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Sample records for break loca analysis

  1. Small break LOCA analysis for Maanshan nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Jer-Cherng Kang; Shou-Chuan Chiang; Lang-Chen Wang [Taiwan Power Company, Taipei (China)

    1994-12-31

    Since 1990, Taiwan Power Company has conducted a LWR LOCA technology transfer program on RELAP5YA computer code from Yankee Atomic Electric Company (YAEC). One objective of this program is to acquire the RELAP5YA computer code from YAEC for Taipower in-house licensing analysis. The RELAP5YA is a computer program developed at YAEC for analysing the dynamic behaviour of thermal-hydraulic systems, and it can cover most of the postulated accidents and transients in light water reactor systems. In this paper, Taipower`s engineers have performed a small break loss of coolant accidents analysis for Maanshan nuclear power plant. Thais action is used to perform the licensing actions for increasing the operation margin on the steam generator tube plugging. The result is shown that the steam generator tube can be plugged slightly without a reduction in safety margins. This analysis covers a spectrum of break size for a small break LOCA. For a complete spectrum of the transient and accident analysis, the large break LOCA and the non-LOCA analysis were performed by the fuel vendor for the reload safety evaluation.

  2. Analysis of large break LOCA in the NPP AP-600: second phase

    International Nuclear Information System (INIS)

    Analysis of large break LOCA in nuclear power plant AP-600 was done by reactor computational simulation using a computer program COBRA IV-I. Large break LOCA is considered as the severest hypothetical accident in the pressurized water reactor. 1/8 symmetrical core is used in the calculation model, and peak cladding temperature is monitored as a LOCA accident criteria. To do this analysis, it was required such system data during the transient condition from the Westinghouse calculation. Calculation results of peak cladding temperature during LOCA is 1500oF, this calculation showed that there is difference <15% with the Westinghouse calculation

  3. Small break LOCA analysis for RCP trip strategy for YGN 3 and 4 emergency procedure guidelines

    International Nuclear Information System (INIS)

    A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3 and 4. The trip setpoint of the first two RCPs for YGN 3 and 4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft2 break size in the hot leg. The analysis results show that YGN 3 and 4 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at worst time. Also, the YGN 3 and 4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3 and 4 can provide improved operator guidance for the RCP operation during accidents. 11 figs., 4 tabs., 9 refs. (Author)

  4. Large break LOCA analysis for retrofitted ECCS at MAPS using modified computer code ATMIKA

    International Nuclear Information System (INIS)

    Full text: Computer code ATMIKA which has been used for thermal hydraulic analysis is based on unequal velocity equal temperature (UVET) model. Thermal hydraulic transient was predicted using three conservation equations and drift flux model. The modified drift flux model is now able to predict counter current flow and the relative velocity in vertical channel more accurately. Apart from this, stratification model is also introduced to predict the fuel behaviour under stratified condition. Many more improvements were carried out with respect to solution of conservation equation, heat transfer package and frictional pressure drop model. All these modifications have been well validated with published data on RD-12/RD-14 experiments. This paper describes the code modifications and also deals with the application of the code for the large break LOCA analysis for retrofitted emergency core cooling system (ECCS) being implemented at Madras Atomic Power Station (MAPS). This paper also brings out the effect of accumulator on stratification and fuel behaviour

  5. The steam-line break inside containment LOCA transient analysis of Lungmen ABWR with TRACE

    International Nuclear Information System (INIS)

    The US NRC is developing TRACE (TRAC/RELAP Advanced Computational Engine), a new thermal-hydraulic code for safety analysis of nuclear power plants. Under the terms of the CAMP (Code Applications and Maintenance Program) contract, the US authority joined most of countries who own nuclear power plants, like Taiwan, together to participate in the research and application of TRACE code. TRACE is a modernized code with the capability to simulate the reactor system and model the thermal-hydraulic phenomena in three-dimensional space. Instead of those out-of-date codes like TRAC and RELAP, TRACE has become the NRC's flagship thermal-hydraulic analysis tool. One of the features of TRACE is its capacity to model the reactor vessel with 3-D geometry. It can support a more accurate and detailed safety analysis of nuclear power plants. TRACE has a greater simulation capability than the other old codes, especially for events like LOCA (Loss of Coolant Accident). Following in Japan's footstep, Taiwan has become the second country who had commenced construction of ABWR (Lungmen nuclear power plant (NPP)). It has two identical units with 3,926 MWt rated thermal power each and 52.2*106 kg/hr rated core flow. The core has 872 bundles of GE14 fuel, and the steam flow is 7.637*106 kg/hr at rated power. There are 10 RIPs in the reactor vessel, providing 111% rated core flow at the nominal operating speed of 1,450 rpm. In this paper, the steam-line break inside containment LOCA transient data from FSAR is used to verify and establish the Lungmen TRACE model. It compares those important thermal parameters at steady state and transient, such as the dome pressure of reactor vessel, steam flow, feedwater flow, and core flow, etc.. It was concluded that the results of TRACE calculations are in agreement with those from FSAR. In summary, our studies concluded that the analysis results trends of Lungmen NPP TRACE model are roughly consistence with FSAR data for the steam-line break inside

  6. PIRT for large break LOCA mass and energy release calculations

    International Nuclear Information System (INIS)

    Pipe ruptures in the primary reactor coolant system are postulated as part of the design basis for containment integrity and equipment qualification validation for Nuclear Power Plants. The mass and energy (M and E) released from a postulated large break LOCA is the primary forcing function used as input for determining the containment response to a LOCA. The current Westinghouse LOCA M and E release calculation methodology was developed in the 1970's, when computing power was limited. The method is somewhat deterministic and includes several simplified, conservative modeling assumptions. Westinghouse is developing a mechanistic LOCA M and E release accident analysis calculation to more realistically, yet conservatively, model the containment response. A good definition of the key LOCA phenomena is needed as part of this development process. The purpose of this document is to discuss the development of the Phenomena Identification and Ranking Table (PIRT) for large break LOCA M and E release calculations. This paper lists the high ranked phenomena from the PIRT, along with the Transient Phase, and Projected Source of Validating Data. This table is the expert opinion of the selected team and is based upon and is an extension of NRC large LOCA PIRT, which was developed as part of the best estimate (BE) LOCA program for ECCS design basis analysis, the Westinghouse large LOCA PIRT developed for the WCOBRA-TRAC BE LOCA model development program, and the Westinghouse large LOCA PIRT, which was developed to address new components as part of the plant development programs

  7. Evaluation of M-RELAP5 code capability for small-break LOCA analysis

    International Nuclear Information System (INIS)

    Mitsubishi Heavy Industries, Ltd. (MHI) has developed M-RELAP5 code to analyze Small-Break LOCA of PWR for licensing calculations. MHI specifically selected RELAP5-3D and modified it as M-RELAP5 in order to meet the requirements in 10CFR Part 50 Appendix K, 'ECCS Evaluation Models'. MHI conducted several analyses for separate effect tests (i.e. ORNL/THTF, FLECHT, CCFL etc.) and for integral effect tests (i.e. ROSA/LSTF, LOFT and Semiscale) to investigate the applicability of M-RELAP5 for the important phenomena (CHF/core dryout, uncovered core heat transfer, core mixture level etc.) under the small-break LOCA, and for the validation against the system effects. The results show that M-RELAP5 well predicts the key phenomena and calculates core heatup conservatively. It is concluded that M-RELAP5 can be applied to the licensing calculation for small-break LOCA of PWR. (author)

  8. Results of small break LOCA analysis for Kuosheng nuclear power plant using the RELAP5YA computer code

    International Nuclear Information System (INIS)

    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)

  9. ALARM-B2: a computer program for analysis of large break LOCA of BWR

    International Nuclear Information System (INIS)

    The computer program ALARM-B2 is a modified version of ALARM-B1 and is a tool to analyse thermo-hydraulic phenomena of BWR during a postulated large break LOCA. The major improvement is to provide one dimensional heat conduction equation, heat transfer correlation package, and point reactor kinetics equation to analyse the heat transfer phenomenon in the core region during a LOCA. Analytical models of the fluid conservation and state equations are the same as in ALARM-B1 code; namely ALARM-B2 solves one-dimensional integral forms of the fluid conservation and state equations under the assumptions common to conventional node-junction type models. The main purpose of this report is to explain the frame-work of ALARM-B2 together with the requirements of input data. The validity of models newly incorporated into the present code is now being examined. (author)

  10. Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

    International Nuclear Information System (INIS)

    The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

  11. Analysis of thermal–hydraulic parameters of WWER-1000 containment in a large break LOCA

    International Nuclear Information System (INIS)

    Highlights: • Evaluation +98 of WWER-1000 containment behavior against LBLOCA accident. • Simulation of WWER-1000 containment by CONTAIN 2.0 code. • Modeling of WWER-1000 containment by a single model. • Validation of results with Bushehr Nuclear Power Plant’s FSAR. - Abstract: The consequences of sever reactor accident depend greatly on containment safety features and containment performance in retaining radioactive material. The specific type of large LOCA is DECL (Double Ended Cold Leg) break which means a total guillotine type of break in cold leg pipe and is one of the most dangerous accidents in the reactor containment. In this paper, thermal–hydraulic parameters (temperature and pressure) of WWER-1000 (Bushehr Nuclear Power Plant) containment in a DECL accident have been simulated by CONTAIN 2.0 code and a single cell model. The containment has been divided to 23 cells in CONTAIN code but for simplicity only one cell has been considered in modeling. The model has been programmed by MATLAB. The accident has been simulated for a short time (initial 200 s) and all of the results have been compared with Bushehr’s Nuclear Power Plant FSAR

  12. Analysis of ROSA-III small-break LOCA experiment RUN 804 by THYDE-B1 computer code

    International Nuclear Information System (INIS)

    THYDE-B1 is a computer code for predicting the thermohydraulic response of the primary system of a BWR during a loss-of-coolant accident (LOCA) aiming at the evaluation of the performance of the emergency core cooling system (ECCS). This code is mainly applied to the analysis of small-break LOCA's with special emphasis on the behavior of the system pressure and the mixture level in the core. Post-test Analysis of a small-break experiment ROSA-III RUN 804 was done for the assessment of the code. ROSA-III facility is a 1/424 scale model of BWR/6 with electrically heated core. RUN 804 was an integral test which simulated a 5% split break at the recirculation pump suction line with ECCS actuation. Sensitivity analyses were also made on important input parameters and models of the code. The temperature rise of fuel rod surface in the experiment was caused by the uncovery of fuel rod to steam from the two-phase mixture. The calculated behavior showed the same trend and the calculated mixture level was in agreement with experimental results. The fuel rod surface temperature after core spray actuation was slightly higher than the experimental results, indicating the need for improvement of heat transfer model. The calculated history of the system pressure showed similar tendency to the experiment, however further examination of the effect of heat loss from the system and of the accuracy of ADS (automatic depressurization system) flow rate in the analysis and experiment is neccesary for better agreement with experimental results. (author)

  13. 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)

    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)

  14. Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

    Energy Technology Data Exchange (ETDEWEB)

    Papini, Davide, E-mail: davide.papini@mail.polimi.i [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Grgic, Davor [Department of Power Systems, FER, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Cammi, Antonio; Ricotti, Marco E. [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy)

    2011-04-15

    Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes). The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.

  15. Break location effects on PWR small break LOCA phenomena

    International Nuclear Information System (INIS)

    The report presents experimental results of a small lower plenum break test of SB-PV-01 conducted at the large-Scale Test Facility (LSTF) of the Rig-of-Safety Assessment (ROSA)-IV program. This test simulates a loss-of-coolant accident (LOCA) caused by instrument tubes break (break area corresponds to 0.5% of the cold leg flow area) in a Westinghouse-type pressurized water reactor (PWR) assuming both manual actuation for all of the high pressure injection (HPI) systems and failure of the auxiliary feedwater systems. The report clarifies long-term system responses, especially the core cooling conditions related to the primary mass inventory. Also it clarifies break location effects on small break LOCA phenomena by comparing other five similar LOCA tests with break locations at cold leg, hot leg, upper head, pressurizer top (TMI-type) and SG U-tubes. It is coucluded that the lower plenum break is the severest on core heatup due to the highest break flow rate and the least primary mass recovery after the ECCS among the six tests. (author)

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

    International Nuclear Information System (INIS)

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

  17. Base input for large break LOCA analysis of commercial PWR with published version of THYDE-P1

    International Nuclear Information System (INIS)

    This report describes input data to be used with the THYDE-P1 interim version SV02L03, which has been published through NEA DATA BANK in April, 1982, and its calculated results. The input data consist of three input data sets, one is for steady state and the following transients and the other two are for restarting, and they are successively used for a through calculation of a large break loss-of-coolant accident (LOCA) of a 1,100 MWe commercial pressurized water reactor (PWR) with ''best estimate'' (BE) options. The major purposes to set up the input data are not only to provide users sample data in publishing THYDE-P1 but also to demonstrate the ability of the published version of THYDE-P1 without any modification to perform a through calculation of a large break LOCA. The results from the present calculation will also be widely utilized as a bench mark in performing sensitivity calculations and further code modifications. In this sense, the input can be called a base input for the published version of THYDE-P1. This report also contains the results from several sensitivity calculations, which show high capability of the version of THYDE-P1 to analyse large break LOCAs. (author)

  18. Thermal-hydraulic analysis for reactor vessel upper-head small break LOCA using SPACE code

    International Nuclear Information System (INIS)

    A small break loss of coolant accident (SBLOCA) in upper-head of a reactor vessel at OPR1000 was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. To assess the capability of SPACE code, upper-head SBLOCA with full plant safeguards was simulated, and compared with results of MARS-KS code. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. Based on the observed thermal-hydraulic features, simulations with the failure of partial plant safeguards were conducted to analyze the safety and performance of OPR1000. Effects of failure to scram and high-pressure safety injection (HPSI) were investigated, and safety assessment was evaluated according to operator actions. Comparative study without any emergency core cooling systems (ECCS) was also conducted to judge the severity of the break location. From the results, this indicated that SPACE code has capabilities to simulate upper-head SBLOCA, and OPR1000 was evaluated to have sufficient safety margin with the application of proper emergency operating procedures.

  19. Thermal-hydraulic analysis for reactor vessel upper-head small break LOCA using SPACE code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [Korea Hydro and Nuclear Power Co., Central Research Inst., Daejeon (Korea, Republic of)

    2015-08-15

    A small break loss of coolant accident (SBLOCA) in upper-head of a reactor vessel at OPR1000 was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. To assess the capability of SPACE code, upper-head SBLOCA with full plant safeguards was simulated, and compared with results of MARS-KS code. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. Based on the observed thermal-hydraulic features, simulations with the failure of partial plant safeguards were conducted to analyze the safety and performance of OPR1000. Effects of failure to scram and high-pressure safety injection (HPSI) were investigated, and safety assessment was evaluated according to operator actions. Comparative study without any emergency core cooling systems (ECCS) was also conducted to judge the severity of the break location. From the results, this indicated that SPACE code has capabilities to simulate upper-head SBLOCA, and OPR1000 was evaluated to have sufficient safety margin with the application of proper emergency operating procedures.

