E.M.

    1984-01-01

    The simulation of the Angra-I nuclear power plant under the condition of large break loss of coolant accident is presented, the thermal-hydraulic analysis of the primary circuit during each phase of the acident and thermal analysis of the hottest fuel rod curing reflooding are shown. Computer codes RELAP4/MOD5 (options EM and FLOOD) and TOODEE 2 are used to perform these computations. Fuel rod peak temperatures reached during the simulation are below the permissible levels. However, during the reflooding phase; the maximum oxidation of the cladding exceeds the limit of 0.17 times the original cladding thickness. (Author) [pt

  7. MABEL-2D: a code to analyse cladding deformation in a loss-of-coolant accident. Part 2

    International Nuclear Information System (INIS)

    Bowring, R.W.

    1985-08-01

    The MABEL series of codes is being developed at Harwell to predict the extent of cladding deformation (ballooning) in pressurized water reactor fuel rods during a loss of coolant accident. MABEL - 2D is an updated version of MABEL - 2C. These are user notes for MABEL - 2D (which is described in a separate report AEEW - R1979). They describe the input data specification; the use of the restart facility; debug printing and quick-running sample problems. The input data are divided into rod data, thermal hydraulic data and creep data. There is an input data flow chart. The main appendix gives the detailed input data specification. (U.K.)

  8. Analysis of the behaviour of pressure and temperature of the containment of a PWR reactor, submitted to a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Silva, D.E. da; Arrieta, L.A.J.; Costa, J.R.; Camargo, C.; Santos, C.M. dos; Rochedo, E.R.R.

    1979-12-01

    The main purpose of this work is to analyse the pressure and temperature behaviour of the metalic containment of a PWR building, submitted to a postulated loss-of-coolant accident (LOCA) caused by a double-ended rupture in the main line of the primary circuit. The scope of the study was directed to verify the Final Safety Analysis Report (FSAR) results for the integrity of the metalic containment of the Angra I power plant. The highest containment pressure peak for this unit is expected for a break in the suction line of one of the main pumps of the primary coolant. Using the same input data, our results are very similar to those presented in the FSAR which shows a reasonable equivalence between the two analytical models. Using as input data the results of a previous LOCA study at CNEN, which yields to more conservative boundary conditions than those presented by the FSAR, the pressure and temperature peak values determined by our model are quite larger than those presented by the cited Safety Report. (author) [pt

  9. Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    International Nuclear Information System (INIS)

    Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B.; Fong, R.W.L.; Coleman, C.E.

    1998-08-01

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHIF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 degrees C. (author)

  10. Thermal hydraulic analysis of aggressive secondary cooldown in small break loss of coolant accident with total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, S. J.; Im, H. K.; Yang, J. U.

    2003-01-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). To use RIA, the present study focuses on the detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study is to evaluate the success criteria of Aggressive Secondary Cooldown (ASC) in Small Break Loss Of Coolant Accident (SBLOCA) with total loss of High Pressure Safety Injection (HPSI) and to enhance the understanding of related thermal hydraulic behavior and phenomena. The accident scenario was 2 inch coldleg break LOCA without HPSI, with 1/2 Low Pressure Safety Injection (LPSI), and performing ASC limited by 55.6 .deg. C /hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip, which successively reaches the LPSI condition for about 1.5hr after starting ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria 1204.4 .deg. C (2200 .deg. F). In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that operator should maintain the adequate ASC operation. However, it is necessary to evaluate uncertainties arisen from the related parameters of the ASC operation

  11. Zircaloy-steam reaction under a simulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kawasaki, Satoru; Furuta, Teruo; Hashimoto, Masao

    1975-07-01

    Under a simulated loss-of-coolant condition, the reaction between zircaloy and steam and the embrittlement of the zircaloy oxidized by this reaction have been studied. The parabolic rate constant, ksub(p), in the zircaloy-steam reaction is represented as ksub(p)=3.24x10 6 exp(-40500/RT) (mg 2 /cm 4 . sec) Ring compression test was made on the steam-reacted zircaloy tubes, and following results were obtained: Embrittlement of the steam-reacted zircaloy tube increases with oxidation at each oxidation temperature. For a given quantity of the oxidation, the incursion of α-phase into β-phase is more remarkable in the specimens reacted at low temperatures than those at high temperatures. The embrittlement, however, is larger in the specimens oxidized at high temperatures than those at low temperatures. (auth.)

  12. Assumptions used for evaluating the potential radiological consequences of a loss of coolant accident for boiling water reactors - June 1974

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. The design basis loss of coolant accident is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant design features, or other factors may require different assumptions which will be considered on an individual case basis. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position

  13. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 4 – Integrated chemical effects testing

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Amir; LaBrier, Daniel [Department of Nuclear Engineering, University of New Mexico (United States); Blandford, Edward, E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry [Department of Civil Engineering, University of New Mexico (United States)

    2016-04-15

    Highlights: • Integrated test explored the material release of a postulated large break LOCA. • Aluminum concentration was very low (<0.1 mg/L) throughout the test duration. • Zinc concentration was low (<1 mg/L) in TSP-buffered system. • Calcium release showed two distinguished release zones: prompt and meta-stable. • Copper and iron has no distinguishable concentration up to first 24 h of testing. - Abstract: This paper presents the results of an integrated chemical effects experiment executed under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at the Vogtle nuclear power plant, operated by the Southern Nuclear Operating Company (SNOC). This test was conducted for closure of a series of bench scale experiments conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum (Howe et al., 2015) and zinc (Pease et al., 2015) from metallic surfaces, and calcium from NUKON fiberglass insulation (Olson et al., 2015) . The integrated test was performed in the Corrosion/Chemical Head Loss Experimental (CHLE) facility with representative amounts of zinc, aluminum, carbon steel, copper, NUKON fiberglass, and latent debris. The test was conducted using borated TSP-buffered solution under a post-LOCA prototypical temperature profile lasting for 30 days. The results presented in this article demonstrate trends for zinc, aluminum, and calcium release that are consistent with separate bench scale testing and previous integrated tests under TSP conditions. The release rate and maximum concentrations of the released materials were slightly different than the separate effect testing as a result of different experimental conditions (temperature, surface area-to-water volume ratio) and/or the presence of other metals and chemicals in the integrated test. Samples of metal coupons and fiberglass were selected for analysis using Scanning Electron Microscopy

  14. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 4 – Integrated chemical effects testing

    International Nuclear Information System (INIS)

    Ali, Amir; LaBrier, Daniel; Blandford, Edward; Howe, Kerry

    2016-01-01

    Highlights: • Integrated test explored the material release of a postulated large break LOCA. • Aluminum concentration was very low (<0.1 mg/L) throughout the test duration. • Zinc concentration was low (<1 mg/L) in TSP-buffered system. • Calcium release showed two distinguished release zones: prompt and meta-stable. • Copper and iron has no distinguishable concentration up to first 24 h of testing. - Abstract: This paper presents the results of an integrated chemical effects experiment executed under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at the Vogtle nuclear power plant, operated by the Southern Nuclear Operating Company (SNOC). This test was conducted for closure of a series of bench scale experiments conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum (Howe et al., 2015) and zinc (Pease et al., 2015) from metallic surfaces, and calcium from NUKON fiberglass insulation (Olson et al., 2015) . The integrated test was performed in the Corrosion/Chemical Head Loss Experimental (CHLE) facility with representative amounts of zinc, aluminum, carbon steel, copper, NUKON fiberglass, and latent debris. The test was conducted using borated TSP-buffered solution under a post-LOCA prototypical temperature profile lasting for 30 days. The results presented in this article demonstrate trends for zinc, aluminum, and calcium release that are consistent with separate bench scale testing and previous integrated tests under TSP conditions. The release rate and maximum concentrations of the released materials were slightly different than the separate effect testing as a result of different experimental conditions (temperature, surface area-to-water volume ratio) and/or the presence of other metals and chemicals in the integrated test. Samples of metal coupons and fiberglass were selected for analysis using Scanning Electron Microscopy

  15. Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability

    Science.gov (United States)

    Ha, Taesung

    A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the accident sequence evaluation procedure (ASEP) approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a time-oriented HRA model (reliability physics model) was applied for the human error probability (HEP) estimation of the core relocation. This model is based on two competing random variables: phenomenological time and performance time. The response surface and direct Monte Carlo simulation with Latin Hypercube sampling were applied for estimating the phenomenological time, whereas the performance time was obtained from interviews with operators. An appropriate probability distribution for the phenomenological time was assigned by statistical goodness-of-fit tests. The human error probability (HEP) for the core relocation was estimated from these two competing quantities: phenomenological time and operators' performance time. The sensitivity of each probability distribution in human reliability estimation was investigated. In order to quantify the uncertainty in the predicted HEPs, a Bayesian approach was selected due to its capability of incorporating uncertainties in model itself and the parameters in that model. The HEP from the current time-oriented model was compared with that from the ASEP approach. Both results were used to evaluate the sensitivity of alternative huinan reliability modeling for the manual core relocation in the LOCA risk model. This exercise demonstrated the applicability of a reliability physics model supplemented with a. Bayesian approach for modeling human reliability and its potential

  16. Loss of coolant accident analysis (thermal hydraulic analysis) - Japanese industries experience

    International Nuclear Information System (INIS)

    Okabe, K.

    1995-01-01

    An overview of LOCA analysis in Japanese industry is presented. The BASH-M code, developed for large scale LOCA reflooding analysis, is given as an example of verification and improvement of US computer programs are given. The code's application to the operational safety analysis concerns the following main areas: 1D drift flux model base computer program CANAC; CANAC-based advanced training simulator; emergency operating procedures. The author considers also the code application to the following new PWR safety design concepts: use of steam generators for decay heat removal at LOCA conditions; use of horizontal type steam generator for maintaining two-phase natural circulation under the reactor coolant system submerged. 9 figs

  17. PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

    Energy Technology Data Exchange (ETDEWEB)

    Hadad, Kamal; Esmaeili-Sanjavanmareh, Mansour [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2017-05-15

    PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

  18. Dose to man from a hypothetical loss-of-coolant accident at the Rancho Seco Nuclear Power Plant

    International Nuclear Information System (INIS)

    Peterson, K.R.; Greenly, G.D.

    1981-02-01

    At the request of the Sacramento Municipal Utilities District, we used our computer codes, MATHEW and ADPIC, to assess the environmental impact of a loss-of-coolant accident at the Rancho Seco Nuclear Power Plant, about 40 kilometres southeast of Sacramento, California. Meteorological input was selected so that the effluent released by the accident would be transported over the Sacramento metropolitan area. With the release rates provided by the Sacramento Municipal Utilities District, we calculated the largest total dose for a 24-hour release as 70 rem about one kilometre northwest of the reactor. The largest total dose in the Sacramento metropolitan area is 780 millirem. Both doses are from iodine-131, via the forage-cow-milk pathway to an infant's thyroid. The largest dose near the nuclear plant can be minimized by replacing contaminated milk and by giving the cows dry feed. To our knowledge, there are no milk cows within the Sacramento metropolitan area

  19. SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

    International Nuclear Information System (INIS)

    Gulden, W.

    1977-08-01

    The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

  20. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Höhne, Thomas; Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter

    2013-01-01

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  1. Strategies for reactor safety: Preventing loss of coolant accidents. Final report

    International Nuclear Information System (INIS)

    Lydell, B.O.Y.

    1997-12-01

    This final report on the NKS/RAK-1.2 summarizes the main features of the PIFRAP PC-program and its intended implementation. Regardless of the preferred technical approach to LOCA frequency estimation, the analysis approach must include recognition of the following technical issues: a) Degradation and failure mechanisms potentially affecting piping systems within the reactor coolant pressure boundary (RCPB) and the potential consequences; b) In-service inspection practices and how they influence piping reliability; and c) The service experience with piping systems. The report consists of six sections and one appendix. A Nordic perspective on LOCA and nuclear safety is given. It includes summaries of results from research in material sciences and current regulatory philosophies regarding piping reliability. A summary of the LOCA concept is applied in Nordic PSA studies. It includes a discussion on deterministic and probabilistic views on LOCA. The R and D on piping reliability by SKI and the PIFRAP model is summarized. Next, Section 6 presents conclusion and recommendations. Finally, Appendix A contains a list of abbreviations and acronyms, together with a glossary of technical terms. (EG)

  2. A methodology for the estimation of the radiological consequences of a Loss of Coolant Accident

    Energy Technology Data Exchange (ETDEWEB)

    Kereszturi, Andras; Brolly, Aron; Panka, Istvan; Pazmandi, Tamas; Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). MTA EK, Centre for Energy Research

    2017-09-15

    For calculation of the radiological consequences of Large Break Loss of Coolant (LBLOCA) events, a set of various computer codes modeling the corresponding physical processes, disciplines and their appropriate subsequent data exchange are necessary. For demonstrating the methodology applied in MTA EK, a LBLOCA event at shut down reactor state - when only limited configuration of the Emergency Core Cooling System (ECCS) is available - was selected. In this special case, fission gas release from a number of fuel pins is obtained from the analyses. This paper describes the initiating event and the corresponding thermal hydraulic calculations and the further physical processes, the necessary models and computer codes and their connections. Additionally the applied conservative assumptions and the Best Estimate Plus Uncertainty (B+U) evaluation applied for characterizing the pin power and burnup distribution in the core are presented. Also, the fuel behavior processes. Finally, the newly developed methodology to predict whether the fuel pins are getting in-hermetic or not is described and the the results of the activity transport and dose calculations are shown.

  3. Atucha I nuclear power plant: Probabilistic safety study. Loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Perez, S.S.

    1987-01-01

    The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10 -2 ; b) probability of failure to demand of the high pressure safety injection = 3 x 10 -3 ; c) probability of emergency cooling system failure = 4.4 x 10 -2 . Therefore, the three possible sequences of core meltdown have the following frequencies: λ 1 = 4 x 10 -6 /year λ 2 = 3 x 10 -7 /year λ 3 = 4.4 x 10 -6 /year. (Author)

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  5. MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.; Haste, T.J.

    1982-04-01

    MABEL can be used to determine the cladding deformation in a PWR during a LOCA. It takes the results of calculations from other codes to define the initial fuel condition and the transient whole core thermal-hydraulic behaviour. The use of MABEL with input data appropriate to different regions of a reactor core allows an overall picture of coolant channel blockage within the core to be obtained. (U.K.)

  6. Comparisons of ROSA-III and FIST BWR loss of coolant accident simulation tests

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Suzuki, Mitsuhiro; Koizumi, Yasuo

    1985-10-01

    A common understanding and interpretation of BWR system response and the controlling phenomena in LOCA transients has been achieved through the evaluation and comparison of counterpart tests performed in the ROSA-III and FIST test facilities. These facilities, which are designed to simulate the thermal-hydraulic response of BWR systems, are operated respectively by the Japan Atomic Energy Research Institute (JAERI) and the General Electric Company. Comparison is made between three types of counterpart tests, each performed under similar tests conditions in the two facilities. They are large break, small break, and steamline break LOCA's. The system responses to these tests in each facility are quite similar. The sequence of events are similar, and the timing of these events are similar. Differences that do occur are due to minor differences in modeling objectives, facility scaling, and test conditions. Parallel channel flow interactions effects in the ROSA-III four channel (half length) core, although noticeable in the large break test, do not result in major differences with the single channel response in FIST. In the small break tests the timing of events is offset by the earlier ADS actuation in FIST. The steamline test responses are similar except there is no heatup in FIST, resulting from a different ECCS trip modeling. Overall comparisons between ROSA-III and FIST system responses in LOCA tests is very good. (author)

  7. On the transient pressure build-up in the full pressure safety shell of watercooled nuclear reactors after a loss of coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1979-08-01

    The thermo-and fluid-dynamic processes in a multichamber full pressure safety containment during a loss of coolant accident have been investigated. Comparison of the calculations carried out with the computer programs, in which ZOCO VI was used as being representative of similar programs, with the experimental results pointed out discrepancies in the determination of time dependent pressure, pressure difference and temperature curves. This led to the development of a new theoretical model and a program COFLOW which pays particular attention to the fluid dynamic processes in the initial phase of a loss of coolant accident. It can also be used to determine the maximum containment pressure towards the end of a loss of coolant accident. Comparison of the COFLOW results with experiments has shown that COFLOW provides a model and a procedure by which the physical processes in a multichamber full pressure safety containment can be simulated satisfactorily

  8. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-01-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

  9. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-12-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

  10. Comparison of methods for calculation of large cladding deformation in the case of a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Fabian, H.; Krugmann, U.; Lassmann, K.; Schwarz, R.

    1975-06-01

    Some results of mechanical computations of cladding deformation are discussed for the case of a loss-of-coolant accident. The models for data-creation realize isothermal and transient conditions. The creep-deformation of the cladding is caused by significant temperature and pressure profiles. In all cases the constitutive creep law of Norton is used. The computations are based on three methods: 1) analytical solution (one-dimensional), 2) finite element solution (two-dimensional), 3) theory of creeping shells (two-dimensional). The differences in the solutions depend on the methods themselves and on computational differences. The influence of the large-deflection theory is discussed. In comparing the results it is evident that the differences in the methods are covered by a small variation of the creep parameters. In conclusion we propose the theory of the creeping shell for extensive computer codes. (orig.) [de

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

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Wachs, D. M.

    1998-01-01

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

  13. Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Conlon, W.M.; Lahey, R.T. Jr.

    1982-12-01

    This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

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

    International Nuclear Information System (INIS)

    Lime, J.F.; Boyack, B.E.

    1996-01-01

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

  15. The oxidation kinetics of zirconium alloys applicable to loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Parsons, P.D.; Miller, W.N.

    1977-10-01

    A review is presented of the available published measurements of the rate of reaction between zirconium alloys and steam and, in some cases, oxygen. Attempts are made to define from all the experimental data a suitable rate equation which is appropriate over the range of temperatures relevant to LOCA conditions. The data reviewed encompass a temperature range 910 0 C to the melting point of zirconium, 1852 0 C. It can be concluded that within 910 to 1577 0 C, Zircaloy-2, Zircaloy-4 and Zr/2 1/2%Nb alloys have the same response to oxidation. (author)

  16. Multi-rod burst behavior under a loss-of-coolant accident condition, (1)

    International Nuclear Information System (INIS)

    Hashimoto, Masao; Otomo, Takashi; Furuta, Teruo; Kawasaki, Satoru; Uetsuka, Hiroshi

    1980-12-01

    Multi-rod burst tests have been planned since 1977 to estimate quantitative channel restriction during a LOCA transient in LWRs. For this purpose, many bundle tests have been making to burst in a steam in varying a few parameters which influence the degree of channel restriction. The purpose of this report is to provide a background document for final reports to be published in the future. This report includes the results of No. 7805 bundle test, namely temperature, internal pressure, burst behavior of rods and channel restriction of the bundle. (author)

  17. Long-term security of electrical and control engineering equipment in nuclear power stations to withstand a loss of coolant accident

    International Nuclear Information System (INIS)

    Mueller, H.

    1996-01-01

    Electrical and control engineering equipment, which has to function even after many years of operation in the event of a fault in a saturated steam atmosphere of 160 C maximum, is essential in nuclear power stations in order to control a loss of coolant accident. The nuclear power station operators have, for this purpose, developed verification strategies for groups of components, by means of which it is ensured that the electrical and control engineering components are capable of dealing with a loss of coolant accident even at the end of their planned operating life. (orig.) [de

  18. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Govers, K.; Verwerft, M.

    2016-01-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive. - Highlights: • We performed Discrete Element Methods simulation for fuel relocation and dispersal during LOCA transients. • The approach provides a mechanistic description of these phenomena. • The approach shows the ability of the technique to reproduce experimental observations. • The packing fraction in the balloon is shown to stabilize at 50–60%.

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

    International Nuclear Information System (INIS)

    Yackle, T.R.

    1979-09-01

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

  20. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  1. Status Report on Spent Fuel Pools under Loss-of-Cooling and Loss-of-Coolant Accident Conditions - Final Report

    International Nuclear Information System (INIS)

    Adorni, M.; Esmaili, H.; Grant, W.; Hollands, T.; Hozer, Z.; Jaeckel, B.; Munoz, M.; Nakajima, T.; Rocchi, F.; Strucic, M.; ); Tregoures, N.; Vokac, P.; Ahn, K.I.; Bourgue, L.; Dickson, R.; Douxchamps, P.A.; Herranz, L.E.; Jernkvist, L.O.; Amri, A.; Kissane, M.P.; )

    2015-01-01

    scenarios, past accidents and precursor events; Chapter 4: Behaviour of spent fuel facilities during the Fukushima Daiichi accident; Chapter 5: Accident phenomenology; Chapter 6: Experiments with relevance to SFP cooling accidents; Chapter 7: Simulation tools; Chapter 8: Conclusions and recommendations; The present report summarizes results of experiments and computational analyses carried out to date to gain understanding of phenomena with significance to SFP cooling accidents. Considering that some knowledge gaps currently exist and that ongoing and planned research projects are expected to produce results that will hopefully narrow these gaps within the foreseeable future, it is recommended that: - a CSNI state-of-the-art report on SFP loss-of-cooling and loss-of-coolant accidents is written as the results of these research projects become available; - a follow-on activity is launched on SFP combining probability of SFP accidents, which was beyond the scope of this document, and mitigation strategies

  2. Sensitivity analysis on the zirconium ignition in a postulated SFP loss of coolant accident

    International Nuclear Information System (INIS)

    Park, Sanggil; Lee, Jaeyoung; Kim, Sun-ki; Chun, Tae-hyun; Bang, Je-geon

    2016-01-01

    From both SFP complete LOCA experiments, it was observed that zirconium alloy cladding temperature was abruptly increased at a certain point and the cladding was almost fully oxidized. To capture this phenomenon, the concept of air oxidation breakaway model was adopted in MELCOR code. This paper examines this air oxidation breakaway model by comparing the SFP project test data and MELCOR code calculation results by using this model. The air oxidation model parameters are slightly altered to see their sensitivities on the occurrence of the zirconium ignition. Through such sensitivity analysis, limitations of the air oxidation breakaway model are revealed in comparison to the actual zirconium ignition phenomenon during air ingress scenarios. In addition, ways to overcome the identified limitations of the air oxidation model are recommended to estimate better the zirconium ignition phenomenon in SFP sequences. In this paper, the zirconium ignition phenomenon was reviewed and the model to capture this phenomenon was investigated. The model is the air oxidation breakaway model in MELCOR code, and its sensitivity of the model parameters on the time to ignition was studied. From the sensitivity analysis, the slight change of model parameters induce the large variation of the time to ignition. The model itself includes its weakness to fully represent both the air oxidation breakaway phenomenon and the followed zirconium ignition behavior. Furthermore, this model considers no effect of N2 on the cladding degradation and its promoted exothermic heat release

  3. Sensitivity analysis on the zirconium ignition in a postulated SFP loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sanggil; Lee, Jaeyoung [Handong Global Univ., Pohang (Korea, Republic of); Kim, Sun-ki; Chun, Tae-hyun; Bang, Je-geon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    From both SFP complete LOCA experiments, it was observed that zirconium alloy cladding temperature was abruptly increased at a certain point and the cladding was almost fully oxidized. To capture this phenomenon, the concept of air oxidation breakaway model was adopted in MELCOR code. This paper examines this air oxidation breakaway model by comparing the SFP project test data and MELCOR code calculation results by using this model. The air oxidation model parameters are slightly altered to see their sensitivities on the occurrence of the zirconium ignition. Through such sensitivity analysis, limitations of the air oxidation breakaway model are revealed in comparison to the actual zirconium ignition phenomenon during air ingress scenarios. In addition, ways to overcome the identified limitations of the air oxidation model are recommended to estimate better the zirconium ignition phenomenon in SFP sequences. In this paper, the zirconium ignition phenomenon was reviewed and the model to capture this phenomenon was investigated. The model is the air oxidation breakaway model in MELCOR code, and its sensitivity of the model parameters on the time to ignition was studied. From the sensitivity analysis, the slight change of model parameters induce the large variation of the time to ignition. The model itself includes its weakness to fully represent both the air oxidation breakaway phenomenon and the followed zirconium ignition behavior. Furthermore, this model considers no effect of N2 on the cladding degradation and its promoted exothermic heat release.