  20. Statistical large-break LOCA analysis for PWRs with combined ECC injection

    Energy Technology Data Exchange (ETDEWEB)

    Seeberger, Gerd-Joachim; Pauli, Eva-Maria; Trewin, Richard; Zeisler, Lars-Peter [AREVA GmbH, Erlangen (Germany)

    2014-04-15

    A statistical analysis methodology based on the code scaling, applicability and uncertainty (CSAU) evaluation approach for predicting the safety margin in case of a postulated large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR) was developed by AREVA. All expected LBLOCA phenomena are listed in the Phenomena Identification and Ranking Table (PIRT) and are prioritized according to their importance on the figure of merit, here the fuel rod peak cladding temperature (PCT). For the high-ranked phenomena parameters are identified, which allow a quantification of the analysis uncertainty. AREVA has updated the PIRT to the state of the art and extended it to the application to pressurized-water reactors with combined emergency core cooling injection of German-type PWRs. This paper describes how the uncertainty distributions, required for a statistical analysis, have been derived and presents the result of an exemplary statistical analysis for a German-type 4-loop plant compared to that of a conservative deterministic analysis. (orig.)

  1. AP1000® Large-Break LOCA BEPU analysis with TRACE code

    International Nuclear Information System (INIS)

    Highlights: • Assessment of AP1000 behavior in LBLOCA sequences. • AP1000 LBLOCA comparison against standard PWR-3L. • TRACE-DAKOTA application to BEPU analysis. - Abstract: The AP1000® is an advanced Pressurized Water Reactor (PWR) design developed by Westinghouse which implements passive safety systems to provide core cooling in case of accident. The development of best-estimate codes produced the evolution of conservative safety analysis towards the so-called best-estimate plus uncertainty (BEPU) analysis in order to obtain more realistic results and larger safety margins. In this sense, Westinghouse used for AP1000 Large Break Loss of Coolant Accident (LBLOCA) the so-called Automated Statistical Treatment of Uncertainty Method (ASTRUM) which was developed to address this kind of BEPU analysis. This paper presents a verification of the AP1000 LBLOCA BEPU analysis by means of TRACE V5.0 patch 2 thermal–hydraulic code with the support of DAKOTA code for uncertainty calculations. The results obtained show lower values for the maximum PCT than the ones obtained by Westinghouse. In both cases the results show that AP1000 can mitigate effectively the occurrence of a postulate LBLOCA and to meet the 10CFR50.46 PCT acceptance criteria with enough margin

  2. Large-break LOCA studies. Computational analysis of clad ballooning and thermohydraulics in a PWR

    International Nuclear Information System (INIS)

    A new multi-pin model of the re-flood phase of a large break loss of coolant accident has been created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. After a brief description of the codes and their linkage, a series of tests to assess the capabilities of the linked codes is described, and their results analysed. It is shown that the current coupled multi-pin code is a stable and reliable tool for ballooning transient analysis. A complete validation process with the simulation of the MT-3 test in the NRU reactor at Chalk River is in progress.(author)

  3. The development of a Realistic LOCA evaluation model applicable to the full range of breaks sizes: Westinghouse full spectrum LOCA (FSLOCA™) methodology

    International Nuclear Information System (INIS)

    Recently changes in the regulatory environment toward a risk informed approach combined with more efficient and demanding fuel power cycles, and utilization of margins put more emphasis in scenarios traditionally defined as Small and Intermediate Break LOCA. As a result, Westinghouse made several upgrades and added several new functionalities to its realistic Large Break LOCA methodology based on the use of the WCOBRA/TRAC code. The new code has been renamed to WCOBRA/TRAC-TF2, for the purpose of extending the Evaluation Model (EM) applicability to smaller break sizes. The new EM is called Westinghouse Full Spectrum LOCA (FSLOCA™) Methodology and is intended to be applicable to a full spectrum of LOCAs, from small to intermediate break as well as large break LOCAs. This paper describes the market and regulatory drivers, the functional requirements for the new evaluation model (EM). An overview of the EM and key conclusions on its applicability to LOCA safety analysis are here summarized. (author)

  4. An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

    Directory of Open Access Journals (Sweden)

    Cesare Frepoli

    2008-01-01

    Full Text Available Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

  5. 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)

    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

  6. A synopsis of experimental activities on small-break LOCA

    International Nuclear Information System (INIS)

    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)

  7. LOCA analysis of the IRIS reactor

    International Nuclear Information System (INIS)

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

  8. Thermal-hydraulic Analysis and Code Assessment for Reactor Vessel Upper-head Small Break LOCA using SPACE code

    International Nuclear Information System (INIS)

    In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper-head as a result of circumferential cracking of a Control Red Drive Mechanism (CRDM) penetration nozzle. Several experimental tests have been performed at the large scale test facility to simulate the behavior of a PWR during an upper-head SBLOCA. Organization for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1 was performed with a break size equivalent to 1.9% cold leg break. Additionally, analysis of an upper-head SBLOCA with high pressure safety injection failed in a Westinghouse PWR was examined taking into account different accident management actions and conditions in order to check the suitability. In this study, the thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 (Optimized Power Reactor 1000 MWe) using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. Inspections of existing nuclear power plants have pointed out the possibility of small break loss of coolant accidents (SBLOCAs) were initiated by a small break located in the upper-head of the reactor pressure vessel. The thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 plant using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. The prediction showed good agreement with the MARS

  9. Thermal-hydraulic Analysis and Code Assessment for Reactor Vessel Upper-head Small Break LOCA using SPACE code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper-head as a result of circumferential cracking of a Control Red Drive Mechanism (CRDM) penetration nozzle. Several experimental tests have been performed at the large scale test facility to simulate the behavior of a PWR during an upper-head SBLOCA. Organization for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1 was performed with a break size equivalent to 1.9% cold leg break. Additionally, analysis of an upper-head SBLOCA with high pressure safety injection failed in a Westinghouse PWR was examined taking into account different accident management actions and conditions in order to check the suitability. In this study, the thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 (Optimized Power Reactor 1000 MWe) using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. Inspections of existing nuclear power plants have pointed out the possibility of small break loss of coolant accidents (SBLOCAs) were initiated by a small break located in the upper-head of the reactor pressure vessel. The thermal-hydraulic analysis was performed for postulated upper-head breaks in OPR 1000 plant using SPACE (Safety and Performance Analysis Code for Nuclear Power Plants) code, which has been developed in recent years by the Korea Hydro and Nuclear Power Company (KHNP). The calculation results were compared with MARS-KS code to assess the capability of the SPACE code to simulate the transient thermal-hydraulic behavior. The prediction showed good agreement with the MARS

  10. Subcooled decompression analysis in PWR LOCA

    International Nuclear Information System (INIS)

    The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such thermo-hydraulic behavior of the coolant during subcooled decompression in PWR LOCA by solving the mass, momentum, and energy conservation equations by the method of characteristics. Detailed studies were made on the transient coolant outflow at the pipe rupture and the effect of frictional loss and heat addition to the coolant on the decompression. Based on the studies, a digital computer code, DEPCO-MULTI, has been prepared and numerical results are compared with the ROSA (JAERI) and the LOFT (NRTS) semiscale test data with various coolant pressures, temperatures, pipe break sizes, and complexity of flow geometry. Good agreement is generally obtained

  11. Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

    International Nuclear Information System (INIS)

    The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft2∼0.02ft2(DVI) 0.5ft2∼0.02ft2(hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA

  12. Computer modelling for LOCA analysis in PHWRs

    International Nuclear Information System (INIS)

    A computer code THYNAC developed for analysis of thermal hydraulic transient phenomena during LOCA in the PHWR type reactor and primary coolant system is described. The code predicts coolant voiding rate in the core, coolant discharge rate from the break, primary system depressurization history and temperature history of both fuel and fuel clad. Reactor system is modelled as a set of connected fluid segments which represent piping, feeders, coolant channels, etc. Method of finite difference is used in the code. Modelling of various specific phenomena e.g. two-phase pressure drop, slip flow, pumps etc. in the code is described. (M.G.B.)

  13. Simulation and analysis of bearing pad to pressure tube contact heat transfer under large break LOCA conditions

    International Nuclear Information System (INIS)

    In some postulated loss-of-coolant accidents (LOCAs) in a CANDU reactor, localized 'hot spots' can develop on the pressure tube as a result of decay heat dissipation by conduction through bearing pad/pressure tube contact locations. Depending on the severity of flow degradation in the channel, these 'hot spots' could represent a potential threat to fuel channel integrity. The most important parameter in the simulation of BP/PT contact is the contact conductance. Since BP/PT thermal contact conductance is a complex parameter which depends upon the thermal and physical characteristics of the material junction and the surrounding environment, contact conductance is determined from experiments relevant to the reactor conditions. A series of twelve full scale integrated BP/PT contact experiments have been conducted at AECL-WRL under CANDU Owner Group (COG). The objective of the experiments was to investigate the effect of BP/PT contact on PT thermal-mechanical behaviour. This paper presents the simulation of BP/PT interaction integrated experiments using SMARTT and MINI-SMARTT computer codes and subsequent derivation of the BP/PT contact conductance by best fitting of the experimental pressure tube temperature measurements. (author)

  14. Uncertainty and sensitivity analysis for a nuclear power plant Large Break Loss of Coolant Accident (LB-LOCA) in the context of OECD BEMUSE programme

    International Nuclear Information System (INIS)

    The second comparative exercise performed as part of BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme, was devoted to Nuclear Power Plant (NPP) application. The BEMUSE programme has been promoted by the Working Group on Analysis and Management of Accidents (WGAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI). The programme was divided into two main steps. The first step was to perform an uncertainty and sensitivity analysis related to the LOFT L2-5 test, and the second step was to perform the same analysis for a Nuclear Power Plant Large Break Loss of Coolant Accident (LB-LOCA). The second step, also known as Phases IV and V, started in May 2006 and was finished in September 2009. Phase IV of BEMUSE Program is connected with previous Phase II, being a necessary step for an uncertainty analysis: the simulation of the reference scenario and sensitivity analysis. The selected plant was Zion 1 NPP, a 4 loop PWR unit. Thirteen participants coming from ten different countries have taken part in the Phase IV of the program. Phase V main objective was to obtain uncertainty bands for the maximum cladding temperature (time trend), upper plenum pressure (time trend), maximum peak cladding temperature (scalar), 1st peak cladding temperature (scalar), 2nd peak cladding temperature (scalar), time of accumulator injection (scalar), time of complete core quenching (scalar). Fourteen groups from twelve organizations and ten countries have participated in BEMUSE Phase V. Both Phases IV and V compared procedures with experience gained in the previous step of the programme. (author)

  15. Large break LOCA experiment at reactor thermohydraulic test loop

    International Nuclear Information System (INIS)

    The experiments of large break LOCA in the reactor Thermohydraulic Test Loop (UUTR) has been done. The experiments were held at hot leg side by use of accumulator safety injection system and without accumulator. Two experiments were done without activating the high and low pressure safety injection system, while make up system was activated until the experiments were stopped. The test were done at 1 MWt power with about 9,4 kg/sec primary coolant flow rate, and pressure 154 bar. The phenomena were only be limited on the effect of accumulator to the system during LOCA. The results of the experiments indicated a similar system depressurization phenomena for both with and without accumulator was activated. The system trip were happened at very closely different time for both at around 10 t h cycle. The significant difference were that in the experiments were the accumulator was activated, the system depressurization was going more slowly than without the accumulator, and the rods and the fluid temperature were more lower. In the case, the water injection system from the accumulator was able to reduced the rod and the cooling temperature as long as the water inventory is available

  16. Progress in realistic LOCA analysis

    International Nuclear Information System (INIS)

    While LOCA is a complex transient to simulate, the state of art in thermal hydraulics has advanced sufficiently to allow its realistic prediction and application of advanced methods to actual reactor design as demonstrated by methodology described in this paper

  17. Progress in realistic LOCA analysis

    International Nuclear Information System (INIS)

    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

  18. Improvement of the PSA model using a best-estimate thermal-hydraulic analysis of LOCA scenarios

    International Nuclear Information System (INIS)

    This study was performed to propose both new success criterion and heading of the event tree by using best-estimate analysis of each LOCA scenario, aiming at the improvement of the PSA models. The MARS code was used for the thermal-hydraulic analysis of LOCA and the Ulchin units 3 and 4 were selected as a reference plant in this study. This study was performed to improve the PSA model of three LOCA scenarios by using best-estimate thermal-hydraulic analysis. The LOCA calculations with various configurations of the safety systems and break sizes were performed. Using the results, we proposed both new success criterion and heading of the small- and middle-break LOCA scenario. The small-break LOCA will be analyzed later in terms of operator actions to depressurize the RCS. The results of this analysis may contribute to improve the PSA model of LOCA. In the probabilistic safety analysis (PSA) of Korean Standard Nuclear Power Plant (KSNP), loss-of-coolant accidents (LOCA) are classified into three scenarios by the break size, such as large-, middle-, and small-break LOCA. The specific break sizes were adopted to identify the boundaries of the three groups in the previous PSA model and the success criteria has been conservatively applied to each state of safety system in the event tree

  19. SPACE code simulation of cold leg small break LOCA in the ATLAS integral test

    International Nuclear Information System (INIS)

    SPACE code is a system analysis code for pressurized water reactors. This code uses a two-fluid and three-field model. For a few years, intensive validations have been performed to secure the prediction accuracy of models and correlations for two-phase flow and heat transfer. Recently, the code version 1.0 was released. This study is to see how well SPACE code predicts thermal hydraulic phenomena of an integral effect test. The target experiment is a cold leg small break LOCA in the ATLAS facility, which has the same two-loop features as APR1400. Predicted parameters were compared with experimental observations. (authors)

  20. Classification of Cold Leg LOCA by Thermal Hydraulic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun; Lim, Ho-Gon; Park, Jin-Hee [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In Level 1 PSA (Probabilistic Safety Assessment), a cold leg LOCA (Loss of Coolant Accident) scenario is significantly considered as an initiating event. The LOCA is formally divided into three groups according as different characteristics of transients which are coming from different break sizes. It came out into the open that the PSA model from traditional grouping of the cold leg LOCA cannot account for the results by the best-estimate TH (thermalhydraulic) code. The one of main issues is that, in small break size LOCA (0.5in-2.0in), only one HPSI (High Pressure Safety Injection) pump satisfies the success criteria in some region of the break size, whereas, bleed operation by SDS (Safety Depressurized System) valve should be used to satisfy the success criteria in the other region of break size. Like this, there are discordances between the present PSA model and TH results for cold leg LOCA. In this paper, TH analyses for the cold leg LOCA are described. Based on the TH results, damage map is illustrated for entire range of break size and characteristics of transient are identified. Using the damage map and characteristics of transient along the break size, recommendations on the re-classification for the PSA model is proposed. In this paper, based on best-estimate TH results for entire range of break size of the cold leg LOCA, specific transient characteristics were identified and four groups re-classification for the cold leg LOCA was suggested.