  4. Multi-rod burst test under a loss-of coolant accident condition, (4)

    International Nuclear Information System (INIS)

    Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi

    1983-06-01

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

  5. Development of a deformation and failure model for Zircaloy at high temperatures for light water reactor loss-of-coolant-accident investigations

    International Nuclear Information System (INIS)

    Raff, S.

    1982-11-01

    To describe Zircaloy-4 deformation and failure behaviour at high temperatures (600 to 1400 0 C), the phenomenological model NORA was developed and verified against numerous experimental results. The model can be applied to the calculation of fuel rod cladding deformation during small and large break loss-of-coolant-accidents. (orig./RW) [de

  6. Sensitivity and uncertainty analysis for Ignalina NPP confinement in case of loss of coolant accident

    International Nuclear Information System (INIS)

    Urbonavicius, E.; Babilas, E.; Rimkevicius, S.

    2003-01-01

    At present the best-estimate approach in the safety analysis of nuclear power plants is widely used around the world. The application of such approach requires to estimate the uncertainty of the calculated results. Various methodologies are applied in order to determine the uncertainty with the required accuracy. One of them is the statistical methodology developed at GRS mbH in Germany and integrated into the SUSA tool, which was applied for the sensitivity and uncertainty analysis of the thermal-hydraulic parameters inside the confinement (Accident Localisation System) of Ignalina NPP with RBMK-1500 reactor in case of Maximum Design Basis Accident (break of 900 mm diameter pipe). Several parameters that could potentially influence the calculated results were selected for the analysis. A set of input data with different initial values of the selected parameters was generated. In order to receive the results with 95 % probability and 95 % accuracy, 100 runs were performed with COCOSYS code developed at GRS mbH. The calculated results were processed with SUSA tool. The performed analysis showed a rather low dispersion of the results and only in the initial period of the accident. Besides, the analysis showed that there is no threat to the building structures of Ignalina NPP confinement in case of the considered accident scenario. (author)

  7. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun, E-mail: pdj@kaeri.re.kr; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-15

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility. - Highlights: • Cr and FeCrAl were coated onto Zr fuel cladding for light water nuclear reactors. • Mo layer between FeCrAl and Zr prevented inter-diffusion at high temperatures. • Coated claddings were tested under loss-of-cooling accident conditions. • Coating improved high-temperature oxidation resistance and mechanical properties.

  8. Computer programmes of the Power Research Institute for the analysis of processes in the primary coolant circuit and in the containment of a WWER plant in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Misak, J.

    1976-01-01

    A brief description is given of computer programmes for the analysis of loss-of-coolant accidents (LOCA) in WWER type reactors. The LENKA programme is intended for the thermal and hydraulic analysis of the consequences of such accidents in the primary coolant circuit. The SICHTA programme is intended for the detailed calculation of the time dependence of the axial and radial distribution of heat in fuel rods from steady-state to the flooding of the core. CHEMLOC is intended for the analysis of the heat history of the core and the extent of chemical reactions in LOCA when the emergency core cooling system is not operating. The TRACO I is intended for the analysis of the initial stage of the transient process in a full-pressure containment after LOCA (the computation of the time and spatial dependences of pressures and temperatures). TRACO III is intended for the computation of the long-term time dependence of pressure and temperature in the full-pressure containment after LOCA. (B.S.)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  11. TRAC calculations of a loss-of-coolant accident in a reactor scale model

    International Nuclear Information System (INIS)

    Pyun, J.J.

    1981-01-01

    The TRAC (Transient Reactor Analysis Code) is being developed at the Los Alamos National Laboratory as an advanced best-estimate computer program for analysis of postulated hypothetical accidents in pressurized water reactors. As a part of the TRAC developmental verification efforts, a TRAC posttest analysis of Semiscale Mod-3 Test S-07-6 was conducted. The results of this analysis show that the agreement between TRAC calculations and experimental data is not very good. In particular, TRAC does not predict the long term doncomer and core liquid level oscillations during the reflood phase

  12. Review of boiling water reactor small break loss of coolant accidents

    International Nuclear Information System (INIS)

    Gururaj, P.M.; Dua, S.S.; Rao, A.S.

    1981-01-01

    This paper presents a review of the analytical and the experimental work performed by the General Electric Company to determine the performance of boiling water reactors (BWR) following postulated small break accidents (SBA). This review paper addresses the following issues: (1) the response of the BWR following small loss of inventory events; (2) methods of analysis and their justification; (3) necessity, if any, of operator action and the length of time available in which such action can be performed; and (4) operator interface following the SBA event. The results from these SBA studies for different BWR product lines show that even with the multiple system failures assumed, the BWR can successfully withstand an SBA. For a typical BWR/6, it takes the failure of 13 water delivery pumps to cause any significant core heatup. The only operator actions determined to be necessary are simple ones and ample time is available to the operator to perform these actions, if needed

  13. Permeability and compression of fibrous porous media generated from dilute suspensions of fiberglass debris during a loss of coolant accident

    International Nuclear Information System (INIS)

    Lee, Saya; Abdulsattar, Suhaeb S.; Vaghetto, Rodolfo; Hassan, Yassin A.

    2015-01-01

    Highlights: • Experimental investigation on fibrous debris buildup was conducted. • Head loss through fibrous media was recorded at different approach velocities. • A head loss model through fibrous media was proposed for high porosity (>0.99). • A compression model of fibrous media was developed. - Abstract: Permeability of fibrous porous media has been studied for decades in various engineering applications, including liquid purifications, air filters, and textiles. In nuclear engineering, fiberglass has been found to be a hazard during a Loss-of-Coolant Accident. The high energy steam jet from a break impinges on surrounding fiberglass insulation materials, producing a large amount of fibrous debris. The fibrous debris is then transported through the reactor containment and reaches the sump strainers. Accumulation of such debris on the surface of the strainers produces a fibrous bed, which is a fibrous porous medium that can undermine reactor core cooling. The present study investigated the buildup of fibrous porous media on two types of perforated plate and the pressure drop through the fibrous porous media without chemical effect. The development of the fibrous bed was visually recorded in order to correlate the pressure drop, the approach velocity, and the thickness of the fibrous porous media. The experimental results were compared to semi-theoretical models and theoretical models proposed by other researchers. Additionally, a compression model was developed to predict the thickness and the local porosity of a fibrous bed as a function of pressure

  14. 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

  15. Experimental investigation of material chemical effects on emergency core cooling pump suction filter performance after loss of coolant accident

    International Nuclear Information System (INIS)

    Park, Jong Woon; Park, Byung Gi; Kim, Chang Hyun

    2009-01-01

    Integral tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for 30 days after a loss of coolant accident. A test rig with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber and the amounts of reactor building materials are scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of calcium-silicate, reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON TM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

  16. Analysis of large break loss of coolant accident with simultaneous injection into cold leg and hot leg

    International Nuclear Information System (INIS)

    Luo Bangqi

    1997-01-01

    When a large break loss of coolant accident occurs, the most part of the safety injection water injected into the cold leg by the safety injection system will flow through the channel between the pressure vessel and the barrel out of the break into the containment, only a little part of the safety injection water can flow into the reactor core. If the safety injection can inject into both the cold leg and the hot leg simultaneously, the safety injection water injected from the cold leg will flow into the core more easily, because the safety injection water injected from the hot leg will carry out more heat from the upper plenum and the core, so the upper plenum and the core is depressed. In addition, a small part of the safety injection water injected from the hot leg will flow down in the core after impinging the guide tubes in the upper plenum, so the core will get more safety injection water than only cold leg injection, and the core will be much safer

  17. Analysis of steam condensation in APR1400 IRWST for loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young Suk

    2006-02-15

    The In-Containment Refueling Water Storage Tank (IRWST) of APR1400 is installed at the bottom of containment building to promote the plant safety functions during an accident. This design feature brings about uncertainty factors which may necessitate conventional prediction of temperature and pressure of containment building improved or revised when an accident occurs. The hot steam which is released from RCS break enters the IRWST through four Pressure Relief Dampers (PRDs). It is expected to be condensed with water stored in IRWST, in which water is colder than incoming steam. The purpose of this study is to investigate the influence of IRWST and pressure relief damper on back pressure and temperature in APR1400 containment codes such as CONTEMPT-LT and GOTHIC. The comparison of codes showed that GOTHIC code be more appropriate for the prediction of containment pressure and temperature under the condition of steam condensation occurring in confined water pool. Especially, the GOTHIC has superior capability to treat multi-compartmentalized geometry This study developed one-compartment (single) model, two-compartment (separated) model, and three-dimension (3-D) model, respectively. Two compartment model separates the IRWST from the other containment compartments. In 3-D model, only the IRWST is nodalized with Cartesian modeling. The single model is developed for comparison with two-compartment model which can analyze PRD's influence. The separated model for predicting PRD's influence divides the space between containment and IRWST. 3-D model for IRWST was generated because it is not symmetric considering location of sparger, pump, and suction sump. Therefore, IRWST is simulated with not only detailed three-dimensional behavior but also independent flow paths for four PRDs. Many experimental studies for the direct-contact heat transfer in stratified steam water flows, cocurrent or countercurrent, have been performed (Segev et al., 1981; Lim et al., 1981

  18. Analysis of steam condensation in APR1400 IRWST for loss of coolant accident

    International Nuclear Information System (INIS)

    Oh, Young Suk

    2006-02-01

    The In-Containment Refueling Water Storage Tank (IRWST) of APR1400 is installed at the bottom of containment building to promote the plant safety functions during an accident. This design feature brings about uncertainty factors which may necessitate conventional prediction of temperature and pressure of containment building improved or revised when an accident occurs. The hot steam which is released from RCS break enters the IRWST through four Pressure Relief Dampers (PRDs). It is expected to be condensed with water stored in IRWST, in which water is colder than incoming steam. The purpose of this study is to investigate the influence of IRWST and pressure relief damper on back pressure and temperature in APR1400 containment codes such as CONTEMPT-LT and GOTHIC. The comparison of codes showed that GOTHIC code be more appropriate for the prediction of containment pressure and temperature under the condition of steam condensation occurring in confined water pool. Especially, the GOTHIC has superior capability to treat multi-compartmentalized geometry This study developed one-compartment (single) model, two-compartment (separated) model, and three-dimension (3-D) model, respectively. Two compartment model separates the IRWST from the other containment compartments. In 3-D model, only the IRWST is nodalized with Cartesian modeling. The single model is developed for comparison with two-compartment model which can analyze PRD's influence. The separated model for predicting PRD's influence divides the space between containment and IRWST. 3-D model for IRWST was generated because it is not symmetric considering location of sparger, pump, and suction sump. Therefore, IRWST is simulated with not only detailed three-dimensional behavior but also independent flow paths for four PRDs. Many experimental studies for the direct-contact heat transfer in stratified steam water flows, cocurrent or countercurrent, have been performed (Segev et al., 1981; Lim et al., 1981; Kim and

  19. Upper plenum dump during reflood in PWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Sudo, Yukio; Griffith, Peter.

    1981-01-01

    Upper plenum dump during reflood in a large break loww-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnetude of water subcooling. (author)

  20. Examples of Unsafe Act Identification from Simulator Training Records for Interfacing System Loss of Coolant Accident

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Yeong; Kim, Yochan; Park, Jinkyun; Kim, Seunghwan; Jung, Wondea [Korea Atomic Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Operating procedures such as EOPs (Emergency Operating Procedures) and AOPs (Abnormal Operating Procedures) have been developed to maximize the operator’s performance during emergency/abnormal situations of critical-safety systems. In this regard, it is very important to point out that one of the significant factors causing accidents or incidents is an inappropriate human performance of operating personnel working in the socio-technical systems. A lot of efforts to collect HRA data by using a simulator of NPP have progressed. We developed a standardized guideline to specify how to gather HRA data from simulator training records, and created IGT (Information Gathering Template) specifying what kinds of measures should be observed during the simulations and defined UA (Unsafe Act) and describe the UA identification method under interactions between crew members to suggest a practical UA type classification scheme under a procedure driven operation. We also developed a framework for data collection and analysis to produce HEPs. The framework is named HuREX (Human Reliability data Extraction) system. In this paper, we described a process to identify UAs as well as UA candidates during an AOP/EOP operation with simulator training records. We presented examples of UA candidates and UAs grouped by consequences based on UA identification criteria. Based on this research, we are to achieve insights about the UA pattern and procedure instruction in which UAs occurred frequently. With this result, we are to analyze the root cause of UAs to find a way to reduce UAs.

  1. Methodology for the Assessment of Confidence in Safety Margin for Small Break Loss of Coolant Accident Sequences

    Energy Technology Data Exchange (ETDEWEB)

    Nagrale, D. B.; Prasad, M.; Rao, R. S.; Gaikwad, A.J., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used concurrently to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis is increasingly being used for deterministic calculation in NPPs. The PSA methodology aims to be as realistic as possible while integrating information about accident phenomena, plant design, operating practices, component reliability and human behaviour. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with the concept of calculating the peak clad temperature with 95 percent confidence and 95 percent probability (PCT{sub 95/95}) in small break loss of coolant accident (SBLOCA) and methodologies for assessing safety margin. Five input parameters mainly, nominal power level, decay power, fuel clad gap conductivity, fuel thermal conductivity and discharge coefficient, were selected. A Uniform probability density function was assigned to the uncertain parameters and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for 5 parameters were randomly mixed by LHS to obtain 25 input sets. A non-core damage accident sequence was selected from the SBLOCA event tree of a typical VVER study to estimate the PCTs and safety margin. A Kolmogorov– Smirnov goodness-of-fit test was carried out for PCTs. The smallest value of safety margin would indicate the robustness of the system with 95% confidence and 95% probability. Regression analysis was also carried out using 1000 sample size for the estimating PCTs. Mean, variance and finally safety margin were analysed. (author)

  2. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

  3. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

  4. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-03-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). The present study focuses on detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model using RIA for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study in this year is to evaluate the success cri-teria of Aggressive Secondary Cooldown (ASC) in a Small Size Loss Of Coolant Accident (SBLOCA) without HPSI and to enhance the understanding of related thermal hydraulic behavior and phenomena. An effort was made to evaluate the system success criteria and a mission time for the recovery action by an operator to prevent the core damage for that accident scenario. The accident scenario for KSNP was a 2 inch coldleg break LOCA with a total loss of High Pressure Safety Injection (HPSI) and 1/2 Low Pressure Safety Injection (LPSI) available and perform-ing ASC limited by 55.6 .deg. C/hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip. It successively reached the LPSI condition for about 1.5hr after starting the ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria of 1204.4 .deg. C (2200 .deg. F). Sensitivity studies were performed for (1) cool-ant average temperature parameters, (2) ASC operation control method, (3) operation start time, (4) 1 inch break size. The present analysis identified thermal hydraulic phenomena and parameters affecting on the behavior, which consist of coolant break flow and inventory, parameters governing secondary heat removal, ASC operation control method, and its reference temperature parameters. In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that an operator should maintain the ade-quate ASC operation. However, it is necessary to evaluate the uncertainties arisen from the

  5. Investigation of a hydrogen mitigation system during large break loss-of-coolant accident for a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohmmad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

  6. Analysis of the effects of the pressure wave generated in loss of coolant accidents in reactor vessels

    International Nuclear Information System (INIS)

    Valero Martinez, M.

    1980-01-01

    The increasing demands in the field of ''Nuclear Safety'', obliges to a perfect knowledge of the causes and effects of every possible accident in a nuclear power plant. In this paper will be analysed the effects of the pressure wave appearing in a LOCA (Loss of collant accident). The pressure wave could deform the following structures: core barrel wall, cover and bottom, control rods and safety coolant system. Any change of the geometry of these structures could provoke and incorrect system reaction after the accident has happened. The basis and hypothesis for the theoretical analysis will be exposed. The structures are considered to be rigid. A typical boiling water be analysed and the developed theory will be verified in comparations with experimental results and the results obtained with some others models. Due to the easy application and short calculation time of the created programmes, they are recommended for parametrical calculations in the analysis of the pressurized water reactors and boiling water reactors. (author)

  7. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  8. Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Callow, R.A.

    1989-01-01

    The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

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

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1975-01-01

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

  10. Utilization of DRUFAN 01/MOD 02 computer code for the depressurization phase analysis of a postulated loss of coolant accident in Angra 2/3 Nuclear Power Plants

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.; Figueiredo, P.J.M.

    1985-08-01

    The DRUFAN 01/Mod 2 developed by Gesellschaft fur Reaktorsicherheit (GRS) mbh to simulate thermohydraulic behavior of the primary circuit of PWR reactors, during the despressurization phase and initial refilling phase of loss of coolant accidents by great ruptures, is presented. The program simulates the system to be analysed by control volumes-concentrated parameters model - and it is based on numerical solution of conservation equations for mass of water, mass of vapor, quantities of motion and energy, and on the control volume homogeneity hypothesis. The possibilities of thermodynamic disequilibrium, determining mass transfer between liquid and vapor phases assuming that one saturated phase, are considered. The process of computer code implantation in the Honeywell Bull 64 DPS 7 system at CNEN, the modifications done into the program and the application to the despressurization phase analysis of a loss of coolant accident at Angra-2 and Angra-3 reactors are considered. (M.C.K.) [pt

  11. Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

    International Nuclear Information System (INIS)

    Zhang Yanzhao; Zhang Fan; Zhao Xinwen; Zheng Yingfeng

    2013-01-01

    Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

  12. Determination of the in-containment source term for a Large-Break Loss of Coolant Accident

    International Nuclear Information System (INIS)

    2001-04-01

    This is the report of a project that focused on one of the most important design basis accidents: the Large Break Loss Of Coolant Accident (LBLOCA) (for pressurised water reactors). The first step in the calculation of the radiological consequences of this accident is the determination of the source term inside the containment. This work deals with this part of the calculation of the LBLOCA radiological consequences for which a previous benchmark (1988) has shown wide variations in the licensing practices adopted by European countries. The calculation of this source term may naturally be split in several steps (see chapter II), corresponding to several physical stages in the release of fission products: fraction of core failure, release from the damaged fuel, airborne part of the release and the release into the reactor coolant system and the sumps, chemical behaviour of iodine in the aqueous and gas phases, natural and spray removal in the containment atmosphere. A chapter is devoted to each of these topics. In addition, two other chapters deal with the basic assumptions to define the accidental sequence and the nuclides to be considered when computing doses associated with the LBLOCA. The report describes where there is agreement between the partner organisations and where there are still differences in approach. For example, there is agreement concerning the percentage of failed fuel which could be used in future licensing assessments (however this subject is still under discussion in France, a lower value is thinkable). For existing plants, AVN (Belgium) wishes to keep the initial licensing assumptions. For the release from damaged fuel, there is not complete agreement: AVN (Belgium) wishes to maintain its present approach. IPSN (France), GRS (Germany) and NNC (UK) prefer to use their own methodologies that result in slightly different values to the proposed values for a common position. There are presently no recommendations of the release of fuel particulates

  13. Utilization of the RELAP4/MOD5/SAS code version in loss of coolant accident in the Angra 1 nuclear power station

    International Nuclear Information System (INIS)

    Sabundjian, G.; Freitas, R.L.

    1991-09-01

    A new version of computer code RELAP4/MOD5 was developed to improve the output. The new version, called RELAP4/MOD5/SAS, prints the main variables in graphical form. In order to check the program, a 36 - volume simulation of the Loss-of-Coolant Accident for Angra - I was performed and the results compared to those of a existing 44 - volume simulation showed a satisfactory agreement with a substantial reduction in computing time. (author)

  14. A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

    International Nuclear Information System (INIS)

    Langman, V.J.; Weaver, K.R.

    1984-05-01

    The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

  15. 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)

  16. Application of realistic (best- estimate) methodologies for large break loss of coolant (LOCA) safety analysis: licensing of Westinghouse ASTRUM evaluation model in Spain

    International Nuclear Information System (INIS)

    Lage, Carlos; Frepoli, Cesare

    2010-01-01

    When the LOCA Final Acceptance Criteria for Light Water Reactors was issued in Appendix K of 10CFR50 both the USNRC and the industry recognized that the rule was highly conservative. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the USNRC began a research program to identify the degree of conservatism in those models permitted in the Appendix K rule and to develop improved thermal-hydraulic computer codes so that realistic accident analysis calculations could be performed. The overall results of this research program quantified the conservatism in the Appendix K rule and confirmed that some relaxation of the rule can be made without a loss in safety to the public. Also, from a risk-informed perspective it is recognized that conservatism is not always a complete defense for lack of sophistication in models. In 1988, as a result of the improved understanding of LOCA phenomena, the USNRC staff amended the requirements of 10 CFR 50.46 and Appendix K, 'ECCS Evaluation Models', so that a realistic evaluation model may be used to analyze the performance of the ECCS during a hypothetical LOCA. Under the amended rules, best-estimate plus uncertainty (BEPU) thermal-hydraulic analysis may be used in place of the overly prescriptive set of models mandated by Appendix K rule. Further guidance for the use of best-estimate codes was provided in Regulatory Guide 1.157 To demonstrate use of the revised ECCS rule, the USNRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology as an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis. More recently the CSAU principles have been generalized in the Evaluation Model Development and Assessment Process (EMDAP) of Regulatory Guide 1.203. ASTRUM is the Westinghouse Best Estimate Large Break LOCA evaluation model applicable to two-, three

  17. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  18. Analysis of Consequences in the Loss-of-Coolant Accident in Wendelstein 7-X Experimental Nuclear Fusion Facility

    Energy Technology Data Exchange (ETDEWEB)

    Uspuras, E., E-mail: algis@mail.lei.lt [Laboratory of Nuclear Installations Safety, Lithuanian Energy Institute, Kaunas (Lithuania)

    2012-09-15

    . The results of analysis demonstrated that proposed burst disk, connecting the plasma vessel with torus hall, opens and pressure inside plasma vessel do not exceed the limiting 1100 kPa absolute pressure. Thus, the plasma vessel remains intact after loss-of-coolant accident during no-plasma operation of Wendelstein 7-X experimental nuclear fusion facility. (author)

  19. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  20. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  1. An assessment of the individual and social risks of Athens population resulting from a hypothetical loss-of-coolant-accident release of the Greek Research Reactor-1

    International Nuclear Information System (INIS)

    Kollas, John; Synodinou, Varvara; Varsamis, G.; Antoniades, John; Catsaros, Nicolas.