  1. Classification of Cold Leg LOCA by Thermal Hydraulic Analysis

    International Nuclear Information System (INIS)

    In Level 1 PSA (Probabilistic Safety Assessment), a cold leg LOCA (Loss of Coolant Accident) scenario is significantly considered as an initiating event. The LOCA is formally divided into three groups according as different characteristics of transients which are coming from different break sizes. It came out into the open that the PSA model from traditional grouping of the cold leg LOCA cannot account for the results by the best-estimate TH (thermalhydraulic) code. The one of main issues is that, in small break size LOCA (0.5in-2.0in), only one HPSI (High Pressure Safety Injection) pump satisfies the success criteria in some region of the break size, whereas, bleed operation by SDS (Safety Depressurized System) valve should be used to satisfy the success criteria in the other region of break size. Like this, there are discordances between the present PSA model and TH results for cold leg LOCA. In this paper, TH analyses for the cold leg LOCA are described. Based on the TH results, damage map is illustrated for entire range of break size and characteristics of transient are identified. Using the damage map and characteristics of transient along the break size, recommendations on the re-classification for the PSA model is proposed. In this paper, based on best-estimate TH results for entire range of break size of the cold leg LOCA, specific transient characteristics were identified and four groups re-classification for the cold leg LOCA was suggested

  2. Research on thermal hydraulic behavior of small-break LOCAs in AP1000

    International Nuclear Information System (INIS)

    Highlights: • A RELAP5 model for RCS and passive safety systems in AP1000 was developed. • A spectrum of cold leg small break LOCAs was analyzed. • The PCTs are far below the limit value of 1478 K and meet the safety criterion. • This article is useful for design and operation of AP1000 and other plants. -- Abstract: As a Generation III+ reactor that received Final Design Approval by U.S. NRC, AP1000 employs a series of nature forces, such as gravity, natural circulation and compressed gas, to enhance plant safety. Although plenty of work has been done around AP600 and its updated version AP1000 both experimentally and theoretically in the past few decades, thermal hydraulic behavior of small break LOCAs in AP1000 has not been fully understood and further studies are still required. In the present study, the response of AP1000 to a spectrum of cold leg small break LOCAs is simulated and analyzed using RELAP5/MOD3.4, including 2-in. break, 4-in. break, 8-in. break as well as 10-in. break which approaches the upper limit size for small break LOCAs in AP1000. Based on the calculation results, it indicates that the passive safety systems employed by AP1000, including CMTs, ACCs, IRWST, PRHRS and ADS, combine to provide continuous passive safety injection and residual heat removal. During cold leg small break LOCAs, the core uncovery and fuel heat up do not occur. The peak cladding temperatures (PCTs) during the accident process are far below the Appendix K limit value of 1478 K/2200 °F and meet the safety criterion. Results show that the accidental consequence can be mitigated effectively and thus the safety of AP1000 during cold leg small break LOCAs is proven

  3. Proceedings of the joint CSNI/CNRA workshop on redefining the large break LOCA: technical basis and its implications

    International Nuclear Information System (INIS)

    -informing technical requirement. In the third paper, the European reactor vendor gave its view on LB-LOCA definition for new reactors (EPR). The proposed concept take into account the LB-LOCA (of the main coolant line) for designing the ECCS and the containment but not for mechanical design of the main coolant lines itself. An important prerequisite for LBB and break exclusion in the EPR is also reliable monitoring and inspection. 2. Does adequate technical basis exist to support a redefinition of the LB-LOCA? None of the participants suggested that the probability of LB LOCA could be so high that it represents a significant contribution to the overall risk. There was a general confidence that the probability of a fast occurring large leak from the main reactor coolant circuit can be made insignificant with the right corrective measures. This is true at least in the new plants where lessons learned during the last 30 years have been implemented. The session had a well coordinated set of four complementary presentations aimed at measuring the technical basis to support a redefinition of the LB-LOCA, through the potential development of a spectrum of break sizes, their expected frequencies and the corresponding consequences. With that aim, the four papers presented, in a sequential manner: the critical issues and technical approaches to the subject from the risk requirements point of view, the known and potential aging mechanisms in primary pipes, the technical and administrative developments to prevent pressure boundary fractures through in service inspections and the new developments to detect such fractures through advanced leak detection technologies. The NRC presentation identified issues related to materials engineering, risk considerations, and plant response analysis, and discussed NRC's ongoing technical approaches to address these issues and develop a technical basis for the risk-informed revision of the rule. The EDF presentation was very insightful as it reflected

  4. Analysis for Passive Safety Injection of IPSS in Various LOCAs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sangho; Chang, Soonheung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-05-15

    The Fukushima accident shows US the possibility of accidents that are beyond a designed imagination. Lots of lessons can be shortly summarized into three issues. First of all, the original cause was the occurrence of a Station Black-Out (SBO). Even if engineers considered the possibility of a loss of offsite power enough to be managed, the failure of EDGs seemed to be unnoticed. The second is poor operation and accident management. They could not understand the overall system and did not check the availability of alternating systems. The third is the large release of radioactive materials outside the containment. Even if SBO occurred and the accident was not managed well, all the means must have prevented the large release out of containment. After that, lots of problems were pointed and numerous actions were carried out in each country. The representative proposals are AAC, additional physical barrier, bunker concept and large big tank. Integrated passive safety system (IPSS) was proposed as one of the solutions for enhancing the safety. IPSS can cope with a SBO and accidents with a SBO. IPSS has five functions which are passive decay heat removal, passive safety injection, passive containment cooling, passive in-vessel retention and filtered venting system. The results showed a high performance of removing decay heat through steam generator cooling by forming natural circulation in the primary circuit. The design concept of passive safety injection system (PSIS) consists of the injection line from integrated passive safety tank (IPST) to reactor vessel. The previous works were only focused on a double ended guillotine break LOCA in SBO. The purpose of this paper is to analyze the performance of PSIS in IPSS for various LOCAs by using MARS (Multi-dimensional Analysis of Reactor Safety) code. The simulated accidents were LOCAs which were accompanied with a SBO. The conditions of the LOCAs were varied only for the size of break. It shall show the capability of PSIS

  5. 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.

  6. 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.

  7. An intermediate break BWR LOCA test (RUN 991) at ROSA-III

    International Nuclear Information System (INIS)

    Double failures on the emergency-core-cooling systems (ECCSs) can be resulted in a case of loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) by assuming an ECCS line break and the single failure criterion on another ECCS. In the Rig-of-Safety Assessment (ROSA)-III program, two BWR LOCA simulation tests with intermediate break areas were performed to experimentally study influences of the ECCS double failures on core cooling phenomena. As there was no break unit in the ROSA-III ECCS lines, two break locations were selected above and below the ECCS line elevation. Namely, one is a main steam line (MSL) break test of RUN 992 which was previously reported. Another one is a single-ended jet pump drive line (JPDL) break test of RUN 991. And this break location effect on the system responses was briefly studied in a report of JAERI 1307. This report presents precise experiment results of RUN 991 with respect to the core cooling phenomena related to transient system mass and also presents additional findings on the influences of ECCS double failures in some intermediate break LOCA tests including above two tests. (author)

  8. RELAP5/MOD3.2 Sensitivity Analysis Using OECD/NEA ROSA-2 Project 17% Cold Leg Intermediate-break LOCA Test Data

    International Nuclear Information System (INIS)

    An experiment simulating a PWR intermediate-break loss-of-coolant accident (IBLOCA) with 17% break at cold leg was conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the experiment, core dryout took place due to rapid drop in the core liquid level before loop seal clearing (LSC). Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to counter-current flow limiting (CCFL) by high velocity vapor flow, causing further decrease in the core liquid level. The post-test analysis by RELAP5/MOD3.2.1.2 code revealed that cladding surface temperature of simulated fuel rods was under-predicted due to later major core uncovery than in the experiment. Key phenomena and related important parameters, which may affect the core liquid level behavior and thus the cladding surface temperature, were selected based on the LSTF test data analysis and post-test analysis results. The post-test analysis conditions were considered as 'Base Case', for sensitivity analysis to study the causes of uncertainty in best estimate methodology. The RELAP5 sensitivity analysis was performed by changing the important parameters relevant to the key phenomena within the ranges to investigate influences of the parameters onto the cladding surface temperature. It was confirmed that both constant C of Wallis CCFL correlation at the core exit and gas-liquid inter-phase drag in the core, as parameters that need to consider for the evaluation of safety margin, are more sensitive to the cladding surface temperature than other chosen parameters. (authors)

  9. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

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

  10. Intermediate-break LOCA analyses for the AP600 design

    International Nuclear Information System (INIS)

    A postulated double-ended guillotine break of a direct-vessel-injection line in an AP600 plant has been analyzed. This event is characterized as an intermediate break loss-of-coolant accident (IBLOCA). Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PFl/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. The key processes occurring in an AP600 during a IBLOCA are primary coolant system depressurization, inventory depletion, inventory replacement via emergency core coolant injection, continuous core cooling, and long-term decay heat rejection to the atmosphere. 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 latter phase of the long-term cooling period, the interval when sump injection and containment cooling processes are important

  11. The ANF [Advanced Nuclear Fuels Corporation]-RELAP small-break LOCA [loss-of-coolant accident] analysis for the Comanche Peak steam electric station

    International Nuclear Information System (INIS)

    The system response code RELAP/MOD2 Idaho National Engineering Laboratory cycle 36.02, with modifications developed by Advanced Nuclear Fuels Corporation (ANF), was used to perform small-break loss-of-coolant accident (SBLOCA) calculations for the Comanche Peak steam electric station (CPSES) unit 1. The ability of the ANF-RELAP code to calculate the SBLOCA system response for the four-loop pressurized water reactor is presented by discussing the overall system response, the system mass distribution, and the core response

  12. Responses to the survey on 'redefining the large break LOCA: technical basis and its implications'

    International Nuclear Information System (INIS)

    The Committee on the Safety of Nuclear Installations (CSNI) of the OECD-NEA co-ordinates the NEA activities concerning the technical aspects of design, construction and operation of nuclear installations insofar as they affect the safety of such installations. The Committee on the Nuclear Regulatory Activities (CNRA) of the OECD-NEA co-ordinates the NEA activities concerning the regulation, licensing and inspection of nuclear installations with regard to safety. In December 2002, the CNRA and the CSNI jointly requested the NEA to organize a workshop on 'Redefining the Large Break LOCA: Technical basis and its implications'. The Workshop was held on June 23-24, 2003 in Zurich, Switzerland hosted by HSK (Swiss Federal Nuclear Safety Inspectorate), PSI (Paul Scherrer Institut) and the OECD/NEA. While the Workshop addressed technical aspects, the survey, completed by member countries, gave the participants a clear view on the current regulatory status and issues. The survey was intended to complement the workshop's discussions and provide general background information. It was designed: - To provide material for discussion; - To clearly summarize current national regulations; - To understand rationales and incentives for changing or not the regulation with regard to the Large LOCA; - To list technical issues to be resolved before implementing a new regulation, if any. The workshop was articulated over three questions: - What drives the need to redefine the LB-LOCA? - Does an adequate technical basis exist to support a redefinition of the LB-LOCA? - What are possible new definitions for the LB-LOCA? What are their implications on current and future reactors? This report makes a synthesis and a compilation of the responses to the survey

  13. Application of GRS uncertainty evaluation methodology for ATHLET calculation of the large break LOCA test

    International Nuclear Information System (INIS)

    Six tests devoted to investigation of the integrity of the bubble condenser system of the accident localization of the 3rd unit of the Kola NPP (Nuclear Power Plant) were carried out at the BC (Bubble Condenser) V-213 test facility [1] within the frame of the TACIS (Technical Assistance for the Commonwealth of Independent States) R2.01/99 project in 2003. ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transients) MOD 1.2/CYCLE D code was used for the simulation of the thermalhydraulic processes in the system of high pressure vessels of the facility. The results, obtained by ATHLET code, were used as boundary conditions for the simulation of the processes in the model of the bubble condenser and hermetic compartments, which was performed by the COCOSYS (Containment Code System) code. The GRS (Gesellschaft fur Anlagenund Reaktorsicherheit) uncertainty evaluation methodology [2] based on Wilk's formula is used in this work to evaluate uncertainty of the ATHLET code results for LB LOCA (Large Break Loss-of-Coolant Accident) experiment performed on the facility. (orig.)

  14. Application of GRS uncertainty evaluation methodology for ATHLET calculation of the large break LOCA test

    Energy Technology Data Exchange (ETDEWEB)

    Melikhov, V.I.; Melikhov, O.I.; Parfenov, Y.V. [Electrogorsk Research and Engineering Center for Safety of Nuclear Power Plants (Russian Federation)

    2008-07-01

    Six tests devoted to investigation of the integrity of the bubble condenser system of the accident localization of the 3{sup rd} unit of the Kola NPP (Nuclear Power Plant) were carried out at the BC (Bubble Condenser) V-213 test facility [1] within the frame of the TACIS (Technical Assistance for the Commonwealth of Independent States) R2.01/99 project in 2003. ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transients) MOD 1.2/CYCLE D code was used for the simulation of the thermalhydraulic processes in the system of high pressure vessels of the facility. The results, obtained by ATHLET code, were used as boundary conditions for the simulation of the processes in the model of the bubble condenser and hermetic compartments, which was performed by the COCOSYS (Containment Code System) code. The GRS (Gesellschaft fur Anlagenund Reaktorsicherheit) uncertainty evaluation methodology [2] based on Wilk's formula is used in this work to evaluate uncertainty of the ATHLET code results for LB LOCA (Large Break Loss-of-Coolant Accident) experiment performed on the facility. (orig.)