    1984-03-01

    In this report the loss-of-coolant-accident consequences for the Greek Research Reactor-1 which is located within the limits of Athens are estimated. The source term emerges from a conservative 20% coremelt with 25 isotopes taken into consideration. Individual and social risks are calculated to a distance of 20 km from the reactor site, an area covering the whole Athens region of 3,081,000 inhabitants. Latent health effects due to both initial an chronic exposure from inhalation of resuspended radionuclides and exposure to groundshine from contaminated ground are assessed. (author)

  2. Laboratory simulation of rod-to-rod mechanical interactions during postulated loss-of-coolant accidents in a PWR involving cladding oxidation

    International Nuclear Information System (INIS)

    Hindle, E.D.; Haste, T.J.; Harrison, W.R.

    1987-01-01

    Creep deformation of Zircaloy cladding in postulated PWR loss-of-coolant accidents may lead to rod-to-rod mechanical interactions. Tests have been performed in the electrically heated FOURSQUARE rig at 750 0 C and 850 0 C in steam to investigate this effect. Conservatisms inherent in a simple 'square with rounded corners' coolant channel blockage model have been quantified; about 5-10% flow area may remain even at strains which in ideal circumstances would give total blockage. Reduction of average burst strains produced by an oxide layer (up to 13 μm) has been demonstrated, resulting from strain concentration at oxide cracks. (author)

  3. Conception of a model for the description of the rewetting phase of reactor fuel pins following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hinderer, B.; Schuetzle, R.

    1976-10-01

    The aim of the present paper has been the development of a model describing rewetting of fuel rods in the reflood phase after a loss of coolant accident of a reactor. Because a suitable solution to the problem could not be found an appropriate model has been implemented into an IKE computer program for transient, two-dimensional heat conductance for a cylindrical rod. Developing this model experimental results of up-to-date literature were used. Remarkable is that very small meshes are necessary around the rewetting front to calculate the rewetting velocity which is strongly dependent on the quench temperature. (orig.) [de

  4. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 3—Calcium

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Sterling; Ali, Amir; LaBrier, Daniel [Department of Nuclear Engineering, University of New Mexico (United States); Blandford, Edward D, E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry [Department of Civil Engineering, University of New Mexico (United States)

    2016-04-15

    Highlights: • Calcium leaching from NUKON fiberglass in borated TSP-buffered solution is independent of the level of fiberglass destruction. • The initial calcium release rate and the maximum calcium concentration increases with increased fiber concentration. • The calcium release in solution has a repeatable pattern of four distinct regions (prompt release, metastable, autocatalytic drop, and stable region) for all experiments. • Magnesium plays a significant role in initiating calcium precipitation in TSP-buffered environment. • Head loss through multi-constituents debris beds was found to increase progressively in all calcium concentration regions. - Abstract: Calcium that leaches from damaged or destroyed NUKON fiberglass in containment post a loss of coolant accident (LOCA) could lead to the formation of chemical precipitates. These precipitates could be filtered through the accumulated fibrous debris on the sump screen and compromising the emergency core cooling system (ECCS) sump pump performance. Reduced-scale leaching experiments were conducted on three solution inventory scales—bench (0.5 L), vertical column (31.5 L), and tank (1136 L) using three different flow conditions, and fiberglass concentrations (1.18–8 g/L) to investigate calcium release from NUKON fiber. All experiments were conducted in simulated post-LOCA water chemistry. (∼220 mM boric acid with ∼5.8 mM trisodium phosphate (TSP) buffer). Prior to the leaching tests, a preliminary experiment was carried out on the bench scale to determine the effect of the fiber preparation (unaltered and blended) method on calcium leaching. Results indicate that the extent of fiberglass destruction does not affect the amount of calcium released from fiberglass. Long-term calcium leach testing at constant temperature (80 °C) in borated TSP-buffered solution had repeatable behavior on all solution scales for different fiberglass concentrations. The calcium-leaching pattern can be divided into

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

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1976-06-01

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

  6. Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

    International Nuclear Information System (INIS)

    Kheshtpaz, H.; Alison, C.

    2006-01-01

    As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

  7. Reduction of the interlocking potential of sump sieves by corrosion products as consequence of loss-of-coolant accidents; Verminderung des Verblockungspotenzials von Sumpfansaugsieben durch Korrosionsprodukte nach Kuehlmittelverluststoerfaellen

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Wolfgang; Kryk, Holger [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany). Inst. fuer Fluiddynamik

    2012-11-01

    In German nuclear power plants thermal insulation fragmentation as a consequence of loss-of-coolant accidents have not been identified, but recently significant pressure increase in the sump sieves due to corrosion products have been observed. The corrosion products are released from hot-galvanized steel grids by steam jet fragmentation. It was shown that critical deposition of corrosion products can occur in the long-term process of the accident. The hazard of sieve blocking could be reduced by zinc containing chemicals or an increase of the pH value (to about 6.7). The possibility of disadvantageous consequences of resulting chemical reactions has to be investigated in the future.

  8. Pressure-temperature response of a full-pressure PWR containment to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Misak, J.

    1976-01-01

    A mathematical model and computer code TRACO III for pressure-temperature transients in the full-pressure containment of PWR during LOCA is described. Main attention is devoted to the analysis of parametric calculations with respect to the estimation of effect of various factors on the transient process and to the comparison of the theoretical and the experimental results on CVTR. (author)

  9. A study on the loss-of-coolant accidents associated with the lung-men nuclear power station

    International Nuclear Information System (INIS)

    Teng, J.T.; Hsu, C.T.; Wang, T.Q.; Chen, Y.H.; Wang, L.C.; Chung, N.M.; Yuann, R.Y.

    2001-01-01

    This study was intended to evaluate the behavior of the nuclear core of the Lung-Men Nuclear Power Station (LMNPS) under postulated LOCA conditions. The LMNPS construction is now in suspense by the Ministry of Economic Affairs, the Republic of China. The assumptions used in this study were in compliance with the requirements specified in 10CFR50.46 and Appendix K. The methodology used was primarily RELAP5YA, which was a modification to the RELAP5/MOD1 Cycle 18. In the paper, features of the thermo-fluids, neutronics, flow systems, trips, and breaks are discussed. Their assumptions and the resulting implications to the outcome of the analyses are emphasized. Also typical sequences of events, the reactor pressure vessel (RPV) pressure, temperature and water inventory transients, and the ultimate core heat-ups for a number of break sizes, ranging from small- to large-break LOCAs, are delineated. The results of this study indicated that for all cases studied, the peak cladding temperature (PCT) was 699.1 Celsius degrees (1290.4 F). This PCT was much lower than the upper temperature limit of 1204.4 Celsius degrees (2200 F) specified in the acceptance criterion of 10CFR50.46. It is to be noted that for all cases studied, the highest PCTs obtained occurred at 4 s after the initiation of the LOCAs. The reason for the occurrence of these PCTs was the internal pump trip, allowing the pump to coast down and the pump to reverse. The next PCTs, resulted from the LOCA, were observed to occur only for the LOCA cases with feedwater line breaks. It did not happen for the cases with steam-line breaks. (authors)

  10. Analytical and experimental assessment of TVS-2006 fuel assembly thermal-mechanical shape deformation at temperature modeling of a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Afanasiev, A.; Semishkin, V.; Makarov, V.; Matvienko, I.; Puzanov, D.

    2015-01-01

    Full or partial core drying-out takes place in loss-of-coolant accidents, which leads to worsening of heat removal from the fuel rods. Depending on the accident scenario the fuel rod cladding temperature can be in a wide range from 350 to 1200°C. It is worth mentioning, that the length of the process can considerably affect the fuel rod cladding loadcarrying capacity and the FA structure as a whole, and in the long run it defines the radiation consequences of the accident and the possibility of postaccident core disassembly at low cost. Most experiments staged of late were devoted to a study of FA behaviour in the temperature range 800-900°C of α→β phase transition that is characterized by a sharp increase in the rate of zirconium alloy creep which leads to fuel rod cladding ballooning and loss of their tightness within a short period of time. The 600-700°C temperature range turned out to be less investigated whereas this is the range where the change of zirconium alloy mechanical properties is also observed but only with the retention of α-phase. The tests of a full-scale FA dummy with the skeleton of guide tubes and spacer grids connected by friction forces, carried out at the testing facility of JSC OKB “GIDROPRESS”, were devoted to a study of FA behaviour in this temperature range. The model was heated up with hot air to 650°C for 6 hours. The tests ended with fuel rod cladding ballooning due to gauge pressure and shape deformation. No loss of fuel rod cladding integrity was observed. Therefore, a conclusion can be made that a long-time core holdup at the parameters implemented at the test facility is permitted and the deformations of the FA structure do not lead to the damage that could considerably complicate the core disassembly. The test results were used for the verification of the calculational model of FA TVS-2006 structure with a welded skeleton by ANSYS code. On the basis of the verified calculational model a calculational model was

  11. Vent clearing during a simulated loss-of-coolant accident in a Mark I boiling-water reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    In this test series, drywell pressurization rate, drywell overpressure, downcomer submergence, and overall vent system loss coefficient were varied to quantify the primary load sensitivities in the pressure suppression system. Extensive tests were conducted on a unique three-dimensional 1/5 scale model of the pressure suppression system a MARK-I BWR. They were focused on the initial or air cleaning phase of a hypothetical loss of coolant accident. As a result of the complete measurement system employed including multiple high speed cameras, the logical interrelationship between measured forces, measured pressures, and the hydrodynamic phenomena observed in high speed photographic pictures were established. The quantitative values from the 1/5 scale experiments can be applied to full scale plants using established scaling laws. (author)

  12. A study on the effect of fluidic device installed in a safety injection tank on thermal-hydraulic phenomena of large break loss of coolant accident

    International Nuclear Information System (INIS)

    Chung, Young Jong; Bae, Kyoo Hwan; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

    1999-03-01

    The performance of the Safety Injection Tank (SIT) with fluidic device (advanced SIT) is analyzed for the large break loss of coolant accident (LBLOCA) using RELAP5/MOD3.1-KREM. First the case is analyzed using the conventional SIT. Among various cases the case with 4-split downcomer, discharge coefficient Cd=0.6, MCP trip with reactor trip and break location of cold leg discharge side with the pressurizer is found to be the most limiting case. For the same condition, the advanced SIT results the similar PCT, however it can maintain adequately the liquid level in the downcomer. By changing the ECCS location from the current injection to the cold leg elevations, PCT is improved by 75 K. (Author). 6 refs., 4 tabs., 54 figs

  13. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  14. Cooling of safety rods in the Savannah River K Reactor during the gamma heating phase of a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Unal, C.; Motley, F.E.; Rodriguez, S.B.

    1992-01-01

    This paper documents the heat-transfer analysis for the safety rod placed in a perforated guide tube during the gamma heating phase of a large-break loss of coolant accident in Savannah River K-reactor. The cooling mechanisms are natural convection to air and radiation to the surrounding structures. The limiting component is the guide tube. The guide tube is shown to remain coolable below its thermal limit for the anticipated reactor powers unless it is contacted by the hotter safety rod. Sample calculations are performed for various contact scenarios, and the results are reported within the paper. The results indicate that the most limiting contact scenario results when the safety rod heats up to its maximum temperature while remaining concentric in the guide tube and then contacts the guide tube. The worse contact location appears to be in line with the slugs-cladding contact and in between the rows of holes in the guide tube

  15. Results of Semiscale Mod-2C small-break (5%) loss-of-coolant accident. Experiments S-LH-1 and S-LH-2

    International Nuclear Information System (INIS)

    Loomis, G.G.; Streit, J.E.

    1985-11-01

    Two experiments simulating small break (5%) loss-of-coolant accidents (5% SBLOCAs) were performed in the Semiscale Mod-2C facility. These experiments were identical except for downcomer-to-upper-head bypass flow (0.9% in Experiment S-LH-1 and 3.0% in Experiment S-LH-2) and were performed at high pressure and temperature [15.6 MPa (2262 psia) system pressure; 37 K (67 0 F) core differential temperature; 595 K(610 0 F) hot leg fluid temperature]. From the experimental results, the signature response and transient mass distribution are determined for a 5% SBLOCA. The core thermal-hydraulic response is characterized, including core void distribution maps, and the effect of core bypass flow on transient severity is assessed. Comparisons are made between postexperiment RELAP5 calculations and the experimental results, and the capability of RELAP5 to calculate the phenomena is assessed. 115 figs

  16. THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

    International Nuclear Information System (INIS)

    Muramatsu, Ken; Akimoto, Masayuki

    1982-08-01

    THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

  17. A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gaballah, I.

    1978-09-01

    A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident. A theory was developed for the calculation of a dispersed two phase flow with heat addition in a channel with general area change. The theory was used to study different thermodynamic and gasdynamic processes, which may occur during the emergency cooling after a LOCA of a pressurized water reactor. The basic equations were formulated and solved numerically. The heat transfer mechanism was examined. Calculations have indicated that the radiative heat flux component is small compared to the convective component. A drop size spectrum was used in the calculations. Its effect on the heat transfer was investigated. It was found that the calculation with a mean drop diameter gives good results. Significant thermal non-equilibrium has been evaluated. The effect of different operating parameters on the degree of thermal non-equilibrium was studied. The flow and heat transfer in a channel with cross-sectional area change were calculated. It was shown that the channel deformation affects the state properties and the heat transfer along the channel very strongly. (orig.) 891 GL [de

  18. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  19. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    International Nuclear Information System (INIS)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

    2017-01-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  20. Stresses and strains in the steel containment resulting from transient pressure and temperature loading during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gruner, P.; Kuntze, W.M.; Jansky, J.

    1985-01-01

    Posttest calculations of stresses and strains in the steel containment of the German research reactor HDR were performed for a simulated LOCA. The results of the theoretical investigations are presented and compared to experimental findings. The pressure and temperature loading of the shell was determined with the thermodynamic code COFLOW on the basis of a multi-compartment model. Using a three-dimensional finite element model the temporal behaviour of the containment was calculated employing the structural mechanics code ASKA. Global bending deformations and local negative straining of the steel shell is discussed. Theoretical and experimental results agree in most cases rather well. Reasons for deviations will be discussed. The specific behaviour of strains found in the vicinity of locally heated areas will be explained by means of analytical considerations. (orig.)

  1. Investigations of the reflood-phase after a loss-of-coolant-accident of an advanced pressurized water reactor (APWR)

    International Nuclear Information System (INIS)

    Schumann, S.; Oldekop, W.

    1983-01-01

    Differences between a high converting advanced pressurized-water reactor (APWR) and a conventional PWR, which are relevant to the reflood-phase after LOCA are presented. The used code and its verification by PWR-reflood experiments is explained. Comparative calculations for APWR and PWR with several conservative assumptions for example cold-leg-injection only, yield nearly the same maximum midplane-temperatures for the average-channel. For the APWR, however, the upper half of the rod shows higher temperatures. Quenchfront and core-water-level increase more slowly. The differences in the reflood-thermohydraulics are analysed in detail. A conservative hot-channel calculation shows maximum temperatures of about 920 0 C. Finally the influence of conservative assumptions is described and the necessity of experiments pointed out. (orig.)

  2. Analysis of the loss of coolant accident due to the faiture in the open position of two pressurizer relief valves, for Angra-1 nuclear power plant

    International Nuclear Information System (INIS)

    Freire, C.F.

    1981-06-01

    A study of the modeling techniques adequate for simulating the loss of coolant accident caused by stuck open pressurizer relief valves, using the RELAP4-MOD5 code, is performed and the model developed is applied to the analysis of this kind of accident for the Central Nuclear Almirante Alvaro Alberto Unit (Angra 1). The thermal hydraulic behavior of the reactor cooling system, when subjected to a loss of main feedwater followed by the failure in the open position of two pressurizer relief valves, is determined. The relief valves are assumed to fail in the totally open position, delivering the maximum massflow through the discharge line. The RELAP4-MOD5 code is shown to be adequate for this kind of analysis, and the detailed prediction of the thermal hydraulic behavior of the Reactor Coolant System is thus possible. The eficiency of the emergency core cooling system of Angra 1 is demonstrated, the fuel elements remaining covered by the coolant during all the accident, and the peak clad temperatures are kept within design limites, ensuring the integrity of the core. (Author) [pt

  3. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

  4. Analysis of three ex-vessel loss-of-coolant accidents in the first wall cooling system of NET/ITER

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1993-01-01

    An ex-vessel LOCA may be caused by a rupture of a cooling pipe located outside the vacuum vessel. No plasma shutdown and no other counteractions have been assumed in order to study the worst case conditions of the accidents. The next three ex-vessel LOCAs in the primary cooling system of the first wall have been analysed: 1. a large break ex-vessel LOCA caused by a rupture of the cold leg (inner diameter 0.314 m) of the main circuit; 2. an intermediate break ex-vessel LOCA caused by a rupture of a sector inlet feeder (inner diameter 0.158 m); 3. an intermediate break ex-vessel LOCA caused by a rupture of the surge line (inner diameter 0.180 m) of the pressurizer. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the first two scenarios, melting in the first wall starts about 90 s after break initiation. In the third scenario, melting in the first wall start about 323 s after break initiation. Special emphasis has been paid to the characteristics of the break flows, the transient thermal-hydraulic behaviour of the cooling system, and the temperature development in the first wall. (orig.)

  5. Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kanzleiter, T.

    1975-10-01

    The phenomena occuring within a containment during a LOCA are currently investigated through experiments with a modelcontainment at Battelle-Institut Frankfurt on behalf of the Bundesministerium fuer Forschung und Technologie, Bonn. The experimental results are to be compared with the results of model calculations in order to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model-containment. The model-containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross section. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiment a PWR-configuration with nine compartments has been istalled. The model scale of the compartment volumes and the overflow areas are about 1:64 compared to the 1,200-MW-PWR-plant Biblis A. Later investigations will also include BWR-experiments and experiments leading to an extremely high load on special containment structures. (orig.) [de

  6. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    International Nuclear Information System (INIS)

    Ghan, L.S.; Ortiz, M.G.

    1991-01-01

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B ampersand W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B ampersand W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions

  7. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

    2017-08-15

    The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  8. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  9. Quantifying reactor safety margins: Application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

    International Nuclear Information System (INIS)

    Boyack, B.; Duffey, R.; Wilson, G.; Griffith, P.; Lellouche, G.; Levy, S.; Rohatgi, U.; Wulff, W.; Zuber, N.

    1989-12-01

    The US Nuclear Regulatory Commission (NRC) has issued a revised rule for loss-of-coolant accident/emergency core cooling system (ECCS) analysis of light water reactors to allow the use of best-estimate computer codes in safety analysis as an option. A key feature of this option requires the licensee to quantify the uncertainty of the calculations and include that uncertainty when comparing the calculated results with acceptance limits provided in 10 CFR Part 50. To support the revised ECCS rule and illustrate its application, the NRC and its contractors and consultants have developed and demonstrated an uncertainty evaluation methodology called code scaling, applicability, and uncertainty (CSAU). The CSAU methodology and an example application described in this report demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. The methodology is systematic and comprehensive as it addresses and integrates the scenario, experiments, code, and plant to resolve questions concerned with: (a) code capability to scale-up processes from test facility to full-scale nuclear power plants; (b) code applicability to safety studies of a postulated accident scenario in a specified nuclear power plant; and (c) quantifying uncertainties of calculated results. 127 refs., 55 figs., 40 tabs

  10. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-01-01

    To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method

  11. Hydride precipitation, fracture and plasticity mechanisms in pure zirconium and Zircaloy-4 at temperatures typical for the postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Pshenichnikov, Anton; Stuckert, Juri; Walter, Mario

    2016-01-01

    Highlights: • All δ-hydrides in Zr and Zircaloy-4 have basal or pyramidal types of habit planes. • Seven orientation relationships for δ-hydrides in Zr matrix were detected. • Decohesion fracture mechanism of hydrogenated Zr was investigated by fractography. - Abstract: The results of investigations of samples of zirconium and its alloy Zircaloy-4, hydrogenated at temperatures 900–1200 K (typical temperatures for loss-of-coolant accidents) are presented. The analyses, based on a range of complementary techniques (X-ray diffraction, scanning electron microscopy, electron backscatter diffraction) reveals the direct interrelation of internal structure transformation and hydride distribution with the degradation of mechanical properties. Formation of small-scale zirconium hydrides and their bulk distribution in zirconium and Zircaloy-4 were investigated. Fractographical analysis was performed on the ruptured samples tested in a tensile machine at room temperature. The already-known hydrogen embrittlement mechanisms based on hydride formation and hydrogen-enhanced decohesion and the applicability of them in the case of zirconium and its alloys is discussed.

  12. Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors

    International Nuclear Information System (INIS)

    1993-07-01

    An interfacing systems loss of coolant accident (ISLOCA) involves failure or improper operation of pressure isolation valves (PIVs) that compose the boundary between the reactor coolant system and low-pressure rated systems. Some ISLOCAs can bypass containment and result in direct release of fission products to the environment. A cost/benefit evaluation, using three PWR analyses, calculated the benefit of two potential modifications to the plants. Alternative 1 is improved plant operations to optimize the operator's performance and reduce human error probabilities. Alternative 2 adds pressure sensing devices, cabling, and instrumentation between two PIVs to provide operators with continuous monitoring of the first PIV. These two alternatives were evaluated for the base case plants (Case 1) and for each plant, assuming the plants had a particular auxiliary building design in which severe flooding would be a problem if an ISLOCA occurred. The auxiliary building design (Case 2) was selected from a survey that revealed a number of designs with features that provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The results were judged not to provide sufficient basis for generic requirements. It was concluded that the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs)

  13. Simulation of the IAEA's fourth Standard Problem Exercise small-break loss-of-coolant accident using RELAP5/MOD.3.1

    International Nuclear Information System (INIS)

    Cebull, P.P.; Hassan, Y.A.

    1995-01-01

    A small-break loss-of-coolant accident experiment conducted at the PMK-2 integral test facility in Hungary is analyzed using the RELAP5/MOD3.1 thermal-hydraulic code. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). Blind calculations of the exercise are presented, and the timing of various events throughout the transient is discussed. A posttest analysis is performed in which the sensitivity of the calculated results is investigated. The code RELAP5 predicts most of the transient events well, although a few problems are noted, particularly the failure of RELAP5 to predict dryout in the core even through the collapsed liquid level fell below the top of the heated portion. A discrepancy between the predicted primary mass inventory distribution and the experimental data is identified. Finally, the primary and secondary pressures calculated by RELAP5 fell too rapidly during the latter part of the transient, resulting in rather large errors in the predicted timing of some pressure-actuated events

  14. Risk-informed analysis of the large break loss of coolant accident and PCT margin evaluation with the RISMC methodology

    International Nuclear Information System (INIS)

    Liang, T.H.; Liang, K.S.; Cheng, C.K.; Pei, B.S.; Patelli, E.