  15. Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

    International Nuclear Information System (INIS)

    This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

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

    International Nuclear Information System (INIS)

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

  17. Large break LOCA uncertainty quantification for boiling natural circulation reactor using latin hypercube sampling

    International Nuclear Information System (INIS)

    This work deals with uncertainty in Peak Clad Temperature (PCT) for double-ended rupture in the primary coolant circuit of Indian natural circulation reactor. This rupture is identified as a critical break size leading to maximum clad temperature from best estimate code RELAP5. Based on initial sensitivity studies, six important parameters are selected for their significant impact on PCT. Latin Hypercube Sampling (LHS) is used to create 500 sets of six independent variables based on their probability distribution and LOCA calculations have been performed. The 95th percentile value of PCT is 1287 K and it is significantly below the core coolability criteria of 1477 K. (author)

  18. Large break LOCA uncertainty evaluation and comparison with conservative calculation

    International Nuclear Information System (INIS)

    is different to the USA. Significant differences of results are presented between conservative calculations according to the USA Code of Federal Regulation which requires to apply conservative models in conformance with the required and acceptable features of ECCS Evaluation Models, and best estimate plus uncertainty evaluation. Consequently, additional margin to licensing criteria is available by changing from conservative evaluation to best estimate calculations plus uncertainty analysis in the USA. This is not the case in other countries where the use of best estimate computer codes is already a common practice for 'conservative' calculations. However, uncertainty of calculation results is especially important when approaching licensing limits, e.g. due to power u prates. This is the reason why a sub-committee of the German Reactor Safety Commission recently recommended the assessment of uncertainty in calculated results in licensing

  19. LOCA analysis evaluation model with TRAC-PF1/NEM

    International Nuclear Information System (INIS)

    Nowadays regulatory rules and code models development are progressing on the goal of using best-estimate approximations in applications of license. Inside this framework, IBERDROLA is developing a PWR LOCA Analysis Methodology with one double slope, by a side the development of an Evaluation Model (upper-bounding model) that covers with conservative form the different aspects from the PWR LOCA phenomenology and on the other hand, a proposal of CSAU (Code Scaling Applicability and Uncertainty) type evaluation, methodology that strictly covers the 95/95 criterion in the Peak Cladding Temperature. A structured method is established, that basically involves the following steps: 1. Selection of the Large Break LOCA like accident to analyze and of TRAC-PF1/MOD2 V99.1 NEM (PSU version) computer code like analysis tool. 2. Code Assessment, identifying the most remarkable phenomena (PIRT, Phenomena Identification and Ranking Tabulation) and estimation of a possible code deviation (bias) and uncertainties associated to the specific models that control these phenomena (critical flow mass, heat transfer, countercurrent flow, etc...). 3. Evaluation of an overall PCT uncertainty, taking into account code uncertainty, reactor initial conditions, and accident boundary conditions. Uncertainties quantification requires an excellent experiments selection that allows to define a complete evaluation matrix, and the comparison of the simulations results with the experiments measured data, as well as in the relative to the scaling of these phenomena. To simulate these experiments it was necessary to modify the original code, because it was not able to reproduce, in a qualitative way, the expected phenomenology. It can be concluded that there is a good agreement between the TRAC-PF1/NEM results and the experimental data. Once average error (ε) and standard deviation (σ) for those correlations under study are obtained, these factors could be used to correct in a conservative way code

  20. Scientific design of the test facility for the KNGR DVI line small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Park, Choon Kyung; Jun, Hyung Gil; Cho, Seok; Kwon, Tae Soon; Song, Chul Hwa; Kim, Jung Taek

    1999-03-01

    Scientific design of the experimental facility (OASIS) for the KNGR (Korea Next Generation Reactor) DVI line SB-LOCA simulation is carried out. Main purpose of the OASIS is to produce thermal-hydraulic data base for determining the best location of the DVI (Direct Vessel Injection) injection nozzle of the KNGR as well as verifying its design performance in view of the ECCS (Emergency Core Cooling System) effectiveness. The experimental facility is designed based on the Ishii's three-level scaling law. The facility has 1/4 height and 1/341 area scaling ratio. It corresponds to the volume scale of 1/1364. The power scaling is 1/682 and the system pressure is prototypic. The OASIS consists of a core, a downcomer, two steam generators, two pump simulators, a break simulator, a collection tank, primary piping as well as a circulation pump for initial test condition. Each component is designed based on the Ishill's global scaling and boundary flow scaling of mass, energy and momentum. In addition, local phenomena scaling is carried out for the design of major components to preserve key local phenomena in each component. Most of the key phenomena are well preserved in the OASIS. However, the local scaling analysis shows that distortions of the void fraction and mixture level can not be avoided in the core. It comes from the basic features of the Ishill's scaling law in case of the reduced-height simulation. However, it is expected that these distortions will be analyzed properly by a best estimate system analysis code. (Author). 22 refs., 20 tabs., 25 figs.

  1. The 'hot rod methodology' for intermediate break LOCA calculations with the CATHARE2 code

    International Nuclear Information System (INIS)

    The nuclear fuel that will be used for next decades in French PWR may be quite different from current ones, including technological evolutions of cladding alloys, new types of assemblies, etc. Beside these structural evolutions, calculation methodologies are also changing with the use of Best-estimate multi-physics multi-scale coupled calculations, and complex physical models development at 3D local scale. In the frame of its LOCA R and D program, the French 'Institut de Radioprotection et de Surete Nucleaire' is developing its own calculation methodology to figure out the maximum cladding temperature and the maximum oxidation rate reached by the hottest rod in the core. This HOT ROD methodology has been worked out for Intermediate Break LOCA calculations using the 'Best-Estimate' CATHARE2 system code. This calculation is based on a first modeling called 1D SYSTEM calculation which simulates the thermohydraulical behavior of the whole reactor. During this computation, key parameters are stored at core ends in order to be used as boundary conditions for a second calculation called 1D CHAINED calculation nodelizing in one dimension the hottest assembly of the core. Based on experimental evidences, the hot assembly has been considered to have no influence on the thermohydraulical behavior of the mean core. Hence its hydraulical environment has been modified at each calculation's timestep in order to fit with the SYSTEM mean one. This paper details the methodology's validation phase on PERICLES 2D BOIL-UP tests, which have quite the same swollen-level evolution's rate as in the boil up phase of an intermediate break LOCA. (author)

  2. Mixing phenomena of interest to boron dilution during small break LOCAs in PWRs

    International Nuclear Information System (INIS)

    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

  3. Potential for boron dilution during small-break LOCAs in PWRs

    International Nuclear Information System (INIS)

    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

  4. Selected examples of natural circulation for small break loca and some severe accidents

    International Nuclear Information System (INIS)

    In all light water reactors (LWRs), natural circulation is an important passive heat removal system. The March 1979 accident at TMI-2 brought into question the capability of natural circulation cooling remove core decay heat, especially during accident situations. Because natural circulation is expected to be an essential core heat rejection mechanism during certain kinds of accidents or transients in a PWR (e.g., small break LOCAs or operational transients involving loss of pumped circulation), a thorough understanding of natural circulation processes and factors that influence the natural circulation response of the reactor system is necessary. In this paper, natural circulation and related major phenomena are discussed with examples for small break LOCA and severe accident cases, e.g., TMLB station black-out. Descriptions of three modes of natural circulation are provided: Single-phase natural circulation, two-phase natural circulation, and reflux condensation/boiling condensation. The basic phenomena associated with the three types of natural circulation being considered for severe accidents are also addressed: In-vessel natural circulation, hot leg countercurrent flow, coolant loop flows. (author)

  5. LOCA analysis for manganese-stabilized steel

    International Nuclear Information System (INIS)

    Manganese-stabilized steels have been proposed as candidate structural materials for fusion reactors, because they have been perceived as ''low-activation'' materials. Depending on the neutron spectra and the neutron fluence, the decay heat in Mn-stabilized steels is about 3--7 times larger than that in the Ni-stabilized steels. This large amount of decay heat could have serious impact in the case of loss of coolant accident (LOCA). A two-dimensional LOCA model has been used to examine the LOCA temperature response of the manganese steel when utilized in an earlier US design of ITER. The results show that the Mn-steel has approached its melting temperature by less than 100 degree C after about 7 hours from the onset of LOCA. On the other hand, the results for the nickel stabilized steel alloy 316SS show that the maximum temperature reached is 532 degree C in about the same time. 14 refs., 13 figs., 2 tabs

  6. Core liquid level depression due to manometric effect during PWR small break LOCA

    International Nuclear Information System (INIS)

    In the previous study, it is reported that the core collapsed liquid level was depressed nearly to the core bottom and the dryout of the core was observed in the early stage of the PWR cold leg small break loss-of-coolant accident (LOCA) experiment. The manometric effect due to the liquid seal formation in the loop seal and the difference of the liquid holdup between the steam generator (SG) upflow-side and downflow-side caused a depression of the core collapsed liquid level. The core liquid level was recovered just after the loop seal was cleared. The bypass between the core side and the downcomer side affects the core liquid depression. Four 5% cold leg break experiments with the different core bypass location, configuration and size were conducted to clarify the bypass effect. When the bypass was relatively small (less than 3% bypass of the initial core flow before the break), the timing of the loop seal clearing delayed with the bypass. When the bypass was relatively large (9.2% of the core flow), the loop seal clearing took place after the break uncovery and the timing was significantly delayed. In general, the smaller minimum core collapsed liquid level was obtained at the earlier timing of loop seal clearing due to the smaller bypass. (author)

  7. Lessons learned from OECD/CSNI ISP on small break LOCA: final report

    International Nuclear Information System (INIS)

    This document presents an overview of the results obtained from five recent OECD/CSNI International Standard Problems (ISPs) dealing with phenomenologies typical of Small Break LOCA in PWR nuclear power plants of western design. The experiment in four Integral test Facilities, Lobi, Spes, Bethsy and Lstf and the recorded data from a steam generator tube rupture transient in the Belgian PWR of Doel, were taken as reference for the calculations. Relevant hardware characteristics of the facilities and of the plant are firstly given, including the correlation between key thermalhydraulic phenomena and the reference experimental scenarios. A statistical evaluation of the general data connected with each ISP is then presented. The lessons learned from the ISPs are then considered. Four areas have been identified: code deficiencies and capabilities, scaling of the data, progress in code capabilities and various additional aspects

  8. The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

    Science.gov (United States)

    Sabundjian, Gaianê; Andrade, Delvonei A.; Belchior, Antonio, Jr.; da Silva Rocha, Marcelo; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; de Souza Lima, Ana Cecília

    2013-05-01

    This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm2, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

  9. The application of Cathare 1 V1.3 to LOBI small break Loca experiments and a comparison with RELAP5/MOD2

    International Nuclear Information System (INIS)

    The paper presents an overview of the application of CATHARE V1.3 to LOBI Small Break LOCA tests, performed at Dipartimento di Costruzioni Meccaniche e Nucleari of Pisa University. In particular, the development of a new nodalization of LOBI facility is discussed along with the analysis of tests A2-81 (1% CL break). A1-83 (10% CL break) and A1-84 (10% HL break). In the second part of the paper, uncertainties are outlined which are typical of the analysis of experiments in integral test facilities. Finally, on the basis of the application of RELAP5/MOD2 to the analysis of test A2-81, a judgement is given about the behaviour of the two codes emphasizing the related advantages and disadvantages

  10. Recirculation pump suction line 5 % split break LOCA test of ROSA-III

    International Nuclear Information System (INIS)

    This report presents the experimental results of ROSA-III small break (SB) LOCA tests RUN 922 and RUN 932. Both tests assumed 5 % sprit break at the recirculation pump suction. An HPCS was also assumed to be failed to start in both tests. The ADS flow area in RUN 932 is decreased to 50 % of the scaled (1/424) BWR ADS flow area in RUN 922. The test data of RUNs 922 and 932 was compared to investigate the effect of the ADS flow area reduction on the core cooling in BWR SBLOCA. The ADS flow area reduction caused the slow depressurization after the ADS actuation resulting in the ECCS actuation delay and the core uncovery period was extended. The PCT was increased by 116 K from 835 K in RUN 922 to 951 K in RUN 932 also because of the ECCS actuation delay. The effect of the ADS flow area reduction on the core cooling was significant resulting in the PCT increase. However, all the heate rods were quenched with LPCS and LPCI and effectiveness of ADS and low pressure ECCS for core cooling has been confirmed. (author)

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

    International Nuclear Information System (INIS)

    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

  12. Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data

    International Nuclear Information System (INIS)

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and cooperative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. Two RD-14M small break loss of coolant accident (SBLOCA) tests, simulating HWR LOCA behaviour, conducted by Atomic Energy of Canada Ltd (AECL), were selected for this validation project. This report provides a comparison of the results obtained from eight participating organizations from six countries (Argentina, Canada, China, India, Republic of Korea, and Romania), utilizing four different computer codes (ATMIKA, CATHENA, MARS-KS, and RELAP5). General conclusions are reached and recommendations made.

  13. BWR 1 % main recirculation line break LOCA tests, RUNs 917 and 918, without HPCS at ROSA-III program

    International Nuclear Information System (INIS)

    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

  14. Analysis of the flammable gas distribution in containment upper space under LOCA with SAMPSON/HYNA

    International Nuclear Information System (INIS)

    Under the AE scenario, which is the scenario of 6-inch hot leg break LOCA without without ECCS and CV spray, of 3-loop steel-dry containment PWR plant, we evaluated the integrity of the containment vessel through the analysis of spatial distribution of hydrogen and steam in the upper space of containment, using HYNA code, which is one of the modules of SAMPSON code. Hydrogen combustion was found to hardly occur in view of the flammable limit of concentration, even if considering steam condensation. (author)

  15. BWR LOCA integral test simulating a 100 % main steam line break outside reactor containment vessel in ROSA-III program, RUN 955

    International Nuclear Information System (INIS)

    This report presents the ROSA-III experimental results of RUN 955, which simulates a 100 % steam line break (SLB) LOCA outside the BWR reactor containment vessel (RCV) with an assumption of high pressure core spray (HPCS) system failure. The ROSA-III test facility simulates a BWR system with volumetric scale of 1/424 and has the principal systems, i.e., four half-length electrically heated fuel bundles, two active recirculation loops, four types of ECCSs, and steam and feedwater systems. The report clarifies that the 100 % SLB LOCA outside the RCV becomes similar to a small SLB LOCA with the safety/relief valve (SRV) operation after the main steam isolation valve (MSIV) closure and that it is analogous to a small recirculation line break (RLB) LOCA with break area less than 2 % of the scaled pipe flow area. (author)

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

    International Nuclear Information System (INIS)

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

  17. Sensitivity Analyses in Small Break LOCA with HPI-Failure: Effect of Break-Size in Secondary-Side Depressurization

    Science.gov (United States)

    Kinoshita, Ikuo; Torige, Toshihide; Yamada, Minoru

    2014-06-01

    In the case of total failure of the high pressure injection (HPI) system following small break loss of coolant accident (SBLOCA) in pressurized water reactor (PWR), the break size is so small that the primary system does not depressurize to the accumulator (ACC) injection pressure before the core is uncovered extensively. Therefore, steam generator (SG) secondary-side depressurization is necessary as an accident management in order to grant accumulator system actuation and core reflood. A thermal-hydraulic analysis using RELAP5/MOD3 was made on SBLOCA with HPI-failure for Oi Units 3/4 operated by Kansai Electoric Power Co., which are conventional 4 loop PWR plants. The effectiveness of SG secondary-side depressurization procedure was investigated for the real plant design and operational characteristics. The sensitivity analyses using RELAP5/MOD3.2 showed that the accident management was effective for a wide range of break sizes, various orientations and positions. The critical break can be 3 inch cold-leg bottom break.

  18. Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

    International Nuclear Information System (INIS)

    A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

  19. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Diamond, D. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-08-30

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  20. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    International Nuclear Information System (INIS)

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  1. Simulation of power pulses during large break LOCAs in natural and slightly enriched cores in the Embalse NPP

    International Nuclear Information System (INIS)

    In the frame of a joint technical feasibility study between Nucleoelectrica Argentina and Atomic Energy of Canada of using slightly enriched uranium fuel (with 0.9 w% U235) in Embalse NPP, a CANDU-6, loss of coolant accidents (LOCAs) simulations were performed. The power pulse due to two large breaks were simulated: 35% of a Reactor Inlet Header (RIH) and 80% of a Reactor Outlet Header (ROH). For each break size four simulations were performed for different initial conditions o scenarios and for Natural Uranium (NU) and slightly enriched uranium (SEU) cores. The power transients have been simulated using the 3D diffusion, spatial kinetics neutronic program PUMA (developed in Argentina) and the thermal-hydraulics program CATHENA. These codes were coupled by an iterative methodology. The CATHENA thermal-hydraulic simulation results (fuel temperatures and coolant temperatures and densities) were used as input of the PUMA calculation and the time dependent power distribution calculated by PUMA was later applied as input for a new CATHENA calculation. The process was repeated up to convergence. Single channel models were developed to calculate the relevant three key safety parameters: the maximum transient fuel centerline temperature, the maximum transient sheath temperature and the maximum transient stored energy. The main results of power pulse calculation show that the behavior of the SEU core are similar to the NU one. The result of the three safety parameter values show that in the hypothetical large break LOCA occurrence the fuel channel integrity is maintained. The maximum fuel temperature values are lower than the melting temperature of UO2 , the maximum stored enthalpies are lower than the fuel break-up limit and the maximum sheath temperature are lower than Zircalloy fusion temperature. The values of these safety parameters are similar or slightly lower for the SEU core compared with the NU one. (author)

  2. Improvement of the LOCA PSA model using a beat-estimate thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Probabilistic Safety Assessment (PSA) has been widely used to estimate the overall safety of nuclear power plants (NPP) and it provides base information for risk informed application (RIA) and risk informed regulation (RIR). For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs) for the Hanul Nuclear Units 3 and 4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT) were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

  3. Uncertainty analysis for the K-reactor FI-LOCA limits

    International Nuclear Information System (INIS)

    A postulated accident scenario for the Savannah River Site (SRS) K-reactor is a Double Ended Guillotine Break Loss of Coolant Accident (DEGB/LOCA) due to a coolant pipe break at the plenum inlet. The DEGB/LOCA consists of two parts, the first of which applies to the first few seconds of the transient. The first part of the DEGB/LOCA is addressed in this paper. In the first few seconds after the pipe break there is a rapid depressurization of the plenum, which results in a rapid reduction in the core flowrate. Safety rod insertion is not assumed to begin until 1 second after the pipe break and the rods are assumed not to be fully inserted until approximately 2 seconds after the break. The resulting flow-power mismatch results in coolant heating and possible flow disruption via a Lendinegg type flow instability. For this reason, the initial phase of the DEGB/LOCA transient is called the Flow Instability (FI) phase

  4. RELAP5 code validation using a medium-size break LOCA experiment at the PMK-2 test facility

    International Nuclear Information System (INIS)

    For the analyses of loss of coolant accidents (LOCA) the thermohydraulic computer code capabilities for eastern-type reactors like VVER-440 must be validated by pre- and post test calculations of suitable experiments. Such experiments are performed on PMK-2 integral-type test facility in KFKI Atomic Energy Research Institute, Budapest, which is a volume-scaled model of the primary and secondary system of the Paks Nuclear Power Plant. One of these experiments is the pressuriser surge line break which correspond to a 22% leak. The most important phenomena of the experiment are the behavior of hot leg loop seal and the core dry-out with refill-reflood. Posttest calculations were performed by use of the code version RELAP5/mod.3.2. The results of the calculation and experiment are compared. The code properly simulate the analyzed transient.(author)

  5. Simulation of Fuel Behaviours under LOCA and RIA Using FRAPTRAN and Uncertainty Analysis with DAKOTA

    International Nuclear Information System (INIS)

    The Tractebel Engineering’s approach to qualifying the FRAPCON/FRAPTRAN fuel codes for simulation of fuel behaviour during LOCA and RIA accidental conditions is first described, followed by the simulation and uncertainty analysis of an OECD fuel rod codes RIA benchmark case (CABRI RIA test CIP3-1) and an OECD LOCA benchmark case (Halden LOCA test IFA-650.5). Those results showed the importance of the uncertainty analysis of the input parameters and the key models. The perspectives for further model improvements and benchmarks are also discussed. (author)

  6. LOCA pipe break criteria for the design of Babcock and Wilcox nuclear steam systems

    International Nuclear Information System (INIS)

    The document describes the criteria applied by B and W to determine design basis break locations, types of breaks, and break sizes in the primary piping system. Appendixes are provided in support of the basic assumptions made in the development of the criteria

  7. Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements

    International Nuclear Information System (INIS)

    The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors. These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from the

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Technical review of the LB-LOCA analysis for China Qinshan-2 nuclear power plant

    International Nuclear Information System (INIS)

    This report is prepared as a first product of LB-LOCA analysis review program for 600 MWe Qinshan nuclear power plant, which was agreed between NPIC and KAERI. Characteristics of Qinshan plant were identified by comparing relevant parameters with those of Kori 2 and Kori 3/4 which are Westinghouse 2- and 3-loop plants in Korea. Technical review on the LB-LOCA analysis which was performed by NPIC for Qinshan plant using RELAP5/MOD2 code is carried out to examine and ascertain the soundness of evaluation methodology and the properness of calculated LOCA phenomena. This review is expected to provide a guidance to future LOCA analysis which will be jointly accomplished by NPIC and KAERI at KAERI site. (Author) 1 fig., 3 tabs

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

    International Nuclear Information System (INIS)

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

  12. Investigation of break orientation effect during cold leg small-break LOCA at ROSA-IV LSTF

    International Nuclear Information System (INIS)

    The Large Scale Test Facility (LSTF) of the Rig-of-Safety Assessment No. 4 (ROSA-IV) Program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (SBLOCAs) and operational transients. Three 2.5 % cold-leg SBLOCA experiments were conducted at LSTF. In the experiments, the break was oriented at the side, bottom and top of the horizontal cold leg, respectively. The loop seal clearing in the bottom break case was later than in the side break case since a larger amount of liquid had to be discharged until the loop seal clearing. The loop seal clearing in the top break case was later than in the side break case because of the smaller discharge flow rate. The core liquid level drop due to boiloff after the loop seal clearing in the bottom break case was earliest among three cases because of the largest mass loss before the loop seal clearing and in the top break case latest because of the latest occurrence of the loop seal clearing. However, the effect on the system transients such as the pressure and core liquid level transients was small since the transient time was quite long. Analyses to the experimental results were performed with the RELAP5/MOD2 code. Shortcomings in the RELAP5 code calculation results were resolved by reducing the interfacial drag in the hot leg and the core. The inclusion of Shrock's model for the side, bottom and top break of a large horizontal pipe was also tested in the analyses. However, consistent results with data were not obtained since the void fraction in the broken cold leg was not calculated properly. (author)

  13. BWR LOCA simulation test (RUN 992) in ROSA-III program for a 10% main steam line break with ECCS double failures

    International Nuclear Information System (INIS)

    A loss-of-coolant accident (LOCA) caused by pipe rupture at the high pressure core spray (HPCS) line is equivalent to a LOCA with double failures on the emergency core cooling systems (ECCSs) in a boiling water reactor (BWR) system by assuming single failure on another ECCS. This report presents the ROSA-III experimental results of RUN 992, which simulates a 10% main steam line break (MSLB) LOCA with double failure assumption on the HPCS and the low pressure core spray (LPCS) systems. The ROSA-III test facility simulates a BWR system with volumetric scale of 1/424 and has the principal systems, i.e., four half-length electrically-heated fuel bundles, two active recirculation loops, four types of ECCSs, and steam and feedwater systems. The report clarifies effectiveness of ECCS even for this double failure assumption in a 10% MSLB LOCA and also clarifies effects of the automatic depressurization system (ADS) on core cooling. In addition to these, mass balance and mass distribution in the system were investigated to clarify the core cooling condition in the small MSLB LOCA test. (author)

  14. LOCA power pulse analysis for CANDU-6 CANFLEX-RU core

    International Nuclear Information System (INIS)

    The power pulses following a large LOCA are analyzed for CANDU-6 reactor core fuelled with CANFLEX-RU fuel. The coupled simulations for reactor physics and channel thermal-hydraulic phenomena are done using RFSP and CATHENA codes. The 55% pump suction, 35% reactor inlet header and 100% reactor outlet header breaks are selected. The highest power pulse is predicted for 100% reactor outlet header break and it is higher than that for the standard 37-element natural fuel. However, the summation of initial stored energy and transient pulse energy of hottest pin has the minimum 17% margin to the fuel break up. Therefore, it is expected that there is no fuel breakup during the LOCA for CANFLEX-RU core

  15. Analysis of Fourth Stage of Automatic Depressurization System Failure to Open in AP1000 LOCA

    Directory of Open Access Journals (Sweden)

    Zhao Guozhi

    2014-01-01

    Full Text Available Automatic Depressurization System (ADS is a very important part of passive core cooling system in passive safety nuclear plant AP1000. ADS have four stages with each stage having two series and only ADS4 utilizes squib valves. During the accident, emergency core injecting is realized by gravity driven passive safety injection system like makeup tank (CMT, accumulator and In-Containment Refueling Water Storage Tank (IRWST. The objective e of this study is to analyze the system response and phenomenon under part of failure of ADS in AP1000 LOCA. The plant model is built by using SCDAP/RELAP5/MOD 3.4 code. The chosen accident scenario is small and medium LOCAs followed by failure of ADS4 to open, whose location is different from the other three stages. The results indicate that long time core cooling from IRWST is postponed greatly through intentional depressurization only by ADS1, 2, 3. In addition, LOCAs with equivalent diameter 25.4 cm and 34.1 cm will not lead to core melt while 5.08 cm break LOCA will. Meanwhile, high water level in the pressurizer will appear during all of three LOCAs.

  16. Uncertainty analysis of minimum vessel liquid inventory during a small-break LOCA in a B ampersand W Plant: An application of the CSAU methodology using the RELAP5/MOD3 computer code

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission (NRC) revised the emergency core cooling system licensing rule to allow the use of best estimate 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 best estimate code uncertainties. The objective of this work was to adapt and demonstrate the CSAU methodology for a small-break loss-of-coolant accident (SBLOCA) in a Pressurized Water Reactor of Babcock ampersand Wilcox Company lowered loop design using RELAP5/MOD3 as the simulation tool. The CSAU methodology was successfully demonstrated for the new set of variants defined in this project (scenario, plant design, code). However, the robustness of the reactor design to this SBLOCA scenario limits the applicability of the specific results to other plants or scenarios. Several aspects of the code were not exercised because the conditions of the transient never reached enough severity. The plant operator proved to be a determining factor in the course of the transient scenario, and steps were taken to include the operator in the model, simulation, and analyses

  17. Core liquid level depression due to manometric effect during PWR small break LOCA

    International Nuclear Information System (INIS)

    The 10, 5 and 2.5 % cold leg break loss-of-coolant accident experiments were conducted by using the Large Scale Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV program. In the early stage of the 5 % break experiment, the core collapsed liquid level was depressed nearly to the core bottom and the dryout of the core was observed. However, the core liquid level depression without the core dryout was observed in the 10 and 2.5 % break experiments. In the three break experiments, the core liquid levels were recovered just after the loop seal clearing. The manometric effect due to the liquid seal formation in the loop seal and the liquid holdup in the steam generator (SG) U-tubes upflow-side caused a depression of the core collapsed liquid level. The liquid holdup in the U-tubes upflow-side was observed after the termination of the two-phase circulation due to the phase separation at the U-tubes top. The counter current flow limiting (CCFL) and the condensation of steam was considered to be the main reason for the liquid holdup. In the 10, 5 and 2.5 % break experiments, the termination of the two-phase circulation and the loop seal clearing were observed approximately at 40 ∼ 60 % and 30 % mass inventory in the primary system, respectively. (author)

  18. Analysis of PWR cladding transient load under LOCA quench conditions

    International Nuclear Information System (INIS)

    LOCA is a classical design basis accident needed to be considered in all LWR safety analyses. The thermal shock induced by quench during LOCA may cause fracture in the claddings which could lead to core damage. Therefore it is necessary to study the cladding behavior during quench. This paper reports the results of LOCA quench experiments and simulations using the RANNS code for evaluating local mechanical and thermal states of axial load on the cladding. The experimental measurements suggest the rate of load gain decreases with an increasing of the ECR value due to the thicker zirconia layer which serves as a thermal barrier. In addition, the temperature-induced stress on the cladding along the axial direction appears uneven. Therefore, it is found that the LOCA simulation needs multiple elements in the axial direction for obtaining a fairly good prediction of the axial load gain. Finally, the RANNS simulation of the pellet center temperature is validated, and the RANNS code shows the capability in predicting the axial load generated during quench for ECR of lower or equal to 15%. (author)

  19. Large LOCA accident analysis for AP1000 under earthquake

    International Nuclear Information System (INIS)

    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

  20. Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation

    International Nuclear Information System (INIS)

    The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). The ROSA facility simulates a PWR Westinghouse design with a four-loop configuration and a nominal power of 3423 MWth. Two intermediate break loss-of-coolant-accident (LOCA) experiments (Tests 1 and 2) have been carried out during 2010. The two tests were analyzed by using the US-NRC TRACE best estimate code, employing the same nodalization previously used for the simulation of small-break LOCA experiments of the ROSA-1 programme. A post-test calculation was performed for each test along with uncertainty analysis providing uncertainty bands for each relevant time trend. Uncertainties in the code modelling capabilities as well as in the initial and boundary conditions were taken into account, following the guidelines and lessons learnt through participation in the OECD/NEA BEMUSE programme. Two different versions of the TRACE code were used in the analysis, providing a qualitatively good prediction of the tests. However, the uncertainty analysis revealed differences between the performances of some models in the two versions. The most relevant parameters of the two experimental tests were falling within the computed uncertainty bands

  1. Blind Calculation of RD-14M Small Break LOCA Tests by CATHENA Code

    International Nuclear Information System (INIS)

    KAERI participated with the computer code CATHENA, which is used to analyze Pressurized Heavy Water Reactors (PHWRs), in an IAEA International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate thermal-hydraulic computer code against qualified data for Small Break Loss of Coolant Accident (SBLOCA) scenario generated on RD-14M Test Facility. Two specific SBLOCA tests selected for this ICSP titled 'Comparison of HWR Code Predictions with SBLOCA Experimental Data', are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow down. This report presents the blind calculation results for these tests conducted by CATHENA code before the test data are distributed to participants. For B9006 test, CATHENA code simulated all the phases of the transient such as blowdown, high-pressure ECI, secondary pressure ramp, refill, switch from high pressure ECI to low pressure ECI, exponential pump ramp, and natural circulation. For B9802 test, CATHENA calculation was intended to predict temperature rise of the FES sheath due to channel boiling, and power supply trip on high FES sheath temperature (600 .deg. C) process protection trip