    2016-01-01

    Highlights: • With RISMC methodology, both aleatory and epistemic uncertainties have been considered. • 14 probabilistically significant sequences have been identified and quantified. • A load spectrum for LBLOCA has been conducted with CPCT and SP of each dominant sequence. • Comparing to deterministic methodologies, the risk-informed PCT margin can be greater by 44–62 K. • The SP of the referred sequence to cover 99% in the load spectrum is only 5.07 * 10 −3 . • The occurrence probability of the deterministic licensing sequence is 5.46 * 10 −5 . - Abstract: For general design basis accidents, such as SBLOCA and LBLOCA, the traditional deterministic safety analysis methodologies are always applied to analyze events based on a so called surrogate or licensing sequence, without considering how low this sequence occurrence probability is. In the to-be-issued 10 CFR 50.46a, the LBLOCA will be categorized as accidents beyond design basis and the PCT margin shall be evaluated in a risk-informed manner. According to the risk-informed safety margin characterization (RISMC) methodology, a process has been suggested to evaluate the risk-informed PCT margin. Following the RISMC methodology, a load spectrum of PCT for LBLOCA has been generated for the Taiwan’s Maanshan Nuclear Power plant and 14 probabilistic significant sequences have been identified. It was observed in the load spectrum that the conditional PCT generally ascends with the descending sequence occurrence probability. With the load spectrum covering both aleatory and epistemic uncertainties, the risk-informed PCT margin can be evaluated by either expecting value estimation method or sequence probability coverage method. It was found that by comparing with the traditional deterministic methodology, the PCT margin evaluated by the RISMC methodology can be greater by 44–62 K. Besides, to have a cumulated occurrence probability over 99% in the load spectrum, the occurrence probability of the

  15. Risk-informed analysis of the large break loss of coolant accident and PCT margin evaluation with the RISMC methodology

    Energy Technology Data Exchange (ETDEWEB)

    Liang, T.H. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); 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); Cheng, C.K.; Pei, B.S. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Patelli, E. [Institute of Risk and Uncertainty, University of Liverpool, Room 610, Brodie Tower, L69 3GQ (United Kingdom)

    2016-11-15

    Highlights: • With RISMC methodology, both aleatory and epistemic uncertainties have been considered. • 14 probabilistically significant sequences have been identified and quantified. • A load spectrum for LBLOCA has been conducted with CPCT and SP of each dominant sequence. • Comparing to deterministic methodologies, the risk-informed PCT margin can be greater by 44–62 K. • The SP of the referred sequence to cover 99% in the load spectrum is only 5.07 * 10{sup −3}. • The occurrence probability of the deterministic licensing sequence is 5.46 * 10{sup −5}. - Abstract: For general design basis accidents, such as SBLOCA and LBLOCA, the traditional deterministic safety analysis methodologies are always applied to analyze events based on a so called surrogate or licensing sequence, without considering how low this sequence occurrence probability is. In the to-be-issued 10 CFR 50.46a, the LBLOCA will be categorized as accidents beyond design basis and the PCT margin shall be evaluated in a risk-informed manner. According to the risk-informed safety margin characterization (RISMC) methodology, a process has been suggested to evaluate the risk-informed PCT margin. Following the RISMC methodology, a load spectrum of PCT for LBLOCA has been generated for the Taiwan’s Maanshan Nuclear Power plant and 14 probabilistic significant sequences have been identified. It was observed in the load spectrum that the conditional PCT generally ascends with the descending sequence occurrence probability. With the load spectrum covering both aleatory and epistemic uncertainties, the risk-informed PCT margin can be evaluated by either expecting value estimation method or sequence probability coverage method. It was found that by comparing with the traditional deterministic methodology, the PCT margin evaluated by the RISMC methodology can be greater by 44–62 K. Besides, to have a cumulated occurrence probability over 99% in the load spectrum, the occurrence probability

  16. Consideration on hydrogen explosion scenario in APR 1400 containment building during small breakup loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kweonha, E-mail: khpark@kmou.ac.kr [Division of Mechanical & Energy Systems Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of); Khor, Chong Lee, E-mail: itachi_829@hotmail.com [Department of Mechanical Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of)

    2015-11-15

    Highlights: • Hydrogen behavior in the containment building of APR1400 nuclear plant up to 15 h after the failure happened. • The risk of hydrogen explosion largely depends on the combination of air, hydrogen and steam in the containment. • Hydrogen explosion risk at different locations in the containment was analyzed. - Abstract: This paper describes the analytical result of the potential risk of hydrogen gas up to 15 h after the failure takes place. The major cause of the disaster occurred in Fukushima Daiichi nuclear reactor was the detonation of accumulated hydrogen in the containment by highly increased reactor core temperatures after the failure of the emergency cooling system. The hydrogen risk should be considered in severe accident strategies in current and future NPPs. A hydrogen explosion scenario is proposed. Hydrogen is accumulated on top of the dome during the hydrogen release period. At this point, there are no risk of explosion due to the steam that resides in upper part of the dome. As the hydrogen concentration increase, substantial amount of steams are released. Subsequently, hydrogen is forced into the lower part of the building with high air density—small explosion and dormant steam condensation phase are possible. The light hydrogen rises up slowly with air, gathering on top of the building with high air density. Massive hydrogen explosion is anticipated upon ignition at this stage.

  17. Evaluation method of iodine re-evolution from an in-containment water pool after a loss of coolant accident, Part II: Evaluation of pH and iodine re-evolution

    International Nuclear Information System (INIS)

    Kim, Tae Hyeon; Jeong, Ji Hwan

    2016-01-01

    Highlights: • It is required to evaluate re-evolved iodine from sump water after LOCA. • Transport of iodine and chemicals influencing pH were analyzed using CFD. • Chemical conditions of the iodine-rich region suppress iodine re-evolution. • The current evaluation method for I 2 re-evolution is excessively conservative. - Abstract: Radioactive iodine that is released during a postulated loss of coolant accident is dissolved into the containment spray water and transported into the in-containment refueling water storage tank (IRWST). The re-evolution of iodine from the water is a safety concern. In this study, three-dimensional computational fluid dynamics (CFD) analyses are conducted in order to analyze the transport of chemical species including iodine in the IRWST and to calculate the amount of iodine that re-evolves from the IRWST water. The CFD analyses demonstrate that the pH of water is high where the iodine concentration is high. Considering that the creation rate of molecular iodine declines as the pH increases, it can be understood that the iodine re-evolution is not so strong in practical situations because the chemical conditions of the iodine-rich region suppress the re-evolution of the iodine. In addition, four different methods for evaluating the amount of re-evolved iodine are examined. The amount of re-evolved iodine calculated using the total-volume-average values, which are currently used for safety analyses, appear to be significantly higher than those determined using other methods. The amount of re-evolved iodine estimated using a realistic method with a conservative assumption of volatilization appears to be approximately one thousandth of that evaluated using the current method. This implies that the current method is very conservative.

  18. Investigation of loss of coolant accidents in pressurized water reactors using the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method for considering of uncertainties in TRACE

    International Nuclear Information System (INIS)

    Sporn, Michael; Hurtado, Antonio

    2016-01-01

    Loss of coolant accident must take uncertainties with potentially strong effects on the accident sequence prediction into account. For example, uncertainties in computational model input parameters resulting from varying geometry and material data due to manufacturing tolerances or unavailable measurements should be considered. The uncertainties of physical models used by the software program are also significant. In this paper, use of the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method to quantify the uncertainties in the TRACE thermal-hydraulic program is demonstrated. For demonstration purposes loss of coolant accidents with breaks of various types and sizes in a DN 700 reactor coolant pipe are used as an example Application.

  19. THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

    1988-04-01

    THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

  20. Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

    International Nuclear Information System (INIS)

    Sdouz, Gert

    2006-01-01

    The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

  1. Fuel Behaviour at High During RIA and LOCA Accidents

    International Nuclear Information System (INIS)

    Barrio del Juanes, M. T.; Garcia Cuesta, J. C.; Vallejo Diaz, I.; Herranz Puebla

    2001-01-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs

  2. Calculation of thermoelastic stresses in the rewetting region of the fuel rod cladding during a loss of coolant accident (loca)

    International Nuclear Information System (INIS)

    Roberty, N.C.; Carmo, E.G.D. do; Tanajura, C.A.S.

    1982-01-01

    A one-dimensional model for axial distribution calculation of temperature and thermal stresses in the fuel rod cladding for a Pressurized Water Reactors (PWR) is developed. The effect of the coolant inlet temperaure, the Leidenfrost and the nucleate boiling in the stress distribution are evaluated. A perturbation in the cladding stress state is obtained. (E.G.) [pt

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

    International Nuclear Information System (INIS)

    Costa, J.R.

    1978-12-01

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

  4. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M T; Garcia Cuesta, J C; Vallejo Diaz, I; Puebla, Herranz

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  5. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  6. Dilatational behaviour of ZrNb1 fuel cans of a WWER-type reactor during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Stephan, M.; Wetzel, L.

    1987-01-01

    Based on an assessment of various factors of influence on the performance of fuel cans during normal operation and imaginable accidents, the necessity of studying creep and burst behaviour of WWER-type fuel cans of ZrNb1 under simulated LOCA conditions has been proved and an experimental facility designed for this purpose is described. Control of fuel can temperature is accomplished through a minicomputer during the creep and bursts experiments. With this, various temperature loading profiles of the fuel cans can be realized. Experimental results on dilatational behaviour of ZrNb1 fuel cans from isothermal creep and burst experiments in air are presented and compared with values for Zircaloy. (author)

  7. Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

    1980-01-01

    A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

  8. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

  9. Determining the boron concentration during long-term cooling of the reactor core after large loss of coolant accident; Dolocenje koncentracij bora pri dolgotrajnoem hladjenju sredice po veliki izlivni nezgodi

    Energy Technology Data Exchange (ETDEWEB)

    Mavko, B; Ravnki, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1988-07-01

    Critical boron concentration before and after postulated loss of coolant accident with long-term cooling recirculation was calculated for cycle 6 of Krsko NPP. The limiting boron concentration curve of containment sump was calculated for equilibrium conditions. The results were analysed and showed that the boron concentration in refueling water storage tank and in safety injection accumulators should be increased from 2000 to 2100 ppm in 6th cycle. In the consequence corresponding chapters of the NPP Krsko technical Specifications were changed as well. (author)

  10. Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm2 in the cold leg of primary loop using RELAP5 code

    International Nuclear Information System (INIS)

    Borges, Eduardo M.; Sabundjian, Gaiane

    2017-01-01

    The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm 2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

  11. Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm{sup 2} in the cold leg of primary loop using RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: borges.em@hotmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm{sup 2} of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

  12. MAAP-CANDU simulations of LOCA/LOECI accidents at Darlington NGS

    International Nuclear Information System (INIS)

    Kwee, M.T.; Choi, M.H.; Leung, R.K.

    1996-01-01

    Severe accidents have been the subject of a great deal of analysis and research, particularly in the light water reactor community. Although severe accident analysis in Canada deuterium-uranium (CANDU) reactors has not been published abundantly, a significant body of research and analysis has been accumulated. This has occurred because CANDU has directly taken into consideration a set of severe accidents [e.g loss-of-coolant accidents (LOCAs) coincident with a loss-of-emergency-coolant injection (LOECI)] in the design basis. These accidents have served to define the design requirements that ensure the integrity of the heat transport system. The CANDU reactor design has inherent heat sinks such as the primary heat transport system, the secondary side, moderator system, and shielding system (shield tank and end shields). These heat sinks are significant and are able to moderate or terminate the progression of severe accidents that go beyond the design base cases. These types of accidents are typically analyzed at Ontario Hydro in conjunction with probabilistic safety analysis (PSA), where the severe accident consequences are analyzed by a series of conservative hand-calculation methods

  13. Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

    Science.gov (United States)

    Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

    1972-01-01

    A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

  14. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  15. Comparison of the cladding deformation measured during the Power Burst Facility loss-of-coolant accident in-pile experiments with recent Oak Ridge National Laboratory out-of-pile results

    International Nuclear Information System (INIS)

    Broughton, J.M.; McCardell, R.K.; MacDonald, P.E.

    1981-01-01

    A series of four large break loss-of-coolant accident fuel behavior experiments have been performed in the Power Burst Facility. The results of these experiments are briefly reviewed and compared with results from the ORNL multirod burst test program. The effect of cladding burst temperature and prior irradiation were investigated. The cladding strain of the previously irradiated test rods was more uniformly distributed around the cladding circumference and larger than for similar unirradiated test rods. The ORNL out-of-pile single rod test results are in good agreement with the Power Burst Facility (PBF) test results with unirradiated test rods, and the ORNL out-of-pile, single-rod test results with heated shrouds and the PBF test results with previously irradiated test rods are comparable

  16. Experimental analysis of upward vertical two-phase flow in four-cusp channels simulating the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident

    International Nuclear Information System (INIS)

    Assad, A.C.A.

    1984-01-01

    The present work deals with an experimental analysis of upward vertical two-phase flow in channels with circular and four-cusp cross-sections. The latter simulates the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident. Simultaneous flow of air and water has been employed to simulate adiabatic steam-water flow. The installation of air-water separators helped eliminate instabilities during pressure-drop measurements. The gamma ray attenuation was utilized for the void fraction determination. For the four-cusp geommetry, new criteria for two-phase flow regime transitions have been determined, as well as new correlatins for pressure drop and void fraction, as function of the Lockhart-Martinelli factor and vapour mass-fraction, respectively. (Author) [pt

  17. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  18. Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

    International Nuclear Information System (INIS)

    Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

    2010-01-01

    SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

  19. A LOCA analysis for AHWR caused by ECCS header rupture

    International Nuclear Information System (INIS)

    Chatterjee, B.; Gawai, Amol; Gupta, S.K.; Kushwaha, H.S.

    2000-01-01

    Loss of coolant accident (LOCA) analyses for the proposed 750 MWth Advanced Heavy Water Reactor (AHWR), initiated by the rupture of 8 inch NB ECCS header has been carried out. This paper narrates the description of AHWR and associated ECCS, postulated scenario with which the analyses is carried out, results, discussion and conclusion

  20. Plan for Structural Analysis of Fuel Assembly for Seismic and Loss of Coolant Accident Loading Considering End-Of-Life Condition for APR1400 NRC Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Hak [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The evaluation of fuel assembly structural response to externally applied forces by earthquakes and postulated pipe breaks in the reactor coolant system is described in standard review plan (SRP) 4.2, appendix A. SRP 4.2, appendix A, section III, states, 'While P(crit) [the crushing load] will increase with irradiation, ductility will be reduced. The extra margin in P(crit) for irradiated spacer grids is thus assumed to offset the unknown deformation behavior of irradiated spacer grids beyond P(crit).' The assumption in the SRP concerning irradiated grids may suggest that only the beginning-of-life (BOL) condition for spacer grid strength needs to be evaluated for fuel assembly integrity under externally applied forces. However, U.S. NRC issued the NRC. To consider the EOL conditions for the structural analysis of the fuel assembly under a seismic and LOCA loading, the simulated fuel assembly for EOL conditions should be considered by determining the gap between the spacer grid and fuel rod. Using the simulated fuel assembly, spacer grid test and fuel assembly mechanical test should be conducted to determine the simplified model of fuel assembly which is used for the structural analysis. The structural analysis will be conducted using the fuel assembly model for EOL condition. The flow damping value will be also used for the structural analysis to reduce the impact force.

  1. Modelling of the steam-water-countercurrent flow in the rewetting and flooding phase after loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Curca-Tivig, F.

    1990-01-01

    A new interphase momentum exchange model has been developed to simulate the Refill- Reflood Phase after LOCAs. Special phenomena of steam/water- countercurrent flow - like limitation or onset of downward-watee penetration - have been modelled and integrated into a flooding model. The interphase momentum exchange model interconnected with the flooding model has been implemented into the advanced system code RELAP5/MOD1. The new version of this code can now be utilized to predict the hot leg emergency-core-cooling (ECC) injection for German PWRs. The interfacial momentum transfer model developed includes the interphase frictional drag, the force due to virtual mass and the momenta due to interphase mass transfer. The modelling of the interfacial shear or drag accounts for the effects of phase and velocity profiles. The flooding model predicts countercurrent-flow limitation, onset of water penetration and partial delivery. The flooding correlation specifies the maximum down flow liquid velocity in case of countercurrent flow through flow restrictions for a given vapor velocity. (orig./HP) [de

  2. Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Larson, T.K.

    1978-01-01

    Examination of the U-tubes in the steam generators of some large commercial pressurized water reactors (PWR) has revealed the existence of leakage and in some cases structural weakening of the tubes. This structural weakening enhances the possibility of tubes rupturing during a hypothesized loss-of-coolant accident (LOCA). Considerable interest has been shown in the analysis of tube ruptures concurrent with a hypothesized LOCA since the presence of tube ruptures has the potential to influence the system thermal-hydraulic response and could foreseeably result in a more severe core thermal behavior than might otherwise occur. To experimentally investigate the influence of steam generator tube ruptures on the thermal-hydraulic response of PWR type system, a series of experiments was conducted in the Semiscale Mod-1 system by EG and G Idaho, Inc., for the U.S. Nuclear Regulatory Commission and the Department of Energy. The primary objective of the experiments was to obtain data which could be used to evaluate the influence of the simulated tube ruptures on the system and core thermal-hydraulic response for a range of tube ruptures that was expected to provide the potential for high cladding temperatures in the Semiscale facility. The experiments were conducted assuming a variety in the number of tubes ruptured during large break loss-of-coolant conditions. The number of experiments conducted permitted determination of the range of tube ruptures for which high peak cladding temperatures could result in the Semiscale Mod-1 system. The paper contains a description of the Semiscale Mod-1 system and a discussion of the steam generator tube rupture tests conducted. The experimental results from the test series and the thermal-hydraulic phenomena found to influence the core thermal response during the experiments are discussed

  3. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

  4. Hydrogen Safety Analysis of the OPR1000 Nuclear Power Plant during a Severe Accident by a Small-Break Loss of Coolant

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Park, Soo Yong; Ha, Kwang Soon; Hong, Seong Wan; Kim, Sang Baik

    2009-01-01

    A huge amount of hydrogen can be generated in a nuclear reactor and released into the reactor containment if a hypothetical severe accident happens. Even for the accident, the hydrogen concentrations must be safely controlled. In order to prove a nuclear power plant (NPP) safe from hydrogen, a simulation of hydrogen distributions in the containment are usually conducted by using a 1-dimensional thermo-hydraulic system code. If there exists a possibility of a hydrogen explosion in the containment, it is required to install a hydrogen mitigation system such as igniters or hydrogen recombiner. For a licensing of NPP construction and operation, the hydrogen combustion and hydrogen mitigation system in the containment is one of the important safety issues. In Korea, two OPR1000 NPPs by the name of Shin-Wolsung 1 and 2 are under construction. The hydrogen safety and its control for the new NPPs will be evaluated in detail until a licensing of the operation. Until now, simulations of the hydrogen behaviors in the OPR1000 have been conducted by a lumped method for each compartment in the containment using CONTAIN or MAAP. This 1-dimensional method is very efficient for a long-term simulation of an accident because of its fast running time, and it is very effective for establishing the averaged hydrogen concentrations in each compartment. But a 3-dimensional flow structure developed by a discharged mass from a reactor vessel and local concentrations of hydrogen are difficult to be resolved by the lumped method. In this study, hydrogen distributions and characteristics of hydrogen mixture cloud such as a possibility of flame acceleration in each compartment of OPR1000 containment were evaluated by using GASFLOW code

  5. Verification tests for GRAD, a computer program to predict nonuniform deformation and failure of Zr-2.5 wt percent Nb pressure tubes during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.; Godin, D.P.

    1985-03-01

    During a postulated loss-of-coolant accident in a CANDU reactor, the temperature of the pressure tubes could rise sufficiently so that ballooning could occur. It is also likely that there would be a variation in temperature around the tube circumference, causing the deformation to be nonuniform. Since the deformation of the pressure tube controls how the core heat is transferred to the surrounding moderator, which is a large heat sink, a computer program, GRAD, has been developed to predict this nonuniform deformation. Numerous biaxial creep tests were done, where the temperature of internally pressurized sections of Zr-2.5 wt percent Nb pressure tubes were ramped to check the ability of GRAD to predict the resulting nonuniform deformation and possible tube failure. GRAD was successful in predicting the average transverse creep strain observed during the tests and the local transverse creep strain at the end of the tests. GRAD was also able to predict the failure time and average transverse creep strain at failure for all the specimens that failed

  6. Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A. [Argonne National Lab., IL (United States); Voeroess, L.; Techy, Z. [VEIKI Inst. for Electric Power Research, Budapest (Hungary); Katona, T. [Paks Nuclear Power Plant (Hungary)

    1998-07-01

    A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

  7. CHEMICAL EFFECTS ON PWR SUMP STRAINER BLOCKAGE AFTER A LOSS-OF-COOLANT ACCIDENT: REVIEW ON U.S. RESEARCH EFFORTS

    Directory of Open Access Journals (Sweden)

    CHI BUM BAHN

    2013-06-01

    Full Text Available Industry- or regulatory-sponsored research activities on the resolution of Generic Safety Issue (GSI-191 were reviewed, especially on the chemical effects. Potential chemical effects on the head loss across the debris-loaded sump strainer under a post-accident condition were experimentally evidenced by small-scale bench tests, integrated chemical effects test (ICET, and vertical loop head loss tests. Three main chemical precipitates were identified by WCAP-16530-NP: calcium phosphate, aluminum oxyhydroxide, and sodium aluminum silicate. The former two precipitates were also identified as major chemical precipitates by the ICETs. The assumption that all released calcium would form precipitates is reasonable. CalSil insulation needs to be minimized especially in a plant using trisodium phosphate buffer. The assumption that all released aluminum would form precipitates appears highly conservative because ICETs and other studies suggest substantial solubility of aluminum at high temperature and inhibition of aluminum corrosion by silicate or phosphate. The industry-proposed chemical surrogates are quite effective in increasing the head loss across the debris-loaded bed and more effective than the prototypical aluminum hydroxide precipitates generated by in-situ aluminum corrosion. There appears to be some unresolved potential issues related to GSI-191 chemical effects as identified in NUREG/CR-6988. The United States Nuclear Regulatory Commission, however, concluded that the implications of these issues are either not generically significant or are appropriately addressed, although several issues associated with downstream in-vessel effects remain.

  8. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

    Kawasaki, S.; Uetsuka, H.; Furuta, T.

    1983-01-01

    In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

  9. Comparison of the Aerospace Systems Test Reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1983-01-01

    This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 10 4 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code

  10. Comparison of the aerospace systems test reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1984-01-01

    This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 10 4 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code. (author)

  11. Application of thermal hydraulic and severe accident code SOCRAT/V3 to bottom water reflood experiment QUENCH-LOCA-0

    International Nuclear Information System (INIS)

    Vasiliev, A.D.; Stuckert, J.

    2013-01-01

    Highlights: ► QLOCA-0 test simulates a design basis LOCA NPP accident with maximum temperature 1300 K. ► Deep understanding of hydraulics and thermal mechanics under accident conditions is necessary. ► We model the test QLOCA-0 with bottom flooding using the Russian code SOCRAT/V3. ► Calculated and experimental data are in a good agreement. ► Experimental procedure is determined to reach a representative LOCA scenario in future tests. -- Abstract: The thermal hydraulic and SFD (severe fuel damage) best estimate computer modeling code SOCRAT/V3 has been used for the calculation of QUENCH-LOCA-0 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (loss of coolant accident) nuclear power plant accident sequence in which the overheated up to 1300 K reactor core would be reflooded from the bottom by ECCS (emergency core cooling system). The first test QUENCH-LOCA-0 was successfully conducted at the KIT, Karlsruhe, Germany, in July 22, 2010, and was performed as the commissioning test for this series. The rod claddings are identical to that used in PWRs. The bundle was electrically heated in steam from 800 K to 1340 K with the heat-up rate of approximately 2.7 K/s. After cooling in the saturated steam the bottom flooding with water flow rate of about 100 g/s was initiated. The SOCRAT calculated results are in a good agreement with experimental data taking into account additional quenching due to water condensate entrainment at the steam cooling stage. SOCRAT/V3 has been used for estimation of further steps in experimental procedure to reach a representative LOCA scenario in future tests

  12. NPP Krsko small break LOCA analysis

    International Nuclear Information System (INIS)

    Mavko, B.; Petelin, S.; Peterlin, G.

    1987-01-01

    Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

  13. Analysis of LOFT loss-of-coolant experiments L2-2, L2-3, and L3-0

    International Nuclear Information System (INIS)

    Leach, L.P.; Linebarger, J.H.

    1979-01-01

    A summary of results from Loss-of-Coolant Experiments (LOCE) L2-2, L2-3, and L3-0, conducted in the Loss-of-Fluid Test (LOFT) facility, and conclusions from posttest analyses of the experimental data are presented. LOCEs L2-2 and L2-3 were nuclear large break experiments and were dominated by a core-wide fuel rod cladding rewet, which limited the maximum fuel temperature. Analytical models only conservatively predicted the measured fuel rod temperatures and will require improvements to provide best estimate predictions in this area. Analysis of a large commercial pressurized water reactor (PWR) indicates that the cladding rewet observed in LOFT is also likely to occur in a large PWR, and that, therefore, safety analysis calculations of large loss-of-coolant accidents (LOCA) are more conservative than previously thought. LOCE L3-0 was an isothermal small break (top of pressurizer) experiment and illustrated that the pressurizer fills after the primary system fluid saturates someplace other than the pressurizer itself, that the indicated pressurizer level is higher than the actual level, and that additional model development and assessment work is necessary in order to predict small LOCAs as accurately as large LOCAs

  14. Application of code scaling applicability and uncertainty methodology to the large break loss of coolant

    International Nuclear Information System (INIS)

    Young, M.Y.; Bajorek, S.M.; Nissley, M.E.

    1998-01-01

    In the late 1980s, after completion of an extensive research program, the United States Nuclear Regulatory Commission (USNRC) amended its regulations (10CFR50.46) to allow the use of realistic physical models to analyze the loss of coolant accident (LOCA) in a light water reactors. Prior to this time, the evaluation of this accident was subject to a prescriptive set of rules (appendix K of the regulations) requiring conservative models and assumptions to be applied simultaneously, leading to very pessimistic estimates of the impact of this accident on the reactor core. The rule change therefore promised to provide significant benefits to owners of power reactors, allowing them to increase output. In response to the rule change, a method called code scaling, applicability and uncertainty (CSAU) was developed to apply realistic methods, while properly taking into account data uncertainty, uncertainty in physical modeling and plant variability. The method was claimed to be structured, traceable, and practical, but was met with some criticism when first demonstrated. In 1996, the USNRC approved a methodology, based on CSAU, developed by a group led by Westinghouse. The lessons learned in this application of CSAU will be summarized. Some of the issues raised concerning the validity and completeness of the CSAU methodology will also be discussed. (orig.)

  15. Special LOFT features for improved monitoring and survival of LOCA transients

    International Nuclear Information System (INIS)

    Goodrich, L.D.; Leach, L.P.; Klingler, T.B.; Morrow, J.C.; Phoenix, W.C.; Satterwhite, D.G.; Sumpter, K.C.; Rouhani, S.Z.; Welland, H.J.

    1980-01-01

    LOFT is designed to monitor and survive Loss-Of-Coolant-Accidents (LOCAs). This report presents the primary design difference from LPWRs that were required to accomplish this. These design differences may be of interest to the nuclear power generator industry. This report should be revised semi-annually or as developments in the LOFT Program require

  16. Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

    International Nuclear Information System (INIS)

    Casal, V.; Malang, S.; Rust, K.

    1982-10-01

    For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

  17. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  18. Estimation of LOCA break size using cascaded Fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA), which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict the progression of LOCAs in nuclear power plants. Therefore, accurate information on the LOCA break position and size should be provided to efficiently manage the accident. In this paper, the LOCA break size is predicted using a cascaded fuzzy neural network (CFNN) model. The input data of the CFNN model are the time-integrated values of each measurement signal for an initial short-time interval after a reactor scram. The training of the CFNN model is accomplished by a hybrid method combined with a genetic algorithm and a least squares method. As a result, LOCA break size is estimated exactly by the proposed CFNN model.

  19. LOCA and RIA studies at JAERI

    International Nuclear Information System (INIS)

    Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi

    2004-01-01

    To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI. (Author)

  20. An estimation of the accident sequence of the LOCA groups for the PSA model of the KSNP

    International Nuclear Information System (INIS)

    Han, Seok Jung; Yang, Joon Eon

    2004-01-01

    A new trend of the probabilistic safety assessment (PSA) technology is to improve and enhance the current PSA model to be adequate for risk-informed applications (RIA). Requirements of a PSA model for the RIA are summarized as (1) reduction of the conservatism in the model utilizing all available information and (2) consideration of the specific features of a plant as designed, as operated. This is because the PSA based on conservatism and insufficient consideration of the plant-specific features resulted in a shadow effect on the assessment results. When a PSA model is used in a risk-informed application, more precise risk-information is more helpful to decision making process, so the reduction of the conservatism and the consideration of the plant-specific features in a PSA model are the most essential elements. Recently, an effort has been performed to modify the current PSA model for the Korea Standard Nuclear Power plant (KSNP) to be used in risk-informed applications. A re-estimation of the accident sequence of the loss of coolant accident (LOCA) groups for the PSA model of the KSNP has been performed

  1. CANDU fuel behaviour under LOCA conditions

    International Nuclear Information System (INIS)

    Kohn, E.

    1989-07-01

    This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

  2. MELCOR ex-vessel LOCA simulations for ITER+

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

    1995-01-01

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

  3. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

    Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

  4. Mitigating Capability Analysis during LOCA for Korean Standard Nuclear Power Plants in Containment Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Park, Soo Yong; Kim, D. H.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    The objective of this paper is to establish Containment spray operational technical bases for the typical Korean Standard Nuclear Power plants (Ulchin units 3 and 4) by modeling the plant, and analyzing a loss of coolant accident (LOCA) using the MAAP code. The severe accident phenomena at nuclear power plants have large uncertainties. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand severe accident sequences and to assess the accident progression accurately by computer codes. Furthermore, it is important to attain the capability to analyze a advanced nuclear reactor design for a severe accident prevention and mitigation.

  5. Fission Product Transport Models Adopted in REFPAC Code for LOCA Conditions in PWR and WWER NPPS

    International Nuclear Information System (INIS)

    Strupczewski, A.

    2003-01-01

    The report presents assumptions and physical models used for calculations of fission product releases from nuclear reactors, their behavior inside the containment and leakages to the environment after large break loss of coolant accident LB LOCA. They are the basis of code REFPAC (RElease of Fission Products under Accident Conditions), designed primarily to represent significant physical processes occurring after LB LOCA. The code describes these processes using three different models. Model 1 corresponds to established US and Russian practice, Model 2 includes all conservative assumptions that are in agreement with the actual state-of-the-art, and Model 3 incorporates formulae and parameter values actually used in EU practice. (author)

  6. Comparison of LOCA safety analysis in the USA, FRG, and Japan

    International Nuclear Information System (INIS)

    Leach, L.P.; Ybarrondo, L.J.; Hicken, E.F.; Tasaka, K.

    1983-01-01

    The bases for loss-of-coolant accident (LOCA) safety analysis required by reactor licensing regulations in the United States of America (USA), Federal Republic of Germany (FRG), and Japan are investigated and related to new data obtained since the regulations were established. The licensing approaches used in the three countries are similar in that a conservative calculation is called for, the necessary conservatism is unspecified, and new research data have had only limited effect on changing the regulations

  7. Accident analyses on TMLB' and LOCA for KNGR using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Choi, Y.; Ahn, K.I

    2000-11-01

    Plant specific phenomenological analyses for the Korean Next Generation Reactor, using MELCOR program, are described in this report. The most important two accident sequences, a station blackout and a loss of coolant scenario, are selected. Complete coverage of corium behavior both in-vessel and ex-vessel, and the corresponding containment responses, are analyzed. The in-vessel progression includes the thermal hydraulics in the primary system, core heat up, hydrogen generation, and melt progression up to the reactor vessel breach. The ex-vessel progression describes molten corium - concrete interaction phenomena and the pressure behavior in the containment atmosphere.

  8. Taipower's approach in development of in-house LOCA analysis capability

    International Nuclear Information System (INIS)

    Wang, L.C.

    2004-01-01

    One lesson learned from the Three Mile Island (TMI) accident was the analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or nuclear fuel suppliers for small break Loss Of Coolant Accident (LOCA) analysis for compliance with appendix K to 10CFR50 should be revised, so, a technology transfer program and a training program of a new LOCA analysis methodology for Taipower's engineers is briefly described in this paper. Also, an other lesson learned from the TMI accident was the plant-specific calculations using NRC-approved models for small-break LOCA to show compliance with 10CFR50.46 should be submitted for NRC approval, so, a study by Taiwan Power Company (TPC) under the guidance of Yankee Atomic Electric Company (YAEC) has been undertaken to perform this analysis for Maanshan nuclear power plant. The results of the 4 inch line break LOCA analysis is described in this paper. (author)

  9. Containment accident analysis using CONTEMPT4/M0D2 compared with experimental data

    International Nuclear Information System (INIS)

    Metcalfe, L.J.; Hargroves, D.W.; Wells, R.A.

    1978-01-01

    CONTEMPT4/MOD2 is a new computer program developed to predict the long-term thermal hydraulic behavior of light-water reactor and experimental containment systems during postulated loss-of-coolant accident (LOCA) conditions. Improvements over previous containment codes include multicompartment capability and ice condenser analytical models. A program description and comparisons of calculated results with experimental data are presented

  10. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

  11. Large LOCA accident analysis for AP1000 under earthquake

    International Nuclear Information System (INIS)

    Yu, Yu; Lv, Xuefeng; Niu, Fenglei

    2015-01-01

    Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

  12. Severe accident sequences simulated at the Grand Gulf Nuclear Station

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1999-01-01

    Different severe accident sequences employing the MELCOR code, version 1.8.4 QK, have been simulated at the Grand Gulf Nuclear Station (Grand Gulf). The postulated severe accidents simulated are two low-pressure, short-term, station blackouts; two unmitigated small-break (SB) loss-of-coolant accidents (LOCAs) (SBLOCAs); and one unmitigated large LOCA (LLOCA). The purpose of this study was to calculate best-estimate timings of events and source terms for a wide range of severe accidents and to compare the plant response to these accidents

  13. Preliminary Results Of LOCA Problem For APR1400 Reactor

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Le Thi Thu; Vo Thi Huong; Le Van Hong

    2011-01-01

    Several features of NPP with APR1400 nuclear reactor during a loss of coolant accident (LOCA) are investigated in this study. The report describes some main design characteristics of an engineering safety systems of APR1400 and the thermal hydraulic calculation results for steady-state using MARS and RELAP/SCDAPSIM codes. Large Break LOCA accident has been analyzed and evaluated based on acceptable criteria for ECCS given by US NRC. The results from cold leg break LOCA with broken area of 0.0465 m 2 in case of high pressure safety injection system (HPSI) failed to operate or 2 and 4 HPSI pumps are activated. The preliminary results of this work is a part of collaboration between INST researchers and KAERI experts in using RELAP tool for safety analysis of NPPs. (author)

  14. Theoretical analysis of the temperature changes and resultant loss of fuel integrity in the IEA-R1 research reactor fuel elements following a loss of coalant accident

    International Nuclear Information System (INIS)

    Garone, J.G.M.

    1983-01-01

    The IEA-R1 core following a loss of coolant accident (LOCA) is analysed. THe AIRLOCA code was used to calculate fuel temperatures, heat generation due to fission product decay and convective and radiative heat transfer from the fuel elements to the surrounding air both during and following the loss of coolant. The influence of certain critical parameters, such as log time, specific power was studied in detail. Representative results are presented and suggestions made to ensure that fuel integrity is maintained following a LOCA. (Author) [pt

  15. LOCA, LOFA and LOVA analyses pertaining to NET/ITER safety design guidance

    International Nuclear Information System (INIS)

    Ebert, E.; Raeder, J.

    1991-01-01

    The analyses presented pertain to loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA). These types of accidents may jeopardise components and plasma vessel integrity and cause radioactivity mobilisation. The analyses reviewed have been performed under the assumption that the plasma facing components are protected by a carbon based armour. Accidental temperatures and pressure transients are quantified, the possibility of reaction products combustion is investigated and worst case accidental public doses are assessed. On this basis, design recommendations are given and design features such as low plasma facing components armour temperatures (on almost the entire surface) and inert gas adjacent to the vacuum vessel have been implemented. (orig.)

  16. Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2009-01-01

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

  17. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    International Nuclear Information System (INIS)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

    1998-01-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs

  18. The 1994 loss of coolant incident at Pickering NGS

    Energy Technology Data Exchange (ETDEWEB)

    Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

  19. Evaluation method of iodine re-evolution from an in-containment water pool after a loss of coolant accident, Part I: pH estimation of a solution with various chemicals

    International Nuclear Information System (INIS)

    Kim, Tae Hyeon; Jeong, Ji Hwan

    2016-01-01

    Highlights: • It is required to evaluate re-evolved iodine from sump water after LOCA. • pH evaluation based on Gibbs free energy minimization. • Program was developed to evaluate chemical equilibrium and pH solutions. • Predictions are in good agreement with experimental data. - Abstract: Radioactive iodine, which is released into the atmosphere of the containment building, is absorbed into the containment spray water and dissolved to be ionized. This iodine-rich water is then transported to the in-containment refueling water storage tank (IRWST) in APR1400 nuclear power plants. When the pH of the water is below 7, the dissolved iodine converts to molecular iodine and re-evolves from the water and returns to the atmosphere. A series of studies have been conducted in order to evaluate the iodine re-evolution from the IRWST. This study consists of two parts: the pH evaluation method and the evaluation of the iodine re-evolution. This paper presents the first part, i.e. the pH evaluation method. The equilibrium concentrations of various chemicals in a solution are determined at the minimum Gibbs’ free energy. This method is useful for complex reactant problems rather than equilibrium constants method because the latter method requires numerous equilibrium constants and there might be missing equilibrium constants associated with the solution. The calculated pH values of solutions are compared with the experimental measurements in order to validate this method and the thermodynamic data of the chemicals incorporated into the program. The estimated values for solutions are in good agreement with the experimental measurements within a difference of less than 3.3%.

  20. Complete BWR--EM LOCA analysis using the WRAP--EM system

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

    1979-01-01

    The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

  1. Fracture mechanics based assessment of postulated flaws in the nozzle region of RPV-KKS under loss of coolant accidents; Bruchmechanische Bewertung von postulierten Fehlern im Stutzenbereich des RDB-KKS bei Kuehlmittelverlust-Stoerfaellen

    Energy Technology Data Exchange (ETDEWEB)

    Siegele, D.; Hodulak, L.; Varfolomeyev, I. [Fraunhofer-Institut fuer Werkstoffmechanik (IWM), Freiburg im Breisgau (Germany); Nagel, G. [Preussen Elektra AG, Hannover (Germany). Hauptverwaltung

    1998-11-01

    Safety assessment of reactor components has to encompass the nozzles, as in the event of a LOCA, the nozzle areas are subject to heavy thermal stresses due to the low temperature there of the cooling agent. The paper refers to three-dimensional, elasto-plastic FEM analyses of the integer RPV and calculations of the J-integral for various crack locations and geometries, for derivation of realistic transients. The J-integral and K values calculated with FEM have been compared with stress intensity factors determined by means of enhanced analytical methods. Calculations for description of the integer nozzle yield plastifications in the cladded and the ferritic areas, so that the K concept of linear-elastic fracture mechanics is restricted in applicability. For postulated cracks beneath the integer cladding, J-integral values are determined which are below the initiating value J{sub 1} of the material used, which excludes crack initiation. For the largest postulated, 20 mm deep surface crack and through-cladding damage, a crack growth of 0.1 mm is derived according to the crack resistance curve. The analytical method for calculating the stress intensity factors has been expanded to also include application to evaluation of nozzle edge cracks under the cladding. (orig./CB) [Deutsch] Bei der sicherheitstechnischen Bewertung von Reaktorkomponenten sind auch die Stutzen zu beruecksichtigen, da der Stutzenbereich bei Kuehlmittelverlust-Faellen, bedingt durch die dort vorliegenden tiefen Temperaturen des Kuehlmediums, hohen thermischen Beanspruchungen ausgesetzt ist. Fuer realistische Transienten werden dreidimensionale, elastisch-plastische FEM-Analysen fuer den integren RDB und Berechnungen des J-Integrals fuer verschiedene Risslagen und Rissgeometrien durchgefuehrt. Die mit FEM berechneten J-Integral- und K-Werte werden mit nach weiterentwickelten analytischen Methoden ermittelten Spannungsintensitaetsfatoren verglichen. Die Berechnungen fuer den integren Stutzen ergeben

  2. A synopsis of experimental activities on small-break LOCA

    International Nuclear Information System (INIS)

    Hein, D.

    1984-01-01

    Through reactor safety studies like WASH 1400 or the ''Deutsche Risiko-Studie'' the attention has turned from large break loss of coolant accidents to small breaks because of the high contribution of this type of accidents to core meltdown. But only after the TMI-2 accident were also the main activities in the experimental fields shifted world-wide to the small break LOCAs. Since TMI numerous research programs have either been finished or are underway. This review paper presents: a classification of the various types of transients according to break size; a discussion of major physical phenomena associated with a small break LOCA, and a description of a few selected research programs and the most important results achieved. (author)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  4. NPP Krsko containment environmental conditions during postulated accident

    International Nuclear Information System (INIS)

    Kozaric, M.; Cavlina, N.; Spalj, S.

    1989-01-01

    This paper presents NPP Krsko containment pressure and temperature increase during Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). Containment environmental condition calculation was performed by CONTEMPT4/MOD4 computer code. Design accident calculations were performed by RELAP4/MOD6 and RELAP5/MOD1 computer codes. Calculational abilities and application methodology of these codes are presented. The CONTEMPT code is described in more detail. The containment pressure and temperature time distribution are presented as well. (author)

  5. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  6. Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

    2018-01-01

    Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

  7. Accomplishments of LOCA/ECCS experimental research at Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Murao, Yoshio; Koizumi, Yasuo

    1984-01-01

    Japan Atomic Energy Research Institute has investigated loss-of-coolant accident (LOCA)/emergency core cooling system (ECCS) from 1970. Major results of the LOCA/ECCS research are summarized in this report. ROSA-II program was LOCA/ECCS research for a pressurized water reactor (PWR) and ROSA-III program was for a boiling water reactor (BWR). The both test facilities were scaled at approximately 1/400 of the respective reference PWR and BWR. Large scale reflood test is research on reflood phenomena during a large break LOCA of PWR. The test facility is scaled at approximately 1/20 of the reference PWR and the research is still being continued. (author)

  8. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Muhammad Subekti

    2009-01-01

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  9. Integral isolation valve systems for loss of coolant accident protection

    Science.gov (United States)

    Kanuch, David J.; DiFilipo, Paul P.

    2018-03-20

    A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

  10. Sensitivity Study on Analysis of Reactor Containment Response to LOCA

    International Nuclear Information System (INIS)

    Chung, Ku Young; Sung, Key Yong

    2010-01-01

    As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

  11. Sensitivity Study on Analysis of Reactor Containment Response to LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ku Young; Sung, Key Yong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2010-10-15

    As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

  12. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    OpenAIRE

    Bo Cheng; Young-Jin Kim; Peter Chou

    2016-01-01

    In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

  13. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  14. Frequency probabilistic analysis of a small break LOCA due to a power operated relief valve (PORV) for Angra-1 pre-TMI and post-TMI

    International Nuclear Information System (INIS)

    Onusic Junior, J.

    1986-01-01

    After the TMI event efforts were aimed towards improvements in the operational and administrative procedures related to the power operated relief valves (PORVs) in order to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by stuck-open power operated relief valve. This paper presents a frequency probabilistic analysis of a small break LOCA due to a stuck open PORV and safety valve to the Angra I nuclear power plant in operating conditions pre-TMI and post-TMI. (Author) [pt

  15. Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

    International Nuclear Information System (INIS)

    Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

    1981-07-01

    The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times

  16. 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)

  17. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  18. Estimative of core damage frequency in IPEN'S IEA-R1 research reactor due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami; Sabundjian, Gaiane; Cabral, Eduardo Lobo Lustosa

    2009-01-01

    The National Commission of Nuclear Energy (CNEN), which is the Brazilian nuclear regulatory commission, imposes safety and licensing standards in order to ensure that the nuclear power plants operate in a safe way. For licensing a nuclear reactor one of the demands of CNEN is the simulation of some accidents and thermalhydraulic transients considered as design base to verify the integrity of the plant when submitted to adverse conditions. The accidents that must be simulated are those that present large probability to occur or those that can cause more serious consequences. According to the FSAR (Final Safety Analysis Report) the initiating event that can cause the largest damage in the core, of the IEA-R1 research reactor at IPEN-CNEN/SP, is the LOCA (Loss of Coolant Accident). The objective of this paper is estimate the frequency of the IEA-R1 core damage, caused by this initiating event. In this paper we analyze the accident evolution and performance of the systems which should mitigate this event: the Emergency Coolant Core System (ECCS) and the isolated pool system. They will be analyzed by means of the event tree. In this work the reliability of these systems are also quantified using the fault tree. (author)

  19. Replacement divider plate performance under LOCA loading

    International Nuclear Information System (INIS)

    Huynk, H.M.; MClellan, G.H.; Schneider, W.G.

    1997-01-01

    A primary divider plate in a nuclear steam generator is required to perform its partitioning function with a minimum of cross leakage, without degradation in operating performance and without loss of structural integrity resulting from normal and accident loading. The design of the replacement divider plate for normal operating conditions is discussed in some detail in reference 1 and 2. This paper describes the structural response of the replacement divider plate to the severe loading resulting from a burst primary pipe. The loads for which the divider plate structural performance must be evaluated are mild to severe differential pressure transients resulting from several postulated sizes and types of pipe break scenarios. In the unlikely event of a severe Loss of Coolant Accident (LOCA) the divider plate or parts thereof must not exit the steam generator nor completely block the outlet nozzle. For the milder LOCA loads, the integrity of the divider plate and seat bars must be maintained. Analysis for the milder LOCA loads was carried out employing a conservative approach which ignores the actual interaction between the structure and the primary fluid. For these load cases it was shown that the divider plate does not become disengaged from the seat bars. For the more severe pipe breaks, the thermal-hydraulic analysis was coupled iteratively with the structural analysis, thereby taking into account divider plate deformation, in order to obtain a better prediction of the behaviour of the divider plate. In this manner substantial reduction in divider plate response to the more severe LOCA loading was achieved. It has been shown that, for the case of a postulated large LOCA (100% reactor inlet header), the disengagement of the divider plate from the seat bars resulted in an opening smaller than 1% of the divider plate area. (author)

  20. Experimental studies on mitigation of LOCA for a high flux research reactor

    International Nuclear Information System (INIS)

    Saxena, A.K.