  2. Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation

    International Nuclear Information System (INIS)

    The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). The ROSA facility simulates a PWR Westinghouse design with a four-loop configuration and a nominal power of 3423 MWth. Two intermediate break loss-of-coolant-accident (LOCA) experiments (Test 1 and 2) have been carried out during 2010. The two tests were analyzed by using the US-NRC TRACE best estimate code, employing the same nodalization previously used for the simulation of small-break LOCA experiments of the ROSA-1 program. A post-test calculation was performed for each test along with uncertainty analysis providing uncertainty bands for each relevant time trend. Uncertainties in the code modeling capabilities as well as in the initial and boundary conditions were taken into account, following the guidelines and lessons learnt through participation in the OECD/NEA BEMUSE program. Two different versions of the TRACE code were used in the analysis, providing a qualitatively good prediction of the tests. However, both versions showed deficiencies that need to be addressed. The most relevant parameters of the two experimental tests were falling within the computed uncertainty bands. (author)

  3. Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jae Won; Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1998-12-31

    Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous`s empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing. 6 refs., 8 figs. (Author)

  4. Applicability of TASS/SMR using drift flux model for SMART LOCA analysis

    International Nuclear Information System (INIS)

    Highlights: • SMART plant and TASS/SMR code have been developed by KAERI. • TASS/SMR code adopts a drift flux model to consider relative velocity under two-phase condition. • Drift flux model in TASS/SMR is validated using separate effect test results. • Applicability of TASS/SMR using drift flux model for SMART LOCA analysis is confirmed. -- Abstract: Small reactors can apply to local power demands or remote areas. SMART, which can produce 90 MWe of electricity and 40,000 tons/day of sea-water desalination for a 100,000 population city, is a promising advanced integral type small reactor. The thermal hydraulic analysis code with a drift flux model, TASS/SMR, was developed for a conservative simulation of a small break loss of coolant accident in SMART. Taking into account SMART-specific inherent characteristics, the code adopts the Chexal–Lellouche correlation for a drift flux model. The capability of TASS/SMR code is validated using the results of the experimental data. The code predicts conservatively the void distribution compared with the experimental data. TASS/SMR calculation predicts reasonably major phenomena for the SBLOCA and is more conservative than or nearly the same as the results of the best estimated realistic system code, MARS. TASS/SMR code can be used for a SBLOCA analysis of SMART

  5. Dynamic analysis of reactor coolant systems under LOCA conditions

    International Nuclear Information System (INIS)

    The procedures described include structural modeling and analytical techniques for a non-linear time history dynamic analysis of a three dimensional coupled model of the reactor coolant system including details of the reactor internals, pressure vessel, supports and piping. The dynamic analysis is performed to determine the response of the reactor coolant system supports to the simultaneous effects of pipe break thrust and external and internal horizontal and vertical asymmetric pressure loads applied to the reactor vessel and internals as a consequence of the postulated pipe rupture. Condensed structural models are created from highly detailed representations of each component by maintaining response frequency characteristics and interface response compatibility. The location, type and size of the pipe breaks have been determined by stress survey and mechanistic break area criteria. Examples of internal and external forcing functions on the vessel and internals, as well as their separate and combined effects on the reactor vessel supports are illustrated. The pressurized water reactor internals, including the fuel and supporting structures, are suspended from the closure flange region of the reactor vessel and surrounded by a cylindrical 'core support barrel' (CSB). The vessel in turn is surrounded by the biological shield wall. The CSB and reactor vessel are essentially concentric cylinders throughout the length of the CSB. The hydraulic loads applied internal to the vessel are determined by use of qualified thermo hydraulic analysis codes such as those of WHAM, FLASH or RELAP series. The dynamic analysis of the mathematical models subjected to these loads is performed using the STRUDL code to define the characteristics of the structure and the CE program DAGS to calculate the non-linear time history response

  6. Core liquid level depression in 5% small break LOCAs: an investigation using subscale data

    International Nuclear Information System (INIS)

    Core heatup has been shown to be possible, during the course of small break loss-of-coolant accidents (SBLOCAs), even when sufficient inventory is present to fully cover the rods with a two-phase mixture. Such behavior occurs when steam, trapped between the loop seal and the core inventory, moves coolant out of the core barrel 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 (STF) and the 1/1705-scale Semiscale facility. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses conducted using the TRAC-PF1/MOD1 version 12.7 thermal-hydraulic code are also described and summarized

  7. Analysis of break test of 54 cluster fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Mitsuo; Kawamata, Nobuhiro; Kamoshida, Hiroshi

    1998-03-01

    A break test of down pipe and main steam tube of 54 cluster fuels were carried out in the Power Reactor and Nuclear Fuel Development Corporation (PNC) in fiscal 1996. The safety evaluation code for `Fugen` was investigated by analysing the break tests by RELAP 5 code. The tests were carried out by ATR safety experimental facility which was consisted of steam drum, lower header, pressure tube, inlet tube, riser, recirculation pump and non-return valve. Break is modified by breaking a rupture disk in both cases of test. Pressure, pressure difference, temperature, water level and flow rate at channel inlet were measured. The results proved the following: The safety evaluation code for `Fugen` estimated the higher temperature of cladding tube after dry out. A return model of the best evaluation for `Fugen` was confirmed to make reappear dry out and quenching phenomena of temperature behavior of cladding tube under the experimental conditions. RELAP 5 code made a reproduction of heat transfer fluid phenomena of LOCA experiment modifying break of down pipe of Fugen. The result proved that the code is also able to use for LOCA analysis of ATR system. (S.Y.)

  8. Thermal-hydraulic mechanisms of core liquid level depression and recovery during small break LOCA experiment

    International Nuclear Information System (INIS)

    A 5% cold leg break loss-of-coolant accident experiment was conducted by using the Large Scale Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV program. In the early stage of experiment, the core collapsed liquid level was depressed to the bottom of the simulated core and the dryout of the core was observed. The maximum superheat of heater rod surface temperature was about 100 K at the middle of core and the entire core was quenched just after the loop seal clearing. The manometric effect due to the liquid seal formation in the loop seal and the liquid holdup in the steam generator (SG) U-tubes upflow-side caused a depression of the core collapsed liquid level in spite of the coolant injection from the emergency core cooling system (ECCS). The liquid holdup in the U-tubes upflow-side was observed after the termination of the two-phase circulation due to the phase separation at the U-tubes top

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

    International Nuclear Information System (INIS)

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

  10. Analysis of LOCA experiments with RELAP4J code

    International Nuclear Information System (INIS)

    The results of analysis with RELAP4J Code are presented for two typical experiments of cold leg break (Runs 413 and 312), in the ROSA-II (Rig of Safety Assessment II) test program. The objectives of analysis are to evaluate validity of the RELAP4J Code, to improve analytical models and to get a better understanding of experimental phenomena. The two tests were performed under actual reactor initial pressure and temperature, in the respective different LPCI locations. Typical factors influencing the pressure history were examined analytically. In conclusion, the predictions of macroscopic-hydraulic phenomena such as pressure transient in each location are good, and the predictions of microscopic-hydraulic phenomena such as steam-water slip velocity, multi-dimentional flow in plenums or core, quenching velocity, cooling of fuel rods by small coolant flow are not good. Experimental phenomena not clarified yet with test data are predicted with the analysis. (author)

  11. Verification study of LOCA analysis code THYDE-P

    International Nuclear Information System (INIS)

    THYDE-P is a code to analyze loss-of-coolant accidents (LOCA) of the pressurized water reactor (PWR). In this report, the blowdown portion of THYDE-P sample calculation Run 10 is presented along with THYDE-P inputs requirements. Run 10 forms a portion of a series of THYDE-P sample calculations to be performed by the evaluation model option on a specified plant design and is characterized by a simple nodalization such as a single active core node and discharge coefficient 0.6. (author)

  12. UPTF/TEST8B/RUN111, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA

    International Nuclear Information System (INIS)

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

  13. UPTF/TEST8A/RUN112, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA

    International Nuclear Information System (INIS)

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

  14. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    International Nuclear Information System (INIS)

    Highlights: ► Circumferential temperature gradient of PT for asymmetric heat-up was 440 °C. ► At 2 MPa ballooning initiated at 450 °C and with strain rate of 0.0277%/s. ► At 4 MPa ballooning initiated at 390 °C and with strain rate of 0.0305%/s. ► At 4 MPa, PT ruptured under uneven strain and steep temperature gradient. ► Integrity of PT depends on internal pressure and magnitude of decay power. -- Abstract: During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2

  15. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Mukhopadhyay, D., E-mail: dmukho@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2013-02-15

    Highlights: ► Circumferential temperature gradient of PT for asymmetric heat-up was 440 °C. ► At 2 MPa ballooning initiated at 450 °C and with strain rate of 0.0277%/s. ► At 4 MPa ballooning initiated at 390 °C and with strain rate of 0.0305%/s. ► At 4 MPa, PT ruptured under uneven strain and steep temperature gradient. ► Integrity of PT depends on internal pressure and magnitude of decay power. -- Abstract: During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2

  16. RAVE code system for 3-D core non-LOCA accident analysis

    International Nuclear Information System (INIS)

    Analysis Reports (SARs) using the point-kinetics model. The SPNOVA core model is the same as the current core design model used with the ANC code. For fluid solutions, the number of radial and axial nodes in the VIPRE model is typically the same as the SPNOVA code. This results in up to 772 coupled core neutronic and T/H channels for a four-loop plant. The VIPRE fuel rod model uses multiple radial mesh points in the fuel pellets and in the clad for each core node. The fuel pellet-to-clad gap heat transfer accounts for changes in fuel dimensions and fill gas pressure during the transient. The VIPRE code is also used in a separate calculation to determine the hot rod minimum Departure from Nucleate Boiling Ratio (DNBR) and fuel temperature versus time. A code application topical report has submitted for licensing review demonstrating the application to several non-LOCA SAR accidents events. The use of the RAVE code system has also been demonstrated by predicting the core and RCS responses of the Phase III main steam-line break (MSLB) benchmark problem sponsored by the Organization for Economic Cooperation and Development (OECD). The RETRAN, SPNOVA and VIPRE licensing models were adjusted in accordance with the OECD problem specifications. The RAVE-system predicted core and RCS responses were in good agreement with the published results from other participants. The results of the OECD MSLB benchmark problem confirm the robustness of the RAVE code system in analyzing the non-LOCA accidents. (authors)

  17. A comparison of LOCA analysis using SMOKIN and CERBERUS codes

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vojtek, I

    1986-09-01

    The general purpose of this project was the investigation of the heat transfer regimes during the high pressure portion of blowdown. The main attention has been focussed on the evaluation of those phenomena which are most important in reactor safety, such as maximum and minimum critical heat flux and forced convection film boiling heat transfer. The experimental results of the 25-rod bundle blowdown heat transfer tests, which were performed at the KWU heat transfer test facility in Karlstein, were used as a database for the verification of different correlations which are used or were developed for the analysis of reactor safety problems. The computer code BRUDI-VA was used for the calculation of local values of important thermohydraulic parameters in the bundle.

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

    International Nuclear Information System (INIS)

    The general purpose of this project was the investigation of the heat transfer regimes during the high pressure portion of blowdown. The main attention has been focussed on the evaluation of those phenomena which are most important in reactor safety, such as maximum and minimum critical heat flux and forced convection film boiling heat transfer. The experimental results of the 25-rod bundle blowdown heat transfer tests, which were performed at the KWU heat transfer test facility in Karlstein, were used as a database for the verification of different correlations which are used or were developed for the analysis of reactor safety problems. The computer code BRUDI-VA was used for the calculation of local values of important thermohydraulic parameters in the bundle

  20. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

    1995-09-01

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

  1. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    International Nuclear Information System (INIS)

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a open-quotes bleed and feedclose quotes procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation

  2. Additional computational analysis of the WWER-1000 fuel rod thermomechanical characteristics at DB LOCA with extended conservatism of the initial and boundary conditions

    International Nuclear Information System (INIS)

    As a continuation of the earlier presented work, an additional computational analysis of the WWER-1000 fuel rod behaviour under hypothetical large break LOCA conditions was carried out applying reasonably maximum conservative initial and boundary conditions. The same multi-stage approach was used, consisting of: (1) a four-year neutron-physical core simulation and selection of fuel rods, without Gd (FRs, no-Gd FRs) and with Gd burnable absorber (Gd-FRs) with the highest linear heat rate (LHR) and burnup achieved before the hypothetical accident start; (2) thermal-hydraulics simulation of the primary system large break LOCA accident and extraction / preparation of boundary conditions (LHR, coolant temperature and pressure, cladding surface temperature, etc., along the cooling channel) applying the RELAP5/MOD3.3 code, and (3) steady-state and accident thermal-mechanical analysis of the selected fuel rods using the extracted boundary conditions, by means of the TRANSURANUS code. Taking into account the new requirement for a reasonably maximum conservative analysis all selected fuel assemblies (FAs), fuel rods (FRs and Gd-FRs), LOCA simulations and boundary conditions generation were carefully prepared to fulfil this goal. Based on preliminary series of calculations, seven FRs and two Gd-FRs, with the highest LHR and burnup achieved in the course of two model four-year cycles, have been selected and analysed with the TRANSURANUS code. The calculated results were compared to the limiting values of the acceptance criteria for LOCA. The influence of some TRANSURANUS cladding mechanical parameter variation leading to more conservative results has been studied and a new more realistic RELAP5/MOD3.3 / TRANSURANUS boundary condition approach was applied and compared to the more conservative base variant. (authors)

  3. 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)

    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)

  4. Large break loss of coolant accident analysis for Kudankulam nuclear power plant

    International Nuclear Information System (INIS)

    Full text: This paper describes the thermal hydraulic analysis for large break loss of coolant accident (LOCA) for VVER-1000 reactor. VVER is water moderated water cooled 1000 MWe pressurised water reactor with four primary coolant loops. This analysis has been carried out using thermal hydraulic code RELAPS /MOD 3.2. During break in the primary circuit the coolant inventory in the system comes down, primary pressure starts decreasing, coolant circulation through the core decreases. As a result of decrease in the coolant inventory in the primary circuit there is decrease in heat removal from the core, which can lead to rise in clad surface temperature. There will be significant rise in clad temperature before emergency core cooling system is valved in. The analysis predicts thermal hydraulic conditions following large break LOCA. Thermal hydraulic parameters like pressure, temperature, and flow at different locations in the PHT are estimated during the transient. The results have been discussed and compared with the acceptance criteria

  5. Thermal-hydraulic analysis of an intermediate LOCA test at the ROSA facility including uncertainty evaluation

    International Nuclear Information System (INIS)