    2006-01-01

    Experimental studies on the rewetting behaviour of hot vertical annular channels were performed to study the mitigation of consequences of loss of coolant accident (LOCA) for a high flux research reactor. Studies were carried out to study the rewetting behaviour with hot inner tube, for bottom flooding and top flow rewetting conditions. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 litres per minute (l pm) (11.7 x 10 -5 m 3 s -1 ). The initial surface temperature of the inner tube was varied from 200 to 500 degC. (author)

  1. Safety studies on LOCA for N.S. Mutsu

    International Nuclear Information System (INIS)

    Kawasaki, Masayuki; Yaguchi, Shinnosuke

    1978-01-01

    A number of safety studies are under way concerning the reactor plant of N.S. Mutsu. One such study relates to Loss of Coolant Accidents (LOCA), which has been conducted to cover mainly the two subjects of experiments to ascertain the integrity of stainless steel fuel cladding under the action of the Emergency Core Cooling System (ECCS), and analysis of containment integrity following a LOCA. The stainless steel cladding tests were conducted to test swelling, rupture, oxidation and compression characteristics. Few reports are known to have been published in this domain, so that the present results should prove useful for future studies related to ECCS evaluation analyses on stainless steel fuel cladding. The containment integrity analysis covered variations of containment pressure and temperature following a LOCA, performed separately for short- and long-term periods. Estimates were also made on the changes in the hydrogen concentration present inside the containment after a LOCA. The results obtained should serve in determining the characteristic response to LOCA of marine reactor plants

  2. Status of efforts to evaluate LOCA frequency estimates using combined PRA and PFM approaches

    International Nuclear Information System (INIS)

    Wilkowski, G.; Rudland, D.; Tregoning, R.; Scott, P.

    2002-01-01

    The risk-informed reevaluation of 10 CFR 50.46 (along with Appendix K and GDC 35), the emergency core cooling system (ECCS) requirements, utilizes loss of coolant accident (LOCA) initiating event frequencies to evaluate the technical basis for potential related rule changes. A longer-term effort is considering redefining the maximum design basis pipe break size for sizing the ECCS system. In the past few years, the U.S. Nuclear Regulatory Commission (NRC) has utilized NUREG/CR-5750 pipe-break LOCA estimated for initiating event frequencies. However, several failure mechanisms have recently emerged at plants which have not been evident within the service period covered by the NUREG/CR-5750 estimates. The concern is that these and other potential aging-related mechanisms may not be adequately represented within the NUREG/CR-5750 LOCA estimates. Additionally, LOCAs can occur from failure of active components (e.g. safety relief valves, reactor coolant pump seals, etc.) and other non-pipe break passive failures (e.g. steam generator tubes). The LOCA contributions from these additional sources must also be considered in deciding the design basis break size. The LOCA estimates must also attempt to capture expected future changes in the LOCA frequencies so that the estimates are pertinent up through the end of the license renewal period. (orig.)

  3. Evaluation of post-LOCA long term cooling performance in Korean standard nuclear power plants

    International Nuclear Information System (INIS)

    Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

    2001-01-01

    The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT) isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences. (author)

  4. Accident sequences simulated at the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1998-01-01

    Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident (LOCA) with the emergency core coolant system (ECCS) on, (2) a station blackout (SBO), (3) a small LOCA (SLOCA) concurrent with SBO, (4) a large LOCA (LLOCA) concurrent with SBO, and (5) a LLOCA concurrent with SBO and with the containment breached at time zero. Timings of important events and source term releases have been calculated for the different sequences analyzed. Under certain weather conditions, the fission products released from the severe accident sequences may travel to southern Florida

  5. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

  6. 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)

  7. An analysis of LOCA sequences in the development of severe accident analysis DB

    International Nuclear Information System (INIS)

    Choi, Young; Park, Soo Yong; Ahn, Kwang-Il; Kim, D.H.

    2006-01-01

    Although a Level 2 PSA was performed for the Korean Standard Power Plants (KSNPs), and it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results. At present, KAERI is calculating the severe accident sequences intensively for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviours. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by knowledge-base technique, and the expected plant behaviour. The plant model used in this paper is oriented to the case of LOCAs related severe accident phenomena and thus can simulate the plant behaviours for a severe accident. Therefore the developed system may play a central role as an information source for decision-making for a severe accident management, and will be used as a training simulator for a severe accident management. (author)

  8. LOCA analysis of SCWR-M with passive safety system

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

    2013-06-15

    Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

  9. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

  10. Scaling criteria and an assessment of Semiscale Mod-3 scaling for small-break loss-of-coolant transients

    International Nuclear Information System (INIS)

    Larson, T.K.; Anderson, J.L.; Shimeck, D.J.

    1982-01-01

    Various methods of scaling fluid thermal-hydraulic test facilities and their relative merits and disadvantages are examined in light of nuclear reactor safety considerations. Particular emphasis is placed on examination of the scaling of the Semiscale Mod-3 system and determination of thermal-hydraulic phenomena thought to be important during a small break loss-of-coolant accident in a pressurized water nuclear reactor. The influence of geometric and dynamic scaling concerns in the Mod-3 system on small break behavior are addressed from an engineering viewpoint and corrective measures contemplated or required to make results from Semiscale tests more meaningful relative to expected PWR response are discussed

  11. The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Abdul-Razzak, A.; Zhang, J.; Sills, H.E.; Flatt, L.; Jenkins, D.; Wallace, D.J.; Popov, N.

    2002-01-01

    The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its proto-typing application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX R fuel. The methodology is consistent with and builds on world practice. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions. (authors)

  12. Accident at Three Mile Island nuclear power plant and lessons learned

    International Nuclear Information System (INIS)

    Ashrafi, A.; Farnoudi, F.; Tochai, M.T.M.; Mirhabibi, N.

    1986-01-01

    On March 28, 1979, the TMI, unit 2 nuclear power plant experienced a loss of coolant accident (LOCA) which has had a major impact among the others, upon the safety of nuclear power plants. Although a small part of the reactor core melted in this accident, but due to well performance of the vital safety equipment, there was no serious radioactivity release to the environment, and the accident has had no impact on the basic safety goals. A brief scenario of the accident, its consequences and the lessons learned are discussed

  13. ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-06-01

    Full Text Available ABSTRAK ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA. Kecelakaan yang diakibatkan oleh kehilangan pendingin (loss of coolant accident / LOCA dari sistem reaktor merupakan kejadian dasar desain yang tetap diantisipasi dalam desain reaktor daya yang mengadopsi teknologi Generasi II hingga IV. LOCA ukuran kecil (small break LOCA memiliki dampak yang lebih signifikan terhadap keselamatan dibandingkan LOCA ukuran besar (large break LOCA seperti terlihat pada kejadian Three-Mile Island (TMI. Fokus makalah adalah pada analisis small break LOCA pada reaktor daya maju Generasi III+ yaitu AP1000 dengan mensimulasikan tiga kejadian pemicu yaitu membukanya katup Automatic Depressurization System (ADS secara tak disengaja, putusnya salah satu pipa Direct Vessel Injection (DVI secara double-ended, dan putusnya pipa lengan dingin dengan diameter bocoran 10 inci. Metode yang digunakan adalah simulasi kejadian pada model AP1000 yang dikembangkan secara mandiri menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Dampak yang ingin dilihat adalah kondisi teras selama terjadinya small break LOCA yang terdiri dari pembentukan mixture level dan transien temperatur kelongsong bahan bakar. Hasil simulasi menunjukkan bahwa mixture level untuk semua kejadian small break LOCA berada di atas tinggi teras aktif yang menunjukkan tidak terjadinya core uncovery. Adanya mixture level berpengaruh pada transien temperatur kelongsong yang menurun dan menunjukkan pendinginan bahan bakar yang efektif. Hasil di atas juga identik dengan hasil perhitungan program lain yaitu NOTRUMP. Keefektifan pendinginan teras juga disebabkan oleh berfungsinya injeksi pendingin melalui fitur keselamatan pasif yang menjadi ciri reaktor daya AP1000. Secara keseluruhan, hasil analisis menunjukkan model AP1000 yang telah dikembangkan dengan RELAP5 dapat digunakan untuk keperluan analisis kecelakaan dasar desain pada reaktor daya maju AP1000. Kata kunci: analisis

  14. Results of small break LOCA analysis for Kuosheng nuclear power plant using the RELAP5YA computer code

    International Nuclear Information System (INIS)

    Wang, L.C.; Jeng, S.C.; Chung, N.M.

    2004-01-01

    One lesson learned from the Three Mile Island (TMI) accident was the analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or nuclear fuel suppliers for small break Loss Of Coolant Accident (LOCA) analysis for compliance with appendix K to 10CFR50 should be revised, documented and submitted for USNRC approval and the plant-specific calculations using NRC-approved models for small-break LOCA to show compliance with 10CFR50.46 should be submitted for NRC approval. A study by Taiwan Power Company (TPC) under the guidance of Yankee Atomic Electric Company (YAEC) has been undertaken to perform this analysis for Kuosheng nuclear power plant. This paper presents the results of the analysis that are useful in satisfying the same requirements of the Republic Of China Atomic Energy Commission (ROCAEC). (author)

  15. Fuel-rod response during the large-break LOCA Test LOC-6

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

  16. Radial heat transfer from fuel to moderator during LOCAs for CANDU PHW reactors

    International Nuclear Information System (INIS)

    Hildebrandt, J.G.; So, C.B.; Gillespie, G.E.; MacLean, G.

    1983-01-01

    In a postulated CANDU-PHW loss-of-coolant accident (LOCA) with coincident impaired emergency cooling, the axial transport of heat from the fuel by convection is reduced. This reduction in heat removal causes the fuel to heat up and the radial heat transfer to the moderator to become significant. This paper deals with two codes that predict the thermal response of fuel channels under LOCA conditions. New channel thermal radiation models in both RAMA, a thermalhydraulic code, and CHAN II, a fuel channel thermo-chemical code, are presented and their predictions are compared with the experimental results of an electrically heated bundle of 37 fuel pins. A second experiment, involving a single heated pin in a channel with flowing steam, is presented. The predictions of RAMA and CHAN II are compared with this experiment to verify the codes' thermo-chemical models. There is good agreement between the predictions of both codes and the experimental results

  17. Embrittlement of pre-hydrided Zircaloy-4 by steam oxidation under simulated LOCA transients

    Energy Technology Data Exchange (ETDEWEB)

    Desquines, J., E-mail: jean.desquines@irsn.fr; Drouan, D.; Guilbert, S.; Lacote, P.

    2016-02-15

    During a Loss Of Coolant Accident (LOCA), the mechanical behavior of high temperature steam oxidized fuel rods is an important issue. In this study, as-received and pre-hydrided axial tensile samples were steam oxidized in a vertical furnace and water quenched in order to simulate a LOCA transient. The samples were then subjected to a mechanical test to determine the failure conditions. Two different rupture modes were evidenced; the first one associated to linear elastic fracture mechanics and the second one is associated to sample failure without applied load. The oxidized cladding fracture toughness was determined relying on intensive metallographic analysis. The sample failure conditions were then back predicted confirming that the main rupture parameters are well captured.

  18. Estimation of maximum pressure in small containments of PWR reactors due to loss of coolant accident in primary circuit; Estimativa da pressao maxima em contencoes de reatores PWR de pequeno porte devido a um acidente de perda de refrigerante no circuito primario

    Energy Technology Data Exchange (ETDEWEB)

    Mendes Neto, Teofilo [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil); Moreira, Joao Manoel Losada [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    This work studies the problem of containment pressurization after a LOCA in reactors with small containment free volumes. The relationship between the reactor power and the containment free volume is described with the ratio between the volumes of the primary circuit and of the containment. The maximum pressure in a containment, following a LOCA, obtained after a correlation based on large containment PWR, is around 185 psia for a primary circuit and containment volumes ratio of 0.025. For the same problem, calculations with the CONTEMPT-LT code produced a maximum pressure of 162 psia. The behavior of the temperature after a LOCA to the containment, as a function of the ratio between the primary circuit and containment volume, is such that it increases reaching asymptotically to a maximum; differently, the pressure increases almost linearly with the ratio of volumes. (author)

  19. Loss of coolant acident analyses on Osiris research reactor using the RELAP5 code

    International Nuclear Information System (INIS)

    Soares, Humberto Vitor; Costa, Antonella Lombardi; Lima, Claubia Pereira Bezerra; Veloso, Maria Auxiliadora Fortini

    2011-01-01

    RELAP5/MOD 3.3 code is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that RELAP5 code can also be applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this paper, a nodalization of the core and the most important components of the primary cooling system of the OSIRIS reactor developed for RELAP5 thermal hydraulic code are presented as well as results of steady state and transient simulations. OSIRIS has thermal power of 70 MW and it is an open pool type research reactor moderated and cooled by water. The OSIRIS reactor characteristics have been used as a base for the development of a model for the Multipurpose Brazilian Reactor (RMB). The aim of the present work is to investigate the behavior of the core during a loss of coolant accident and the possible damage of the fuel elements due an inadequate heat removal. Although the core coolant reached the saturation point due the large break, the fuel element conditions were out of the damage zone. (author)

  20. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  1. A comparison of LOCA analysis using SMOKIN and CERBERUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Younis, M H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Gaboury, G [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    This paper presents the results of a comparison of the analyses of a postulated Loss of Coolant Accident (LOCA) in Pickering NGS A reactors using the two neutron kinetics codes SMOKIN and CERBERUS. Both codes have been used to simulate the space-time neutronic behaviour of CANDU-PHWR reactors. The main objective of the present study is to evaluate the accuracy with which SMOKIN can predict power transients compared to CERBERUS. The comparison shows that the two codes produce similar bulk power and reactivity transients. However, SMOKIN was found to overestimate the power transient (relative to CERBERUS) in some regions of the core, which is indicative of the spatial differences between the two codes. It was demonstrated that part of this overestimate is due to the use of reaction-rate averaged fuel properties in SMOKIN, compared to instantaneous fuel properties in CERBERUS. (author). 5 refs., 3 tabs., 6 figs.

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  4. Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong Kwan, E-mail: jksuh@khnp.co.kr [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kim, Jae Won; Kwon, Sun Guk; Lee, Jae Yong [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Cho, Hyoung Kyu; Park, Goon Cherl [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

    2015-08-15

    Highlights: • In-vessel downstream effect tests were performed in the presence of LOCA-generated debris. • Available driving heads under each LOCA scenario were verified using experimental data. • Fibrous debris was prepared to satisfy the length distribution obtained from the bypass test. • Limiting test conditions were identified through sensitivity studies. - Abstract: Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

  5. Hydrogen and steam distribution following a small-break LOCA in large dry containment

    Institute of Scientific and Technical Information of China (English)

    DENG Jian; CAO Xuewu

    2007-01-01

    The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations.Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident (LOCA) without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump (RCP)compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.

  6. The Effect of External Vessel Cooling for a 2 inch LOCA Severe Accident Scenario at SMART with MIDAS/SMR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

    2010-01-01

    KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. This feature may prevent the large size of LOCA. However it is necessary to estimate the hypothetical severe accidents progression for improving the degree of safety and identifying the unknown weakness of the system against an accident. To simulate a hypothetical severe accident for the SMART, we adopt the MIDAS/SMR code which was developed by KAERI

  7. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  8. Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term

    Directory of Open Access Journals (Sweden)

    Sunarko

    2017-01-01

    Full Text Available Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2 PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region.

  9. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  10. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  11. JCO criticality accident as POST-LOCA: Poor structure induced loss of organizational control accident

    International Nuclear Information System (INIS)

    Furuhama, Yutaka

    2000-01-01

    Some problems in operation and business management of JCO (Japan Nuclear Fuel Conversion Co.) have been studied as background factors of the criticality accident. Open information about business conditions of JCO suggests that the cause of the accident is not so simple as to be attributed only to economic pressure, but includes immanent problems in JCO. We investigate the problems from five viewpoints, organization of safety management, system of operation management, activities for business improvement, risk awareness, and restructuring of business, and discuss the effects and causality of background factors as well as remedies for them. (author)

  12. Progress in realistic LOCA analysis

    International Nuclear Information System (INIS)

    Young, M.Y.; Bajorek, S.M.; Ohkawa, K.

    2004-01-01

    In 1988 the USNRC revised the ECCS rule contained in Appendix K and Section 50.46 of 10 CFR Part 50, which governs the analysis of the Loss Of Coolant Accident (LOCA). The revised regulation allows the use of realistic computer models to calculate the loss of coolant accident. In addition, the new regulation allows the use of high probability estimates of peak cladding temperature (PCT), rather than upper bound estimates. Prior to this modification, the regulations were a prescriptive set of rules which defined what assumptions must be made about the plant initial conditions and how various physical processes should be modeled. The resulting analyses were highly conservative in their prediction of the performance of the ECCS, and placed tight constraints on core power distributions, ECCS set points and functional requirements, and surveillance and testing. These restrictions, if relaxed, will allow for additional economy, flexibility, and in some cases, improved reliability and safety as well. For example, additional economy and operating flexibility can be achieved by implementing several available core and fuel rod designs to increase fuel discharge burnup and reduce neutron flux on the reactor vessel. The benefits of application of best estimate methods to LOCA analyses have typically been associated with reductions in fuel costs, resulting from optimized fuel designs, or increased revenue from power upratings. Fuel cost savings are relatively easy to quantify, and have been estimated at several millions of dollars per cycle for an individual plant. Best estimate methods are also likely to contribute significantly to reductions in O and M costs, although these reductions are more difficult to quantify. Examples of O and M cost reductions are: 1) Delaying equipment replacement. With best estimate methods, LOCA is no longer a factor in limiting power levels for plants with high tube plugging levels or degraded safety injection systems. If other requirements for

  13. Monitoring severe accidents using AI techniques

    International Nuclear Information System (INIS)

    No, Young Gyu; Ahn, Kwang Il; Kim, Ju Hyun; Na, Man Gyun; Lim, Dong Hyuk

    2012-01-01

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  14. Monitoring severe accidents using AI techniques

    Energy Technology Data Exchange (ETDEWEB)

    No, Young Gyu; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of); Lim, Dong Hyuk [Korea Institute of Nuclear Nonproliferation and Control, Daejon (Korea, Republic of)

    2012-05-15

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  15. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    International Nuclear Information System (INIS)

    Roth, P.A.; Schultz, R.R.; Choi, C.J.

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests

  16. A PC Mathcad-based computational aid for severe accident analysis and its application to a BWR small LOCA sequence

    International Nuclear Information System (INIS)

    Wu, Laung-Kuang T.; Lee, S.J.

    2004-01-01

    A PC-based Mathcad program is used to develop a computational aid for analyzing severe accident phenomena. This computational aid uses simple engineering expressions and empirical correlations to estimate key quantities and timings at various stages of accident progressions. In this paper, the computational aid is applied to analyze an early phase of a BWR small LOCA sequence. The accident phenomena analyzed include: break flow rates, boiled-up water level in the core, core uncovery time, depressurization of the reactor pressure vessel, core heat-up, onset of clad oxidation, hydrogen generation, and onset of fuel relocation. The results are compared with those obtained running the MAAP 3.0B code. This PC-based computational aid can be used to train plant personnel in understanding severe accident phenomena and to assist them in managing severe accidents. (author)

  17. Improvement on models associated with LOCA and loss of RHR accidents during shutdown

    International Nuclear Information System (INIS)

    Chang, W. P.; Chung, Y. J.; Kim, W. S.; Kim, K. D.; Lee, S. J.; Jung, J. J.; Ha, G. S.; Son, Y. S.; Chung, B. D.; Han, D. H.; Lee, Y. J.; Hwang, T. S.; Lee, S. Y.; Park, C. Y.; Choi, H. R.; Lee, S. Y.; Choi, J. H.; Ban, C. H.; Bae, G. H.

    1997-07-01

    The characteristics of the best estimate codes available in Korea have been studied through literature surveys for the reliability on LOCA analyses and then, a feasibility study on reduction of capacities of existing safety systems in YGN 3/4 have been carried out using the codes. Since it has been expected to adopt DVI + 4 -Train HPSI in the next generation reactor, the core uncoveries under one DVI line break and 6 cold leg break, which is a requirement for advance d reactor by EPRI, in addition to LBLOCA for reduction effect of SIT capacity, have been analyzed. Finally, an effort on finding the way how the system could be simplified, has been made through the analysis of SIT injection characteristics. On the other hand, the best estimate methodology consisting of uncertainties of the code itself, bias, and application have been developed first and quantification of the uncertainty has been made the case of KORI unit 3 afterward. The prediction capabilities of the best estimate codes and major physical models on the accident under loss of RHR during shutdown have been assessed suing the large scale experimental data delivered from France and then, the assessed codes have been used to provide essential data required for description of operation procedures in YGN 3/4. (author). 64 refs., 45 figs

  18. Thermal stresses at nozzles of nuclear steel containments under LOCA-conditions

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, G.; Bergmann, A.N.

    1986-01-01

    During a loss of coolant accident (LOCA) of a PWR-nuclear power plant, a considerable heating of the containment atmosphere is expected to occur. Transient thermal stresses will appear at the containment as a consequence of a non-uniform rise of its temperature. Applying computer codes based on the finite element method, dimensionless general thermal stresses at nozzles of spherical steel containment have been calculated, varying the principal geometrical parameters and the Biot number for the containment internal surface. Atmosphere temperature and Biot number are assumed constant after the accident. Several plots of the maximum principal stresses are provided, which constitute general results applicable to stress analysis of any particular containment of this kind. (orig.)

  19. Investigation of bubble-condenser operation under large break LOCA conditions

    International Nuclear Information System (INIS)

    Blinkov, V.; Melikhov, O.; Melikhov, V.; Davydov, M.; Sokolin, A.; Hoffmann, D.; Simon, U.; Bajsz, J.

    2000-01-01

    In the framework of the PHARE/TACIS project, the experimental test facility for bubble condenser experimental qualification was built at Electrogorsk Research and Engineering Centre. The test facility contains high pressure system, compartments upstream of the bubble condenser and a section of the bubble condenser system. The scaling of the test facility is 1:100. The high pressure system consists of five vessels to appropriately model the leak functions (mass flow rate and enthalpy) during the loss of coolant accidents postulated in the design of VVER-440/V213. Design basis accident (LB LOCA) was experimentally and analytically considered. Results of pre-test analysis with ATHLET and DRASYS codes for determination of necessary test parameters and post-test analysis of three tests are presented. (author)

  20. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  1. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    International Nuclear Information System (INIS)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S.

    2015-01-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Adams, J.P.

    1982-01-01

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

  4. Evaluation of PWR response to main-steamline break with concurrent steam-generator tube rupture and small-break LOCA

    International Nuclear Information System (INIS)

    Laaksonen, J.T.; Sheron, B.W.