    Highlights: ► Pre and post test calculations of an IBLOCA test at the ROSA facility are presented. ► Both analyses included an evaluation of the uncertainties. ► The differences between the two uncertainty evaluations are analysed. ► The article highlights the impact of the base case on the final uncertainty band. - Abstract: The goal of uncertainty analyses is to provide best-estimate simulations with uncertainty bands, which take into account uncertainties in the code modeling capabilities as well as uncertainties in the initial and boundary conditions of a given transient scenario. In the present paper, the experience acquired at the Paul Scherrer Institut (PSI) through participations to previous international programs (among these the OECD/NEA BEMUSE program) is used to evaluate a blind calculation of the ROSA-2 Test 1. The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). The ROSA facility simulates a PWR Westinghouse design with a four-loop configuration and a power of 3423 MWth. Test 1 of the ROSA-2 program consists of an intermediate break located in one of the hot legs, in particular it represents the rupture of the pressurizer surge line. Using a TRACE model, previously validated against experiments of the ROSA-1 programme, namely small break LOCA Tests 6-1 and 6-2, a blind case calculation was performed for Test 1/ROSA-2. An uncertainty analysis was carried out together with the simulation of Test 1, in order to provide uncertainty bands for each time trend and finally determine whether the TRACE simulation is able to capture the experimental results within the uncertainty bands. Since the uncertainty bands did not envelop the experimental data, a post-test analysis was carried out. The post-test analysis was helpful in determining which relevant physical phenomena had not been included in the pre

  6. Implementation of a cladding failure model for a Loss of Coolant Accident (LOCA)-analysis in Transuranus

    International Nuclear Information System (INIS)

    To simulate the behaviour of the LWR-fuel rod in a LOCA-transient the entire lifetime of the fuel rod has to be considered. In the German fuel licensing process, the pre-transient life-time of the fuel rod is calculated with Transuranus and the transient part is analyzed by means of different codes that were developed specifically for LOCA conditions. Data transfer from Transuranus to the LOCA-specific code is limited to a few variables. Within these LOCA-specific codes the behaviour of the fuel e.g. fission gas release and thermal-mechanical properties are not treated or only considered in simplified models. In order to refine the calculation it was decided to use Transuranus for both, the pre-transient part and the subsequent LOCA-transient, in one consistent simulation. An additional advantage to integrate a LOCA-transient calculation in Transuranus is to reduce the data amount and efforts on data handling for full core analysis methods. To calculate the fuel rod behaviour during a LOCA with Transuranus one had to implement for the cladding materials Zircaloy and M5 a high temperature creep model, a high temperature corrosion model and a failure criterion model. The high temperature creep model considers the shift in phase transition temperatures and alpha-beta phase fraction due to heating rates. The creep and failure models for LOCA-conditions, the high temperature oxidation model as well as the material properties are fully integrated in the Transuranus code. Model change from the normal to the transient conditions is performed by using the Transuranus restart option. Several test runs have shown the numerical stability of the code. The ongoing validation process of the LOCA-creep models shows so far an excellent compliance with burst tests. (Author)

  7. Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

    International Nuclear Information System (INIS)

    This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

  8. Comparison of LOCA safety analysis in the USA, FRG, and Japan

    International Nuclear Information System (INIS)

    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

  9. TRACE Analysis of LOCA Transients Performed on FIX-II Facility

    OpenAIRE

    Hu, Xiao

    2012-01-01

    As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. In the pr...

  10. Analysis of a convection loop for GFR post-LOCA decay heat removal

    International Nuclear Information System (INIS)

    A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA gas-cooled fast reactor (GFR). The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO2 outdoes helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops. (authors)

  11. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    International Nuclear Information System (INIS)

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

  12. ISP - 26 OECD/NEA/CSNI International standard problem n. 26. Rosa-4 LSTF Cold-Leg Small-Break Loca experiment. Comparison report

    International Nuclear Information System (INIS)

    This report is the final comparison report for the CSNI ISP-26 which was performed for a 5 pc cold-leg small-break LOCA experiment conducted in the ROSA-IV Large scale test facility (LSTF), and simulation of the thermal-hydraulic response of a pressurized water reactor during a small break loss-of-coolant accident or an operational transient. The facility has an overall scaling factor of 1/48, with hot and cold legs sized to conserve the volume scaling. Both the initial steady-state conditions and the test procedures are designed to minimize the effects of scaling compromises. 10 seconds into the break, the turbine throttle valve was closed at a pressurizer pressure of 12.97 MPa. The turbine bypass was inactive, since loss-of-offsite power was assumed concurrently with the scram. The reactor coolant pumps stopped at 265 seconds. The core was uncovered between 120 seconds and 155 seconds into the break. The comparison report presents all the nineteen submitted calculations. A summary description of the participants as well as the computer codes and input decks used by them is provided

  13. A Stylistic Analysis of Break,Break,Break

    Institute of Scientific and Technical Information of China (English)

    李瑶

    2015-01-01

    Break, Break, Break is a poem by Alfred Lord Tennyson, the Poet Laureate during the Queen Victoria's reign. This exquisite little poem is wel known for the poet's grief-stricken feelings and heart-broken emotions over the premature death of his best friend, Arthur Henry Halam. Most of the previous studies on this poem focus on the emotional level to consider it as an elegy, expressing sorrow and lamentation for the death of a particular person. However, in order to have a deep understanding in general, this paper analyzes the poem based on the stylistic theory, concerning on the lexical level and the semantic level. It aims at helping the readers to cultivate a sense of appropriateness, to sharpen the understanding and appreciation of literary works and to achieve adaptation in translation.

  14. Size and number density change of droplet populations above a quench front during reflood in a PWR after a large break LOCA

    International Nuclear Information System (INIS)

    The dispersed flow regime above a quench front during reflood in a PWR after a large break LOCA, cannot be described adequately by employing a single averaged drop size. This is due to the fact that two distinct droplet generation mechanisms are present (Fragmentation of the bubble film and hydrodynamic) leading to a bi-spectral droplet distribution. The evaporation rate of these droplet populations is analyzed by taking into account the relevant convective and radiative heat transfer mechanisms. It is found that the droplet population changes significantly between the point of creation of the droplets and the next channel restriction. The rapid evaporation and diminishing of the small droplet spectrum could explain the exponential rate of energy transfer from the cladding surface in the vicinity above the quench front

  15. Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

    Science.gov (United States)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    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 the 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 reasons 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.

  16. Replacement divider plate performance under LOCA loading

    Energy Technology Data Exchange (ETDEWEB)

    Huynk, H.M. [Quebec Hydro, Montreal, PQ (Canada); MClellan, G.H.; Schneider, W.G. [Babcock and Wilcox, Cambridge, ON (Canada)

    1997-07-01

    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)

  17. MELCOR code analysis of a severe accident LOCA at Peach Bottom Plant

    International Nuclear Information System (INIS)

    A design-basis loss-of-coolant accident (LOCA) concurrent with complete loss of the emergency core cooling systems (ECCSs) has been analyzed for the Peach Bottom atomic station unit 2 using the MELCOR code, version 1.8.1. The purpose of this analysis is to calculate best-estimate times for the important events of this accident sequence and best-estimate source terms. Calculated pressures and temperatures at the beginning of the transient have been compared to results from the Peach Bottom final safety analysis report (FSAR). MELCOR-calculated source terms have been compared to source terms reported in the NUREG-1465 draft

  18. A study on the effect of the CHF correlations to the LOCA analysis

    International Nuclear Information System (INIS)

    The critical heat flux (CHF) is a major parameter which determines the cooling performance and therefore the prediction of CHF is of importance for the design and safety analysis in boiling systems; such as nuclear reactors, conventional boilers, and other various two-phase flow systems. Until now, many CHF correlations have been developed and for the actual design a correlation has been selected in consideration of its characteristics. For the analysis of Loss of Coolant Accident (LOCA) in a Nuclear Power Plant, which shows the drastic parameters change during the system transient, a correlation having a reasonable degree of accuracy over a wide range is preferred, rather than that having accuracy for a specific range. It is required to have tangible insight about the effects of the CHF correlation to the LOCA analysis for the purpose of computer code development and nuclear regulation. The related research is further recommended. The purpose of this research is to obtain an insight and/or intuition about the above effect and to evaluate the selected CHF correlations. To achieve these purposes LOCA is analysed for the UL-JIN 3 and 4 nuclear power plant, the Korea Standard Type Nuclear Power Plant and the Loss of Flow Test (LOFT) L2-5 experiment is simulated using the RELAP5/MOD3.1 computer code for each selected CHF correlation. The selected correlations are the AECL-UO Lookup Table, adapted in RELAP5 code; the K110 CHF correlation, developed by KAERI; and the original W-3 CHF correlation, developed by L.S. Tong. LOFT is also simulated using the AECL-UO Lookup Table having the CHF multiplication factors 0.5 and 1.5, and then compared with the result of the original Lookup Table and the experiment result. In the LOCA analysis, the CHF correlations affect the magnitude of peak cladding temperatures, but does not seriously affect the occurrence points of time. The effect of each CHF correlation to the fuel cladding temperature behavior becomes apparent at the end of

  19. Coldleg Loss of Coolant Accident (LOCA) Analysis of the Modified Reactor Thermohydraulic Test Facility Using CATHENA Computer Code

    International Nuclear Information System (INIS)

    A LOCA analysis at coldleg of the reactor thermal hydraulic test facility using CATHENA computer code has been completely conducted. The analysis was performed by modeling the reactor thermal hydraulic test loop into generic models of the CATHENA such as PUMP, VALVE, VOLUME, ACCUMULATOR, TANK, RESERVOIR, DISCHARGE, GENERALIZED TANK, and GENHTP. The primary system was simulated at power of 1 MWatt with pressure and mass flow at 15.5 MPa and 9.4 kg/s respectively. At secondary side, feedwater flowed at 15.0 kg/s with temperature of 27 oC and pressure of 0.8 MPa. The calculation showed that during steady state the inlet and outlet temperature of test section were 121 oC and 146 oC. After calculating steady state condition, the calculation was followed by transient calculation. The transient was triggered by pipe break at coldleg with diameter of the break was 2 mm. Due to this break, the pressure decreased dramatically. When the pressure reached 4.2 MPa, the accumulator started supplying water into the system. A moment later, the pump was also tripped because of the continuing pressure drop that reached 4.0 MPa. As a consequence the coolant flow was also dropped At the coolant 40 % of normal flow, the power of heated rods then shut down. The result of calculation showed that during the transient, the maximum coolant temperature was 159 oC and the maximum temperature of heated rods was 223 oC. Based on these results, it can be concluded that during the transient, the heated rods were not in danger. (author)

  20. Preliminary Results Of LOCA Problem For APR1400 Reactor

    International Nuclear Information System (INIS)

    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 m2 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)

  1. Computation programs for the thermofluidodynamic transient analysis in the containment system following a LOCA

    International Nuclear Information System (INIS)

    This works briefly describes the features of the computation codes available at the Istituto di Impianti Nucleari of the Pisa University for the analysis of the thermofluidodynamic transient in the containment system of a nuclear power plant following a LOCA (RELAP 4/MOD.S, COMPARE, FUMO and CONTEMPT-LT/026). More details are contained in the Annex. Particular attention has been devoted to the opportunity to study, through the computation codes, the effects of the sub division of a full pressure containment system

  2. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

  3. Large break loss-of-coolant accident analysis for China Qinshan-2 nuclear power plant

    International Nuclear Information System (INIS)

    Large break LOCA analysis for China Qinshan-2 nuclear power plant has been performed using realistic evaluation model which has been being developed by KAERI. RELAP5/MOD3/KAERI code, which is a modified version of RELAP5/MOD3, is coupled with CONTEMPT4/MOD5 and is used as a best estimate code to predict the thermal hydraulic behavior of the system. PCT uncertainty which stems from code uncertainty, plant application uncertainty, scaling uncertainty and PCT bias are discussed. Among them, plant application uncertainty is described in detail. The licensing PCT is calculated by adding all the uncertainties to the best-estimate PCT. The result indicates the Qinshan-2 nuclear power plant has at least 37 deg C safety margin for large break LOCA. (Author) 10 refs., 47 figs., 14 tabs

  4. 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)

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

    International Nuclear Information System (INIS)

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

  6. SCDAP/RELAP5 analysis of station blackout with pump seal LOCA in Surry plant

    International Nuclear Information System (INIS)

    During a station blackout of PWR, the pump seal will fail due to loss of the seal cooling. This particular transient-LOCA sequence designated as S3-TMLB' analyzed by SNL with MELPROG/TRAC for Surry plant showed that the depressurization due to the pump seal LOCA would result in early accumulator injection and subsequent core cooling which lead to the delay of reactor pressure vessel (RPV) meltthrough. The present analysis was performed with SCDAP/RELAP5 to evaluate this scenario shown in the MELPROG/TRAC analyses. Additionally, the calculated results were compared with the similar experimental studies of JAERI's ROSA-IV program. The present analyses showed that: (1) During S3-TMLB', the loop seal clearing would occur and cause a slight delay of accident progression. (2) It is unlikely that the accumulator injection, which leads to the delay of RPV meltthrough by approximately 60 min, is initiated automatically during S3-TMLB'. Accordingly, an intentional depressurization using PORVs is recommended for the mitigation of the accident consequences. (3) The present SCDAP/RELAP5 analyses did not show significant delay of accident progression. It was found that non-realistic lower heat generation and higher core cooling models used in the MELPROG/TRAC analysis are attributed to this discrepancy. (author)

  7. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  8. An analysis of LOCA sequences in the development of severe accident analysis DB

    International Nuclear Information System (INIS)

    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)

  9. Application of the statistical safety evaluation method to the small break LOCA with high pressure injection failure. Sensitivity analyses to determine the break conditions

    International Nuclear Information System (INIS)

    By applying a statistical safety evaluation method, the uncertainties of best estimate results can be estimated quantitatively, and as a consequence, excessive conservatism can be reasonably removed to obtain evaluation results with enhanced reliability. Application of a statistical evaluation method is being made to analyses of the “low pressure injection by intentional depressurization of the steam generator secondary side” which is an accident management approach in a SBLOCA (small break loss-of-coolant accident) with HPI (high pressure injection) failure. At the time of a SBLOCA, the break conditions such as the break size are important parameters since they influence PCT (peak cladding temperature). In this research, sensitivity analyses about the break size, direction and position were carried out for a system plant under a condition which the start timing of the steam generator secondary side intentional depressurization is severer than an actual abnormal operating condition. From the result of the sensitivity analyses, differences in the phenomena progression which change depending on the break conditions were evaluated, and a 3 inch facing-down break of the cold-leg was determined as the base case of a statistical safety evaluation. (author)

  10. Preliminary analysis of downcomer effective water head during reflood phase in PWR LOCA

    International Nuclear Information System (INIS)

    The results are described of preliminary analysis on the downcomer effective water head, which is the driving force to feed the emergency coolant into the core during the reflood phase in PWR LOCA. Due to the vapor generated by heat release from the downcomer walls (reactor vessel etc.), a two-phase flow appears in the downcomer, so that the downcomer effective water head is reduced. Taking into consideration thermo-hydraulic characteristics of the downcomer, the sensitivities were studied concering factors influencing the effective water head, and the conceptual histories were obtained of the downcomer effective water head during the reflood phase. Geometrical scalings, run parameters and test procedures were also obtained on the basis of the results of preliminary analysis. (auth.)