    1982-12-01

    In 1980, the NRC staff raised a potential safety issue involving a coincident steamline break, steam generator tube rupture, and small-break loss-of-coolant accident (LOCA). The bases for this concern were that the system response, primarily the maintenance of core cooling, was unanalyzed and the adequacy of the present guidance to operators to respond to combination LOCAs was unknown. This report discusses the staff evaluations performed to assess the system response and the adequacy of the present emergency operator guidelines. In all of the analyzed cases the primary coolant shrinkage, caused by overcooling, and the simultaneous loss of coolant can be compensated by the high pressure emergency core cooling system. The core remains covered with liquid, and the primary coolant remains subcooled, except in the vessel upper head. If the steamline break is outside the containment and cannot be isolated, the radiological consequences could be more severe than in any accident currently analyzed in a typical plant Final Safety Evaluation Report (FSAR). To decrease the risk of elevated offsite releases, an early diagnosis of the tube rupture has to be ensured. This can be done by upgrading operator instructions. The appropriate mitigating actions are in the existing instructions

  5. Development and Assessment of the Appendix K Version of RELAP5-3D for LOCA Licensing Analysis

    International Nuclear Information System (INIS)

    Liang, Thomas K.S.; Chang, C.-J.; Hung, H.-J.

    2002-01-01

    In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss-of-coolant accident (LOCA) would limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during a LOCA, it generally takes much more resources to develop. Instead, implementation of evaluation models required by Appendix K of 10CFR50 on an advanced thermal-hydraulic platform such as RELAP5, TRAC, etc., also can gain significant margin for the PCT calculation. Through compliance evaluation against Appendix K of 10CFR50, all of the required evaluation models have been implemented in RELAP5-3D. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effects experiments and eight sets of LOCA integral experiments were adopted. Through the assessments against separate-effects experiments, the success of the code modification in accordance with Appendix K of 10CFR50 was demonstrated. Besides, one set of a typical integral large-break LOCA from Loss-of-Fluid Test Facility experiments (L2-5) has also been applied to preliminarily evaluate the integral performance of the Appendix K version of RELAP5-3D. The PCT predicted by the evaluation models is greater than the one from best-estimate calculation in the whole LOCA history with the conservatism of 150 K, and the measured PCTs of L2-5 are also well bounded by the evaluation model calculation. Another seven sets of integral-effect experiments will be further applied in the next step to ensure the reasonable integral conservatism of the newly developed LOCA licensing analysis code (RELAP5-3DK/INER), which can cover all the phases of both large- and small LOCA in one code

  6. Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Youngmin; Kim, Young-In; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    KAERI (Korea Atomic Energy Research Institute) is developing a passive safety injection system (PSIS) to supply cold borated water into a reactor coolant system (RCS) without any operator actions or AC power under the occurrence of postulated design basis accidents. The PSIS consists of four independent trains, each of which is furnished with a gravity drained core makeup tank (CMT) and a safety injection tank (SIT). The CMT is designed to provide makeup and boration functions to the RCS during the early stage of a loss of coolant accident (LOCA) and a non-LOCA. In this paper, we investigate numerically the fluid mixing characteristics in the upper annular space of SMART, especially when single-phase natural circulation is formed between the CMT and RCS following a non-LOCA such as a main steam line break. In this paper, the fluid mixing characteristics in the upper annular space of SMART are investigated numerically when single-phase natural circulation is formed between the RCS and CMT during the early stage of a non-LOCA.

  7. Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

    International Nuclear Information System (INIS)

    Bae, Youngmin; Kim, Young-In; Kim, Keung Koo

    2015-01-01

    KAERI (Korea Atomic Energy Research Institute) is developing a passive safety injection system (PSIS) to supply cold borated water into a reactor coolant system (RCS) without any operator actions or AC power under the occurrence of postulated design basis accidents. The PSIS consists of four independent trains, each of which is furnished with a gravity drained core makeup tank (CMT) and a safety injection tank (SIT). The CMT is designed to provide makeup and boration functions to the RCS during the early stage of a loss of coolant accident (LOCA) and a non-LOCA. In this paper, we investigate numerically the fluid mixing characteristics in the upper annular space of SMART, especially when single-phase natural circulation is formed between the CMT and RCS following a non-LOCA such as a main steam line break. In this paper, the fluid mixing characteristics in the upper annular space of SMART are investigated numerically when single-phase natural circulation is formed between the RCS and CMT during the early stage of a non-LOCA

  8. Analysis of the metallic containment integrity of Angra 2/3 reactor under the effects of the design basis accident

    International Nuclear Information System (INIS)

    Costa, J.R.

    1981-06-01

    The application of Condru 4 computer code, developed to determine the maximum values of pressure and temperature that occur inside the metallic containment building of PWR nuclear power plants, in case of a hypothetic accident - LOCA - considered as a Design Basic Accident - DBA. The hypothesis, input and results for the simulation of a loss of coolant in the hot leg of the Angra-2/3 reactors, considered as the most critical case for that Kind of project, are presented. The analysis was made with input provided by the manufacturer. (Author) [pt

  9. Severe damage analysis of VVER 1000 following large break LOCA using Astec code

    International Nuclear Information System (INIS)

    Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

    2007-01-01

    Severe accident analysis of a reactor is an important aspect in the evaluation of source term. This in turn helps in emergency planning. An analysis has been carried out for VVER-1000 (V320) reactor following Large Break LOCA (loss of coolant accident) along with Station Blackout (SBO). Computer code ASTEC (jointly developed by IRSN, France, and GRS, Germany) is used for analyzing the transient. This integral code has been designed to be used as reference code for PSA2 studies. Severe accident analysis is carried out for an accident initiated by Large break LOCA along with SBO. Two cases have been analysed with the version ASTEC V1.2-rev1. In the first case hydro-accumulators are considered not available while the second case has been analysed with hydro accumulators. In this paper, ASTEC predictions have been studied for the in-vessel phase of the accident till vessel failure. The vessel failure was observed at 6979 s when accumulators were assumed not available. The vessel failure was quite delayed (19294 s) with operating accumulators. The hydrogen production was found to be very large (22% of total Zr inventory) in the case with accumulators compared to the case without accumulators (1.5% of total Zr inventory)

  10. Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of excedeence of damage by integrated Safety Analysis Methodology

    International Nuclear Information System (INIS)

    Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

    2013-01-01

    The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Excedence Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

  11. Primary LOCA in VVER-1000 by pressurizer PORV failure

    International Nuclear Information System (INIS)

    Sabotinov, L.; Lutsanych, S.; Kadenko, I.

    2013-01-01

    The paper presents the calculations and analysis of the design basis accident of a standard VVER-1000/V320 reactor with inadvertent opening and stuck in open position of the pressurizer pilot operated relief valve (PORV). The objective is the independent assessment of this accident applying the French best estimate thermal-hydraulic computer code CATHARE 2 and verification to meet the safety criteria for such kind of the accident. The 'Inadvertent opening and stuck in open position of PORV' is a design basis accident classified as Medium Break Loss of Coolant Accident (MB LOCA) with the equivalent diameter of the break D- 68 mm. This accident is particularly interesting to be calculated and analyzed, because it took place at operating NPP with VVER-1000 reactors (Rovno NPP) in 2009. The calculations have been carried out with conservative conditions as usual for DBA analysis. The NPP model corresponds to a real VVER-1000/V320 configuration and comprises all safety systems, adopted for one of the latest CATHARE 2 versions. The results of CATHARE 2 calculations are compared with available results of RELAP5 calculations. There is similarity of the thermal-hydraulic parameters behavior, but also some differences can be observed basically due to the break flow prediction, which causes differences in primary pressure evaluation. Both calculations show that there is no boiling crisis in the reactor core and reliable cooldown is achieved. The calculations performed with CATHARE2 code demonstrate the ability of the code to predict reasonably the break flow, pressures, temperatures etc. for considered LOCA scenario and to be applied for safety studies

  12. A simplified time-dependent recovery model as applied to RCP seal LOCAs

    International Nuclear Information System (INIS)

    Kohut, P.; Bozoki, G.; Fitzpatrick, R.

    1991-01-01

    In Westinghouse-designed reactors, the reactor coolant pump (RCP) seals constantly require a modest amount of cooling. This cooling function depends on the service water (SW) system. Upon the loss of the cooling function due to the unavailability of the SW, component cooling water system or electrical power (station blackout), the RCP seals may degrade, resulting in a loss-of-coolant accident (LOCA). Recent studies indicate that the frequency of the loss of SW initiating events is higher than previously thought. This change significantly increases the core damage frequency contribution from RCP seal failure. The most critical/dominant element in the loss of SW events was found to be the SW-induced RCP seal failure. For these potential accident scenarios, there are large uncertainties regarding the actual frequency of RCP seal LOCA, the resulting leakage rate, and time-dependent behavior. The roles of various recovery options based on the time evolution of the seal LOCA have been identified and taken into account in recent NUREG-1150 probabilistic risk assessment PRA analyses. In this paper, a consistent time-dependent recovery model is described that takes into account the effects of various recovery actions based on explicit considerations given to a spectrum of time- and flow-rate dependencies. The model represents a simplified approach but is especially useful when extensive seal leak rate and core uncovery information is unavailable

  13. Statistics and integral experiments in the verification of LOCA calculations models

    International Nuclear Information System (INIS)

    Margolis, S.G.

    1978-01-01

    The LOCA (loss of coolant accident) is a hypothesized, low-probability accident used as a licensing basis for nuclear power plants. Computer codes which have been under development for at least a decade have been the principal tools used to assess the consequences of the hypothesized LOCA. Models exist in two versions. In EM's (Evaluation Models) the basic engineering calculations are constrained by a detailed set of assumptions spelled out in the Code of Federal Regulations (10 CFR 50, Appendix K). In BE Models (Best Estimate Models) the calculations are based on fundamental physical laws and available empirical correlations. Evaluation models are intended to have a pessimistic bias; Best Estimate Models are intended to be unbiased. Because evaluation models play a key role in reactor licensing, they must be conservative. A long-sought objective has been to assess this conservatism by combining Best Estimate Models with statisticallly established error bounds, based on experiment. Within the last few years, an extensive international program of LOCA experiments has been established to provide the needed data. This program has already produced millions of measurements of temperature, density, and flow and millions of more measurements are yet to come

  14. Bulging of pressure tubes at hot spots under LOCA conditions

    International Nuclear Information System (INIS)

    Manu, C.; Shewfelt, R.S.W.; Wright, A.C.D.; Aboud, R.; Lau, J.H.K.; Sanderson, D.B.

    1996-01-01

    During certain postulated loss-of-coolant accidents (LOCA) in a CANDU reactor, some fuel channels can become highly voided within a very short time. Although the pressure tubes are heated mainly by convection and thermal radiation during the LOCA transient, additional heat flow occurs through the bearing pads that are in contact with the pressure tribe. This contact can lead to local hot spots and associated thermal stresses in the pressure tube wall. The two factors that affects the behavior of the pressure tubes during LOCA conditions are the internal pressure and the local heating. Although the effect of internal pressure and of axially uniform temperature has been studied elsewhere, the effect of the local heating on the pressure tube behavior has not been modelled before. This paper shows that the bulging of a pressure tube at a hot spot is the result of the thermal stresses that are developed in a pressure tube during a LOCA transient. To isolate the local heating effect from the internal pressure, a series of single-effect experiments was performed. In these experiments, sections of a CANDU pressure tube were subjected to local heating only. The thermal profile and the local deformation were measured function of time. To quantify the effect of the thermal stresses on the bulging of pressure tubes at hot spots and to develop numerical tools that can predict such bulging, finite element analyses were performed rising the ABAQUS finite element computer code. Use of the measured thermal profiles in the ABAQUS finite element analysis, resulted in very good agreement between the predicted and measured displacements. (author)

  15. SB LOCA analyses for Krsko Full Scope Simulator verification

    International Nuclear Information System (INIS)

    Prosek, A.; Parzer, I.; Mavko, B.

    2000-01-01

    Nuclear power plant simulators are intended to be used for training and maintaining competence to ensure safe, reliable operation of nuclear power plants throughout the world. The simulator shall be specified to a reference unit and its performance validation testing shall be provided. In this study a small-break loss-of-coolant accident (SB LOCA) response of Krsko nuclear power plant (NPP) was calculated for full scope simulator verification. The investigation included five cases with varying the break size in the cold leg of reactor coolant system. The plant specific and verified RELAP5/MOD2 model of Krsko nuclear power plant (NPP), developed in the past for 1882 MWt power, was adapted for 2000 MWt power (cycle 17) including the model for replacement steam generators. The results showed that the plant system response to breaks with small break area was slower compared to breaks with larger break area. The core heatup occurred in most of the cases analyzed. The acceptance criteria for emergency core cooling system were also met. The predicted results of the SB LOCA analysis for Krsko NPP suggest that they may be used for verification of the Krsko Full Scope Simulator performance. (author)

  16. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  17. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

    Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

    1991-10-01

    The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

  18. Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of exceedance of damage by integrated Safety Analysis Methodology; Arboles de sucesos dinamicos aplicados a secuencias Full Spectrum LOCA. Calculo de la frequencia de excedencia del dano mediante la metodologia Analisis Integrados de Seguridad (ISA)

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

    2013-09-01

    The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Exceedance Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

  19. The fuel and channel thermal/mechanical behaviour code FACTAR 2.0 (LOCA)

    International Nuclear Information System (INIS)

    Westbye, C.J.; Mackinnon, J.C.; Gu, B.W.

    1996-01-01

    The computer code FACTAR 2.0 (LOCA) models the thermal and mechanical response of components within a single CANDU fuel channel under loss-of-coolant accident conditions. This code version is the successor to the FACTAR 1.x code series, and features many modelling enhancements over its predecessor. In particular, the thermal hydraulic treatment has been extended to model reverse and bi-directional coolant flow, and the axial variation in coolant flow rate. Thermal radiation is calculated by a detailed surface-to-surface model, and the ability to represent a greater range of geometries (including experimental configurations employed in code validation) has been implemented. Details of these new code treatments are described in this paper. (author)

  20. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  2. Small break LOCA RELAP5/MOD3 uncertainty quantification: Bias and uncertainty evaluation for important phenomena

    International Nuclear Information System (INIS)

    Ortiz, M.G.; Ghan, L.S.; Vogl, J.

    1991-01-01

    The Nuclear Regulatory Commission (NRC) revised the Emergency Core Cooling System (ECCS) licensing rule to allow the use of Best Estimate (BE) computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability and Uncertainty (CSAU) to evaluate BE code uncertainties. The CSAU methodology was demonstrated with a specific application to a pressurized water reactor (PWR), experiencing a postulated large break loss-of-coolant accident (LBLOCA). The current work is part of an effort to adapt and demonstrate the CSAU methodology to a small break (SB) LOCA in a PWR of B and W design using RELAP5/MOD3 as the simulation tool. The subject of this paper is the Assessment and Ranging of Parameters (Element 2 of the CSAU methodology), which determines the contribution to uncertainty of specific models in the code

  3. Review of RIA and LOCA criteria for WWER fuel

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    The RIA and LOCA fuel safety criteria are under revision in the international community of fuel suppliers, authorities and research organizations. The main criteria will be reviewed in the paper for WWER fuel. Experimental data on the fuel failure behaviour under reactivity-initiated accident (RIA) conditions produced in the last decade in French and Japanese test reactors indicated low failure enthalpy for high burnup fuel compared to fresh fuel. However the high burnup was not the only phenomenon influencing the fuel failure. The oxide scale on the external surface of the fuel rod, hydrogen content of the Zr cladding and the local hydriding seemed also be responsible for the failure at low enthalpy. Furthermore differences have been found between Western design fuel and Russian type WWER fuel. The burnup dependence of fuel failure for WWER fuel was found much less, probably due to the low oxidation during normal operational conditions compared to other PWRs. The recently published Vitanza and KAERI correlations for RIA failure enthalpy have been applied to 23 WWER tests. Experimental data from Russian IGR and BIGR reactors have been used. The calculations have shown that both burnup and cladding oxidation effects must be considered, however the pulse width dependence of failure enthalpy has not been confirmed. During loss of coolant accidents (LOCA) the peak cladding temperature and local oxidation criteria have to be met. The oxidation criterion is under discussion today in many laboratories. The AEKI carried out several experimental series with Zr1%Nb cladding used in WWER reactors. The paper will describe the main results of the tests and present the limit for ductile-brittle transition derived from ring compression test. The behaviour of Zr1%Nb (E110) and Zircaloy-4 claddings under LOCA conditions will be compared as well. (author)

  4. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    International Nuclear Information System (INIS)

    Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-01-01

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  5. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  6. Strategies of modeling the cognitive tasks of human operators for accident scenarios in nuclear power plant control rooms

    International Nuclear Information System (INIS)

    Cheon, Se Woo; Sur, Sang Moon; Lee, Yong Hee; Lee, Jeong Wun

    1993-01-01

    This paper presents the development strategies of cognitive task network modeling for accident scenarios in nuclear power plant control rooms. Task network modeling is used to provide useful predictions of operator's performance times and error rates, based upon plant procedures and/or control room changes. Two accident scenarios, small-break loss of coolant accident (LOCA) and steam generator tube rupture (SGTR), are selected for task simulation. To obtain the input data for the model, task elements are extracted by the task analysis of emergency operating procedures. The input data include task performance time, communication ink, panel location, component operating mode, and data for performance shaping factors (PSFs). Operator's verbs are categorized according to the elements of cognitive behavior. The simulation of the task network for the small-break LOCA scenario is presented in this paper. (Author)

  7. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

    2008-01-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

  8. Simulated LOCA Test and Characterization Study Related to High Burn-Up Issue

    International Nuclear Information System (INIS)

    Park, D. J.; Jung, Y. I.; Choi, B. K.; Park, S. Y.; Kim, H. G.; Park, J. Y.

    2012-01-01

    For the safety evaluation of fuel cladding during the injection of emergency core coolant, simulated Loss-of-coolant accident (LOCA) test was performed by using Zircaloy-4 fuel cladding samples. Zircaloy-4 tube samples with and without prehydring were oxidized in a steam environment with the test temperature of 1200 .deg. C. Prehydrided cladding was prepared from as-fabricated Zircaloy-4 to study the effects of hydrogen on mechanical properties of cladding during high temperature oxidation and quench conditions. In order to measure the ductility of the tube samples embrittled by quenching water, ring compression test was carried out by using 8 mm ring sample sectioned from oxidized tube sample and microstructural analysis was also performed after simulated LOCA test. The results showed that hydrogen increases oxygen solubility and pickup rate in the beta layer. This reduces ductility of prehydrided fuel cladding compared with as-fabricated cladding. Trend in ductility decrease for prehydrided sample under simulated LOCA condition was very similar with data obtained from tests conducted using irradiated high burn-up fuel claddings

  9. Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

    International Nuclear Information System (INIS)

    Arshad, M. W.

    2012-01-01

    Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

  10. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  11. LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The third OECD LOFT experiment was conducted on 14 July 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with delayed pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  12. LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The second OECD LOFT experiment was conducted on 23 June 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with early pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  13. Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Larson, T.K.

    1978-01-01

    Tests which simulated rupture of steam generator tubes during loss-of-coolant experiments in a PWR type system have been conducted in the Semiscale Mod-1 system. Analysis of test data indicates that high rod cladding temperatures occured only for a band of tube ruptures (between 12 and 20 tubes) and that the peak cladding temperatures attained within this band were strongly dependent on the magnitude of the tube rupture flow rates. Maximum cladding temperature of about 1255 K was observed for tests which simulated tube ruptures within this narrow band. (author)

  14. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  15. Effective water cooling of very hot surfaces during the LOCA accident.

    Czech Academy of Sciences Publication Activity Database

    Štepánek, J.; Bláha, V.; Dostál, V.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 1211-1214 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : LOCA * Quenching * Divertor cooling * Heat transfer * Rewetting Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617303733

  16. Current status of accident analysis for Korean HCCR TBS

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Jin, Hyung Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.

  17. Current status of accident analysis for Korean HCCR TBS

    International Nuclear Information System (INIS)

    Ahn, Mu-Young; Jin, Hyung Gon; Cho, Seungyon; Lee, Dong Won; Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin

    2014-01-01

    Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements

  18. APR1400 CEA Withdrawal at Power Accident Analysis using KNAP

    International Nuclear Information System (INIS)

    Lee, Dong-Hyuk; Yang, Chang-Keun; Kim, Yo-Han; Sung, Chang-Kyung

    2006-01-01

    KEPRI (Korea Electric Power Research Institute) has been developing safety analysis methodology for non- LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code. RETRAN code is a non- LOCA safety analysis code developed by EPRI. The new methodology will replace existing CE (Combustion Engineering) supplied codes and methodologies currently used in non-LOCA analysis of OPR1000. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400). The CEA (Control Element Assembly) withdrawal at power accident is one of the 'reactivity and power distribution anomalies' events and the results are typically described in the chapter 15.4.2 of SAR (Safety Analysis Report). The APR1400 has been designed to generate 1,400MWe of electricity with advanced features for greatly enhanced safety and economic goals. The CEA withdrawal at power analysis in APR1400 SSAR (Standard Safety Analysis Report) is analyzed with CESEC-III computer code. In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, CEA withdrawal at power accident is analyzed using RETRAN code and it is compared with results from APR1400 SSAR

  19. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall

  20. A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

    International Nuclear Information System (INIS)

    Ramani, D.T.

    1977-01-01

    The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

  1. MABEL 2: a code to analyse cladding deformation in a loss of coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.; Nye, M.T.S.

    1983-06-01

    The calculation strategy of MABEL-2 and the hierarchy and purpose of its subroutines are described so that a programmer can readily identify both the overall structure of the code and the functions of its constituent parts. Also, to assist those who wish to examine the coding in detail, the common block variables are defined and a list is given of all variables used in the code, together with the subroutines in which they are used. (author)

  2. Investigation of straitified and countercurrent flows in horizontal piping during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bourteele, J.P.

    1980-06-01

    The ECTHOR program consists in a loop having as objective to study the flow regimes in horizontal pipings (stratification, countercurrent flows) in conditions representative of small break transients within commercial PWR. The ECTHOR tests are in process. Experimental results are already available and are presented in this paper: scaling problem, U tube experiments, hot leg experiments, high pressure tests

  3. Definition of loss-of-coolant accident radiation source: summary and conclusions

    International Nuclear Information System (INIS)

    Bonzon, L.L.; Lurie, N.A.; Houston, D.H.; Naber, J.A.

    1978-05-01

    The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as time-dependent beta and gamma-ray energy release rates and spectra which are fundamental quantities that are not specific to a plant design but are generally applicable to any nuclear power station

  4. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  5. System for mitigating consequences of loss of coolant accident at nuclear power station

    International Nuclear Information System (INIS)

    Bukrinsky, A.M.; Rzheznikov, J.V.; Shvyryaev, J.V.; Zlatin, D.A.; Kuznetsov, J.A.; Babenko, E.A.; Tatarnikov, V.P.; Lapshin, A.L.; Sanovich, V.I.

    1981-01-01

    The system according to the invention comprises a first room which accommodates a reactor plant and an active-type sprinkler means. As pressure rises in the first room due to a release of steam from the lost coolant, most of the air contained in this first room is driven out through holes provided in walls of the first room in immediate proximity to a floor of the first room, wherefrom it proceeds to a second room through channels and a basin-type condenser accommodated in the second room. The length of the channels is selected so as to form a water seal in these channels to prevent the back-flow of air from the second room to the first room and thus produce rarefaction in the first room. (author)

  6. An outcome of nuclear safety research in JAERI. Case study for LOCA, FP, criticality and reprocessing

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Ito, Keishiro; Kawashima, Kei; Katsuki, Chisato; Shirabe, Masashi

    2009-09-01

    An outcome of nuclear safety research done by JAERI was case studied by the bibliometric method. (1) For LOCA (loss-of-coolant accident) a domestic share of JAERI in monoclinic research paper was 63% at the past (20) 1978-1982 but was decreased to 40% at the present 1998-2002. For co-authored papers a domestic share between JAERI and PS (public sectors) is almost zero at past (20) but increased to 4% at the present. Research cooperation is active between Tokyo University and JAERI or between JAERI and Nagoya University. (2) Project-type research is to have a large monopolization in papers and that of basic-type research is to have a large development of research networking (DRN). (3) For FP, a share of co-authored paper is high due to an enhanced cooperation among JAERI-PO (Public Organization)-PS. For criticality, research activity was enhanced after JCO accident, especially at NUCEF. (4) For reprocessing, PS had a monopolistic position with a domestic share of 71% and a share of JAERI was about 20%. (5) LOCA and RIA outputs born by NSR-JAERI coincided partly to those of the Safety Licensing Guidelines but a share of contribution done by JAERI was ambiguous due to the lack of necessary information. (author)

  7. LOCA Analysis of KAIST-Micro Modular Reactor with Modified GAMMA+ code

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Bong Seong; Ahn, Yoon Han; Kim, Seong Gu; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The supercritical carbon dioxide (S-CO{sub 2}) power cycle is being seriously investigated around the world due to its simple layout, quite high efficiency around 500 .deg. C turbine inlet temperature, etc. By combining these two ideas, the KAIST research team developed a S-CO{sub 2} cooled SMR, called KAIST-Micro Modular reactor (MMR), which is targeting transportability and electricity supply for remote region. Therefore, requirements of MMR design are factory fabrication of the total system including power conversion system to be transported and air cooling to be independent from the site selection. Until now, steady performances and sizes of components were evaluated. Thus, in this paper a transient performance of the MMR are simulated with special focus on the loss of coolant accident (LOCA) at cold leg pipe. The MMR is a newly suggested innovative small modular reactor concept by the KAIST research team. Since the MMR is cooled by supercritical CO{sub 2}, general safety codes for conventional reactors have limitations. Thus, GAMMA+ code for the transient analysis of a gas-cooled reactor was selected and modified for the S-CO{sub 2} power system. After the modification of GAMMA+ code, LOCA is simulated, which is considered as one of the most limiting accidents in terms of safety of nuclear power plant.

  8. The phenomenology of a small break LOCA in a complex thermal hydraulic loop

    International Nuclear Information System (INIS)

    Di Marzo, M.; Almenas, K.K.; Hsu, Y.Y.; Wang, Z.

    1988-01-01

    A phenomenological description of the thermal hydraulics events that take place during a simulated Small Break Loss of Coolant Accident (SB-LOCA) is presented. The SB-LOCA transient is described in detail and the various mass and energy transport modes are identified. Similar behavior is observed in other facilities designed for the simulation of this type of accidents. Previous investigations suggest a simple modelling of the phenomena based on fluid mechanic considerations. An extensive experimental program conducted at the experimental facility of the University of Maryland reveals that condensation is a dominant driving force for this type of transients. This finding has significant implications in the modelling of enthalpy transport for some of the flow modes which occur during the transient. In particular it affects the Interruption and Resumption Mode (IRM) during which enthalpy is transported by periodic flow of a two phase mixture. The efforts to predict the flow interruption based on fluid mechanic criteria of phase separation in the hot leg are shown to be misdirected since thermodynamic phenomena taking place in the horizontal portion of the cold legs and in the reactor vessel downcomer are mostly responsible for that transition. For flow resumption to occur the liquid-vapor mixture swelling in the vertical portion of the hot leg determines the occurrence of the liquid spill over the top of the candy cane. (orig.)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. Comparison of gamma densitometer detectors used in loss of coolant studies

    International Nuclear Information System (INIS)

    Shipp, R.L.

    1979-01-01

    Ionization chamber type gamma detectors are used in water-steam density measurements in loss of coolant studies at Oak Ridge National Laboratory. Ionization chambers have replaced current-mode scintillation detectors to obtain stability and freedom from magnetic field interference. However, this change results in some loss of fast transient response. Results of studies comparing the transient response of ionization chamber detectors, plastic scintillation detectors, and sodium iodide (NaI) detectors to rapid changes in gamma intensity demonstrate that plastic scintillation detectors have the fastest response and most closely reproduce the transient; ionization chambers have an initial fast response followed by a slower response, which may produce errors in fast transient measurements; and NaI scintillation detectors have a moderately fast initial response followed by an extremely slow response, which produces errors in even slow transient measurements

  11. Requalification of the LOFT reactor following a loss of coolant experiment (Level I)

    International Nuclear Information System (INIS)

    Cannon, J.W.

    1979-01-01

    During a Loss of Coolant Experiment (LOCE), the LOFT reactor experiences an acceleration of 10 G's and fuel cladding temperature changes at a rate of 1100 0 K/sec. These unparalleled conditions present a unique startup problem to the LOFT program: How can the integrity of the fuel be confirmed so as to minimize operation if damage has occurred. The Level I Requalification Program is designed to accomplish this. It is a progressive series of tests, designed to detect damage at the earliest possible time, and thus preclude or minimize operation if damage exists. First, fuel specialists examine the LOCE data for possible damaging conditions and the results of primary coolant sample analysis for signs of failed fuel. Second, the requalification program proceeds to a series of mechanical and physics tests

  12. KUPOL-M code for simulation of the VVER's accident localization system under LOCA conditions

    International Nuclear Information System (INIS)

    Efanov, A.D.; Lukyanov, A.A.; Shangin, N.N.; Zajtsev, A.A.; Solov'ev, S.L.

    2004-01-01

    Computer code KUPOL-M is developed for analysis of thermodynamic parameters of medium within full pressure containment for NPPs with VVER under LOCA conditions. The analysis takes into account the effects of non-stationary heat-mass transfer of gas-drop mixture in the containment compartments with natural convection, volume and surface steam condensation in the presence of noncondensables, heat-mass exchange of the compartment atmosphere with water in the sumps. The operation of the main safety systems like a spray system, hydrogen catalytic recombiners, emergency core cooling pumps, valves and a fan system is simulated in KUPOL-M code. The main results of the code verification including the ones of the participation in ISP-47 International Standard Problem on containment thermal-hydraulics are presented. (author)

  13. Safety assessment of the advanced CANDU reactor in postulated LOCA/LOECC events

    International Nuclear Information System (INIS)

    Hazen Hezhi Fan; Zoran Bilanovic

    2005-01-01

    The Advanced CANDU Reactor TM (ACR TM ) retains the proven strengths and features of CANDU reactors, and incorporates innovative new features and state-of-the-art technology. In addition to the enhanced emergency core cooling system, the reserve water system is designed to be available to inject reserve water by gravity into the reactor inlet headers after a postulated loss-of-coolant accident (LOCA). To assist in the ACR design and analysis of beyond the design basis events, simulations are needed to demonstrate the effectiveness of these two independent systems on core cooling, and to assess the consequences of the postulated accident coincident with the impairment of either of the two systems. The current paper is subject to an assessment of a postulated large LOCA coincident with loss of the emergency core cooling (LOECC) system. A postulated LOCA/LOECC has very low probability, in the range usually associated with severe core damage events. However, in the CANDU design, including ACR, the presence of moderator water surrounding the fuel channels acts as an effective heat sink, together with other safety features, to prevents severe core damage following a postulated LOCA/LOECC. Therefore, it is possible to analyse LOCA/LOECC using the same deterministic tools that are used for analysis of events with much higher frequencies, in the design basis event range. The assessment is conducted based on the current ACR-700 design. However, the analysis methodology, scope, computer tools, and the results in principle, are applicable to larger ACR designs. This assessment includes system (circuit), fuel channel, and fuel analyses. Some assessment results are needed in subsequent moderator analysis and containment analysis. In the assessment, several simulations were performed to analyse the full circuit and individual fuel channel transient behaviours, as well as the fission product release behaviour. The assessment has captured the key responses of the reactor heat

  14. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    Energy Technology Data Exchange (ETDEWEB)

    Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

    2015-11-15

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  15. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    International Nuclear Information System (INIS)

    Urbonavičius, E.; Povilaitis, M.; Kontautas, A.

    2015-01-01

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  16. Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

    Energy Technology Data Exchange (ETDEWEB)

    De, T K; Collins, W M; Holmes, R W [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    CANDU nuclear reactors use D{sub 2}0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs.

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

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

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

  18. Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

    International Nuclear Information System (INIS)

    De, T.K.; Collins, W.M.; Holmes, R.W.

    1995-01-01

    CANDU nuclear reactors use D 2 0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs

  19. BWR 1 % main recirculation line break LOCA tests, RUNs 917 and 918, without HPCS at ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Okazaki, Motoaki; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1988-07-01

    In a case of small break loss-of-coolant accident (LOCA) at a boiling water reactor (BWR) system, it is important to lower the system pressure to cool down the reactor system by using either the high pressure core spray (HPCS) or the automatic depressurization system (ADS). The report presents characteristic test results of RUNs 918 and 917, which were performed at the rig-of-safety assessment (ROSA)-III program simulating a 1 % break BWR LOCA with an assumption of HPCS failure, and clarifies effects of the ADS delay time on a small break LOCA. The ROSA-III test facility simulates principal components of a BWR/6 system with volumetric scaling factor of 1/424. It is experimentally concluded that the ADS delay time shorter than 4 minutes results in a similar PCT as that in a standard case, in which the PCT is observed after actuation of the low pressure core spray (LPCS). And the ADS delay time longer than 4 minutes results in higher PCT than in the standard case. In the latter, the PCT depends on the ADS time, a 220 K higher PCT, for example, in a case of 10 minutes ADS delay compared with the standard case. (author) 52 refs. 299 figs

  20. Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

    International Nuclear Information System (INIS)

    Griggs, D.P.

    1993-03-01

    The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

  1. 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.

  2. 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.

  3. A thermohydraulic analysis for LOCA accident of a CANDU 600 reactor core charged with SEU 43 fuel by means of FIREBIRD code

    International Nuclear Information System (INIS)

    Serbanel, M.; Catana, A.

    2001-01-01

    This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters

  4. Analysis and development of the automated emergency algorithm to control primary to secondary LOCA for SUNPP safety upgrading

    International Nuclear Information System (INIS)

    Kim, V.; Kuznetsov, V.; Balakan, G.; Gromov, G.; Krushynsky, A.; Sholomitsky, S.; Lola, I.

    2007-01-01

    The paper presents the results of the study conducted to support planned modernization of the South Ukraine nuclear power plant. The objective of the analysis has been to develop the automated emergency control algorithm for primary to secondary LOCA accident for SUNPP WWER-1000 safety upgrading. According to the analyses performed in the framework of safety assesment report, given accident is the most complex for control and has the largest contribution into the core damage frequency value. This is because of initial event diagnostics is difficult, emergency control is complicated for personnel, time available for decision making and actions performing is limited with coolant inventory for make-up, probability of steam dump valves on affected steam generator non-closing after opening is high, and as a consequence containment bypass, irretrievable loss of coolant and radioactive materials release into the environment are possible. Unit design modifications are directed on expansion of safety systems capabilities to overcome given accident and to facilitate the personnel actions on emergency control. Safety systems modification according to developed algorithm will allow to simplify accident control by personnel and enable to control the ECCS discharge limiting pressure below the affected steam generator steam dump valve opening pressure, and decrease the probability of the containment bypass sequences. The analysis of the primary-to-secondary LOCA thermal-hydraulics has been conducted with RELAP5/Mod 3.2, and involved development of the dedicated analytical model, calculations of various plant response accident scenarios, conducting of plant personnel intervention analyses using full-scale simulator, development and justification of the emergency control algorithm aimed on the minimization of negative consequences of the primary-to-secondary LOCA (Authors)

  5. Effect of air on speed of insulating material deterioration under simulated LOCA environment. [Gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Kusama, Yasuo; Yagi, Toshiaki; Ito, Masayuki; Okada, Sohei; Yoshikawa, Masato (Japan Atomic Energy Research Inst., Takasaki, Gunma. Takasaki Radiation Chemistry Research Establishment)

    1982-12-01

    To examine the quality approval testing method for the electric cables used for nuclear reactors, various covering insulating materials employed for the cables have been investigated from all angles. The factors which are considered to affect the deterioration of cable materials in a simulated LOCA (loss of coolant accident) environmental test are numerous. This paper reports on the result of investigation on the effect of air on the rate of deterioration of various organic materials usually used as the insulating and covering materials for the cables. Five kinds of polymer sheets (1 mm thick) used for reactor cables were employed as samples. The samples of both standard compounding ratio and the compounding ratio for practical reactor use were tested. As the deterioration prior to LOCA simulation, the thermal deterioration corresponding to 40 years aging (at 121 deg C for 7 days) was given, and subsequently, 50 Mrad gamma -irradiation at 1 Mrad/h was performed in the air. After that, the samples were subject to LOCA simulated environment. Since the results were different according to the kinds of samples, those are described separately for Hypalon, ethylene propylene rubber, cross-linked polyethylene, chloroprene and silicone rubber. The existence of air under LOCA environment accelerated the deterioration of insulation materials except silicone rubber, though its influence differed to the polymers. These materials swelled in the presence of air, and the degree of swelling increased with the temperature, having the close relation to oxidation deterioration. Polyethylene was more susceptible to the effect of air, and silicone rubber was rather stable. The samples of fire-retardant compounding ratio more swelled by water absorption than those of standard compounding ratio.

  6. Simulation of fuel rod behaviour during various break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Gadalla, A.A.; El-Fawal, M.M.

    1996-01-01

    During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

  7. Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S; Streit, R D; Chou, C K

    1980-01-01

    This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10{sup -12}). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

  8. Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

    International Nuclear Information System (INIS)

    Lu, S.; Streit, R.D.; Chou, C.K.

    1980-01-01

    This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10 -12 ). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

  9. LOCA simulation tests in the RD-12 loop with multiple heat channels

    International Nuclear Information System (INIS)

    Ardron, K.H.; McGee, G.R.; Hawley, E.H.

    1985-11-01

    A series of tests has been performed in the RD-12 loop to study the bahaviour of a CANDU-type, primary heat transport system (PHTS) during the blowdown and injection phases of a loss-of-coolant accident (LOCA). Specifically, the tests were used to investigate flow stagnation and refilling of the core following a LOCA. RD-12 is a pressurized water loop with the basic geometry of a CANDU reactor PHTS, but at approximately 1/125 volume scale. The loop consists of U-tube steam generators, pumps, headers, feeders, and heated channels arranged in the symmetrical figure-of-eight configuration of the CANDU PHTS. In the LOCA simulation tests, the loop contained four horizontal heated channels, each containing a seven-element assembly of indirectly heated, fuel-rod simulators. The channels were nominally identical, and were arranged in parallel pairs between the headers in each half-circuit. Tests were carried out using various restricting orifices to represent pipe breaks of different sizes. The break sizes were specifically chosen such that stagnation conditions in the heated channels would be likely to occur. In some tests, the primary pumps were programmed to run down over a 100-s period to simulate a LOCA with simultaneous loss of pump power. Test results showed that, for certain break sizes, periods of low flow occurred in the channels in one half of the loop, leading to flow stratification and sheath temperature excursions. This report reviews the results of two of the tests, and discusses possible mechanisms that may have led to the low channel flow conditions observed in some cases. Plans for future experiments in the larger scale RD-14 facility are outlined. 5 refs

  10. Development of loca calculation capability with relap5-3D in accordance with the evaluation model methodology

    International Nuclear Information System (INIS)

    Liang, T.K.S.; Huan-Jen, Hung; Chin-Jang, Chang; Lance, Wang

    2001-01-01

    In light water reactors, particularly the pressurized water reactor (PWR), the severity of a LOCA (loss of coolant accident) will limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the PCT (peak cladding temperature) evaluation during LOCA, it generally takes more resources to develop. Instead, implementation of evaluation models required by the Appendix K of 10 CFR 50 upon an advanced thermal-hydraulic platform can also enlarge significant margin between the highest calculated PCT and the safety limit of 2200 F. The compliance of the current RELAP5-3D code with Appendix K of 10 CFR50 has been evaluated, and it was found that there are ten areas where code assessment and/or further modifications were required to satisfy the requirements set forth in the Appendix K of 10 CFR 50. The associated models for LOCA consequent phenomenon analysis should follow the major concern of regulation and be expected to give more conservative results than those by the best-estimate methodology. They were required to predict the decay power level, the blowdown hydraulics, the blowdown heat transfer, the flooding rate, and the flooding heat transfer. All of the ten areas included in above classified simulations have been further evaluated and the RELAP5-3D has been successfully modified to fulfill the associated requirements. In addition, to verify and assess the development of the Appendix K version of RELAP5-3D, nine separate-effect experiments were adopted. Through the assessments against separate-effect experiments, the success of the code modification in accordance with the Appendix K of 10 CFR 50 was demonstrated. We will apply another six sets of integral-effect experiments in the next step to assure the integral conservatism of the Appendix K version of RELAP5-3D on LOCA licensing evaluation. (authors)

  11. Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

  12. Comparative analysis of a LOCA for a German PWR with ASTEC and ATHLET-CD

    International Nuclear Information System (INIS)

    Reinke, N.; Chan, H.W.; Sonnenkalb, M.

    2013-01-01

    This paper presents the results of a comparative analysis performed with ASTEC V2.02 and a coupled ATHLET-CD V2.2c /COCOSYS V2.4 calculation for a German 1300 MWe KONVOI type PWR. The purpose of this analysis is mainly to assess the ASTEC code behaviour in modelling of both the thermal-hydraulic phenomena in the coolant circuit arising during a hypothetical severe accident and the early phase of the core degradation versus the more mechanistic code system ATHLET-CD/COCOSYS. The performed analyses cover a loss of coolant accident sequence (LOCA). Such comparison has been done for the first time. The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed since 1996 by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. The thermal-hydraulic mechanistic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by GRS for the analysis of the whole spectrum of leaks and transients in PWRs and BWRs. For modeling of core degradation processes the CD part (Core Degradation) of ATHLET can be activated. For analyses of the containment behavior, ATHLET-CD has been coupled to the GRS code COCOSYS (COntainment COde SYStem). (orig.)

  13. Core heat transfer analysis during a BWR LOCA simulation experiment at ROSA-III

    International Nuclear Information System (INIS)

    Yonomoto, T.; Koizumi, Y.; Tasaka, K.

    1987-01-01

    The ROSA-III test facility is a 1/424-th volumetrically scaled BWR/6 simulator with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA). Heat transfer analyses for 5, 15, 50 and 200% break tests were conducted to understand the basic heat transfer behavior in the core under BWR LOCA conditions and to obtain a data base of post-critical heat flux (CHF) heat transfer coefficients and quench temperature. The results show that the convective heat transfer coefficient of dried-out rods at the core midplane during a steam cooling period is less than approximately 120 W/m 2 K. It is larger than existing data measured at lower pressures during a spray cooling period. Bottom-up quench temperatures are given by a simple equations: The sum of the saturation temperature and a constant of 262 K. Then the heat transfer model in the RELAP4/MOD6/U4/J3 code was revised using the present results. The rod surface temperature behavior in the 200% break test was calculated better by using the revised model although the model is very simple. (orig.)

  14. Mixing phenomena of interest to boron dilution during small break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.; Cheng, Z.

    1995-01-01

    This paper presents the results of a study of mixing phenomena related to boron dilution during small break loss of coolant accidents (LOCAs)in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler-condenser mode. A problem may occur when subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core are examined in this paper. Bounding calculations for boron concentration of coolant entering the core during a small break LOCA in a typical Westinghouse-designed four-loop plant are also presented

  15. Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

    International Nuclear Information System (INIS)

    Ferreira Guimaraes, Antonio Cesar; Franklin Lapa, Celso Marcelo; Lamego Simoes Filho, Francisco Fernando; Cabral, Denise Cunha

    2011-01-01

    Research highlights: → This article presents an uncertainty modelling study using a fuzzy approach. → The AP1000 Westinghouse NPP was used and it is provided of passive safety systems. → The use of advanced passive safety systems in NPP has limited operational experience. → Failure rates and basic events probabilities used on the fault tree analysis. → Fuzzy uncertainty approach was employed to reliability of the AP1000 large LOCA. - Abstract: This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

  16. LOCA assessment experiments in a full-elevation, CANDU-typical test facility

    International Nuclear Information System (INIS)

    Ingham, P.J.; McGee, G.R.; Krishnan, V.S.

    1990-01-01

    The RD-14 thermal-hydraulics test facility, located at the Whiteshell Nuclear Research Establishment, is a full-elevation model representative of a CANDU primary heat transport system. The facility is scaled to accommodate a single, full-scale (5.0 MW, 21 kg/s), electrically heated channel per pass. The steam generators, pumps, headers, feeders and heated channels are arranged in a typical CANDU figure-of-eight geometry. The loop has an emergency coolant injection system (ECI) that may be operated in several modes, including typical features of the various ECI systems found in CANDU reactors. A series of experiments has been performed in RD-14 to investigate the thermal-hydraulic behaviour during the blowdown and injection phases of a loss-of-coolant accident (LOCA). The tests were designed to cover a full range of break sizes from feeder-sized breaks to guillotine breaks in either an inlet or an outlet header. Breaks resulting in channel flow stagnation were also investigated. This paper reviews the results of some of the LOCA tests carried out in RD-14, and discusses some of the behaviour observed. Plans for future experiments in a multiple-channel RD-14 facility, modified to contain multiple flow channels, are outlined. (orig.)

  17. Potential for boron dilution during small-break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.; Cheng, Z.

    1995-01-01

    This paper documents the results of a scoping study of boron dilution and mixing phenomena during small break loss of coolant accidents (LOCAs) in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler condenser mode. A problem may occur when the subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core were examined in this report. A bounding evaluation of the range of boron concentration entering the core during a small break LOCA in a typical Westinghouse-designed, four-loop plant is also presented in this report

  18. 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)

  19. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Bolt, S.E.

    1977-01-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions

  20. 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

  1. 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

  2. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  3. Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

    Full Text Available The loss-of-coolant accident (LOCA simulation in the boiling water reactor (BWR of Laguna Verde Nuclear Power Plant (LVNPP at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

  4. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    Energy Technology Data Exchange (ETDEWEB)

    Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  5. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    International Nuclear Information System (INIS)

    Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

    2017-01-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  6. Identification of NPP accidents using support vector classification

    Energy Technology Data Exchange (ETDEWEB)

    Back, Ju Hyun; Yoo, Kwae Hwan; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    In case of the accidents that happens in a nuclear power plants (NPPs), it is very important to identify its accidents for the operator. Therefore, in order to effectively manage the accidents, the initial short time trends of major parameters have to be observed and NPP accidents have to accurately be identified to provide its information to operators and technicians. In this regard, the objective of this study is to identify the accidents when the accidents happen in NPPs. In this study, we applied the support vector classification (SVC) model to classify the initiating events of critical accidents such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Input variables were used as the initial integral value of the signal measured in the reactor coolant system (RCS), steam generator, and containment vessel after reactor trip. The proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. In this study, the proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. We used an initial integral value of the simulated sensor signals to identify the NPP accidents. The training data was used to train the SVC model. And, the trained model was confirmed using the test data. As a result, it was known that it can accurately classify five events.