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

    International Nuclear Information System (INIS)

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

  12. Development and qualification of the LOCA analysis system CUPIDON-DEMETER

    International Nuclear Information System (INIS)

    As a support to the power reactor safety analysis, codes describing PWR-type fuel behaviour under LOCA conditions have been developed by CEA under the sponsorship of IPSN. These codes are CUPIDON for unirradiated single fuel rod and DEMETER devoted to the prediction of the behaviour of a whole assembly. An extensive qualification program is underway. The main models: heat transfer, zircaloy oxidation and cladding deformation and rupture, have been separately benchmarked using PHEBUS experiments, German and Japanese transient tests and EDGAR transient tests. Qualification with global tests FLASH and PHEBUS is now underway. From the first calculations, it is concluded that CUPIDON gives a fair account of the behaviour of the rods tested in FLASH and PHEBUS

  13. Data report for ROSA-IV LSTF 5% cold leg break LOCA experiment Run SB-CL-08

    International Nuclear Information System (INIS)

    Experimental data for the 5% cold leg break test, Run SB-CL-08, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on July 10, 1986, are presented. This test was part of a test series which addressed the steam generator (SG) liquid holdup effects on the core liquid level depression associated with loop seal clearing during a cold leg small-break loss-of-coolant accident (SBLOCA) in a pressurized water reactor (PWR). The test assumed total failure of both high pressure injection (HPI) and auxiliary feedwater systems. Core uncovering occurred twice in the test. The first level drop which occurred during loop seal clearing was enhanced by the asymmetry of liquid holdup between the SG upflow and downflow sides. The core was uncovered temporarily, nearly to the core bottom, but was recovered as soon as loop seals were cleared. The second core uncovering occurred due to vessel coolant inventory boil-off and was terminated by an automatic actuation of accumulators. (author) 8 tabs. 94 figs. 10 refs

  14. ROSA-III tests on BWR pump suction-line 200% break LOCAs with partial and total ECCS failure

    International Nuclear Information System (INIS)

    This report presents experiment data for three recirculation-pump suction-line 200% (double-ended) break tests, RUNs 902, 905 and 924, conducted in the ROSA-III test facility, a 1/424 volumetrically-scaled simulator of a boiling water reactor (BWR). These three tests, together with two other tests that have been already reported elsewhere (RUNs 926 and 901), make a set of five tests which addressed the effects of ECCS failure modes on core cooling performance for this break geometry. Results from these five tests are compared in this report. RUNs 902, 924 and 926 simulated three different single-failure modes of emergency core cooling system (ECCS) associated with failure of one out of the three emergency diesel generators (DGs) in the BWR. RUN 926 simulated DG failure disabling the high pressure core spray (HPCS) pump; RUN 924 the low pressure core spray (LPCS) pump and one out of the three low pressure coolant injection (LPCI) pumps; and RUN 902 two out of the three LPCI pumps. RUN 905 simulated hypothetical total failure of ECCS, and RUN 901 a fully-functional ECCS. For all the ECCS DG single-failure tests (RUNs 902, 924 and 926), the measured peak cladding temperatures (PCTs) were considerably lower than the licensing criterion of 1473 K. The HPCS failure (RUN 926) resulted in the severest core heatup among these three tests. (author)

  15. ROSA-IV/LSTF 5% cold leg break LOCA experiment run SB-CL-18 data report

    International Nuclear Information System (INIS)

    This report presents the experimental data obtained for 5 % cold leg break test with the assumption of high pressure injection system (HPIS) failure, Run SB-CL-18, conducted at the Large Scale Test Facility (LSTF) of the RCSA-IV program. In the test, core uncovery was observed twice. The first core uncovery occurred during loop seal clearing. The core uncovery was amplified by the manometric effect caused by imbalance in the coolant holdup in the steam generator (SG) U-tubes and SG plena between the upflow and downflow sides. The peak cladding temperature (PCT) in the test was observed during this temporary core uncovery just before the loop seal clearing. The second core uncovery occurred due to core boil-off; however, the core cooling was recovered after automatic actuation of the accumulators (ACC). This report includes all the data for the test. The experimental data are presented in engineering units. (author)

  16. Severe accident analyses of a BWR with MAAP5 code. Station blackout and large-break LOCA

    International Nuclear Information System (INIS)

    Calculations were performed for a station blackout (TBU) sequence and a large-break loss-of-coolant accident (AE) sequence of a typical BWR-5 plant with modified Mark-II type containment by the MAAP5 code. The core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment including hydrogen production were investigated. Sensitivity analyses focusing on direct contact heating (DCH) and zirconium oxidation, which affect on the consequences of severe accidents, were also performed. If extensive DCH does not occur in the TBU sequence, failure of the containment vessel can be postponed. On the other hand, concrete ablation at a floor and a side wall in the pedestal due to molten core - concrete interaction (MCCI) significantly increases, because a large amount of debris with high temperature stays inside the pedestal. Although the hydrogen production is affected by the zirconium oxidation model, the differences of hydrogen production are within ± 10% in the case of TBU sequence. (author)

  17. Recirculation pump suction line 1 % split break LOCA test of ROSA-III (RUNs 921 and 931 with HPCS failure)

    International Nuclear Information System (INIS)

    This report presents the experimental results of ROSA-III SBLOCA tests RUN 921 and RUN 931. Both tests assumed 1 % sprit break at the recirculation pump suction. An HPCS was also assumed to be failed to start in both tests. The ADS was actuated 77 s earlier in RUN 931, when the liquid level in the downcomer decreased to L2 level with a time delay of 120 s (L2 + 120 s), than in RUN 921 (L1 + 120 s). The test data of RUNs 921 and 931 was compared to investigate the effect of the ADS early actuation on the core cooling in BWR SBLOCA. The ADS early actuation caused the early core uncovery behavior as a whole. The PCTs in two tests were observed at the same location, midplane (position 4) of the peak power rod A87, and were almost the same value each other, 751 K for RUN 921 and 765 K for RUN 931, because the uncovered period of the core midplane was almost the same in two tests. The ADS early actuation resulted no significant difference in the thermal hydraulic behavior. The mixture level in the downcomer in both tests stayed relatively above the upper tie plate in the core even when the whole core was uncovered. Almost all heater rods were quenched with only LPCS before the LPCI actuation and the effectiveness of the low pressure ECCS for core cooling has been confirmed. (author)

  18. Comparative analysis of a LOCA for a German PWR with ASTEC and ATHLET-CD

    International Nuclear Information System (INIS)

    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.)

  19. Comparative analysis of a LOCA for a German PWR with ASTEC and ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Reinke, N.; Chan, H.W.; Sonnenkalb, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2013-07-01

    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.)

  20. 小型动力堆码头中破口失水事故大气扩散研究%Atmospheric Dispersion Research of Mid-break LOCA for Small Reactor of Nuclear-powered Device in Dock

    Institute of Scientific and Technical Information of China (English)

    王伟; 张帆; 陈力生; 晏峰

    2014-01-01

    Using the model of Gaussion subsection plume ,the radioactive nuclide atmos‐pheric dispersion rule in the terrain of 20 km of the coastal w as estimated w hen the design basis accident with 29.4% equivalent diameter break size happened .The source term was captured by the calculation program of severe accident named MELCOR ,and the result was used as input in the analysis software of atmospheric dispersion named MACCS .The results show that the mid‐break LOCA would lead to the radioactive pol‐lution for the area of dock .The slower the wind blows and the more steady the weather is ,the larger the radioactive polluted area is .%采用高斯分段烟羽模型估算了某小型动力堆在码头内发生破口尺寸为29.4%当量直径的设计基准事故时,放射性核素在码头20 km 区域范围内的大气扩散规律。源项采用严重事故计算程序M ELCOR仿真获得,并将计算结果输入到大气扩散分析软件M ACCS进行分析计算。计算结果表明:中破口失水事故会造成码头区域的放射性污染,风速越小、气象条件越稳定,放射性的影响范围越大。

  1. RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

    International Nuclear Information System (INIS)

    The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

  2. System Response Analysis of Rod Ejection Accident for OPR1000 Using Korea Non-LOCA Analysis Package

    International Nuclear Information System (INIS)

    Korea Electric Power Research Institute (KEPRI) of Korea Electric Power Corporation (KEPCO) has been developed the non-loss-of-coolant accident (non-LOCA) analysis methodology, called as the Korea Non-LOCA Analysis Package (KNAP), for the typical Optimized Power Reactor 1000 (OPR1000) plants. The RETRAN hot spot model (HSM) of KNAP has been contrived to replace the functions of STRIKIN-II code of ABB-CE, which is used for the rod ejection accident (REA) analysis. The HSM could be used to estimate the fuel temperature, fuel enthalpy, cladding surface temperature, etc., which are used to confirm the safety limits of REA. In this work, to estimate the feasibility of HSM, the typical cases of REA were analyzed and the results were compared with those calculated by the CESEC-III and STRIKIN-II, which were used to prepared the final safety analysis report (FSAR) of Ul-Chin Units 3 and 4 (UCN-3/4). Through the study, it was concluded that the HSM of KNAP showed the acceptable results

  3. Development, assessment and application of TRAC-BF1/v2001.2 for beyond design basis BWR LOCA transients

    International Nuclear Information System (INIS)

    In preparation for the possible transition to risk based licensing, it is increasingly important to demonstrate the applicability of Best Estimate codes to more extreme conditions corresponding for example to limited equipment availability. With this idea in mind, we have reviewed the application of TRAC-BF1 to Large and Small Break LOCAs. In this context this work describes the assessment and applications of the Penn State University (PSU) version 2001.2 of TRAC-BF1 with all the PSI updates, to the analysis of hypothetical Large Break (LB) LOCAs in a BWR/4 with postulated limited ECC availability. Since in contrast to a LB-LOCA in a BWR with full ECC availability, the rod surface temperatures reach relatively high values, additional assessment of the code under such conditions is required. Hence, after analyzing bottom flooding separate-effect experiments in two different heater rod bundles and a TLTA LB-LOCA test, we shall present and discuss the results of the analysis of a LB-LOCA with highly restricted Emergency Core Cooling flow in a BWR/4. In this context, we shall also assess different descretization of some terms of the 3D momentum equations, which was found to be important in the analysis of BWR Small Break LOCAs

  4. Hungarian approach to LOCA analyses for SARs

    International Nuclear Information System (INIS)

    The Hungarian AGNES project in the period of 1992-94 was performed with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised - among others - a complete design basis accident (DBA) analysis. Major part of the thermal-hydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with conservative approach. In the medium size LOCA calculations and the PTS studies the six reactor cooling loops of the WWER-440/213 system were modelled by three loops (a single, a double and a triple loop). In the further developed version of the input model used in small break LOCA and other DBA analyses the six loops were modelled separately. The nodalisation schemes of the reactor vessel and the pressurizer, moreover the single primary loops are identical in the two input models. For the six-loop inputs model the trip cards, general tables and control variables are generated by using a RELAP5 object-oriented pre-processing interactive code, the TROPIC 4.0 code received from TRACTEBEL Belgium. The six-loop input model for WWER-440/V213 system was verified by the data of two operational transients measured in Paks NPP. The analysis of large break LOCAs, where the combined simultaneous upper plenum and downcomer injection results in a rather complicated process during reflooding phase, was carried out by using the ATHLET mod 1.1 Cycle code version (developed by GRS) in the framework of a bilateral German-Hungarian cooperation agreement using two-loop (1+5) input model. Later on in our safety analysis activities the application of best estimate methodology gained ground. In the last years AEKI in framework of different projects as US CAMP activity, EU PHARE and 5th Framework Programmes, as well as national projects to support the plant operation performed also many cases of LOCA analysis including primary to secondary leakages, feedwater and steam line breaks. These can be the preparation for a new DBA Analysis project

  5. Development and qualification of the LOCA analysis system CUPIDON-DEMETER

    International Nuclear Information System (INIS)

    As a support to the power reactor safety analysis, codes describing PWR-type fuel behaviour under LOCA conditions have been developed by CEA under the sponsorship of IPSN (Institute for Nuclear Safety and Safeguards). These codes are CUPIDON for unirradiated single fuel rod and DEMETER devoted to the prediction of the behaviour of a whole assembly. An extensive qualification program is underway. The main models: heat transfer, zircaloy oxidation and cladding deformation and rupture, have been separately benchmarked using PHEBUS experiments, German and Japanese transient tests and EDGAR transient tests. Qualification with global tests FLASH and PHEBUS is now underway. From the first calculations, it is concluded that CUPIDON gives a fair account of the behaviour of the rods tested in FLASH and PHEBUS. The new developments of the codes will mainly include: the influence of azimuthal differences of cladding temperature on zircaloy creep rate; the influence of the irradiation effect on mechanical behaviour if it proves significant; the introduction of a subchannel flow blockage index, using in particular PHEBUS phase II results. (author)

  6. Light Water Reactor LOCA Safety Parameter Analysis Using The Reactor Dynamics Code Dynode and Subchanflow

    International Nuclear Information System (INIS)

    Safety margin evaluation for LWR can be strongly improved by means of multi-physics and multi-scale methodologies. Thus, projects like NURISP have been focused on incorporating into one computational platform the latest advances in reactor simulation tools. An important task of KIT in the frame of NURISP was the improvement, extension and integration of the pin power reconstruction method (PPR) of DYN3D. The flexibility of the new development allows LOCA refinement in the spatial mesh for specific regions of interest or even having a whole core with pin-by-pin resolution. This integration is a step forward in the direction of two-level coupling with the subchannel code FLICA, being one of the major objectives of NURISP. In order to investigate the scope of the new DYN3D implementation, an off-line fast running methodology for the evaluation of LOCA safety parameters using DYN3D and the subchannel code SUBCHANFLOW has been developed. The DYN3D-PPR method has been used to obtain a time-dependent 3D pin power distribution in the hottest assembly of a mini-core during a transient. The bundle power maps obtained in this way are used as an input for SUBCHANFLOW to evaluate the time-dependent variation of LOCA safety parameters such as DNBR or the maximum cladding temperature. As a result, a 3D-map of LOCA safety parameters is available for each simulation time step. On the other hand, the stand-alone version of DYN3D contains an internally coupled 1D thermal-hydraulic model LOCA that provides the on-line thermal-hydraulic feedback for the updating of cross sections. Additionally it allows also evaluating LOCA safety parameters using the new methodology and DYN3D stand-alone calculation is presented. As expected, the use of SUBCHANFLOW yields a more detailed and accurate prediction of LOCA safety parameters comparing with the 1D thermal-hydraulics model LOCA

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

    International Nuclear Information System (INIS)

    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)

  8. Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/LSTF test 7

    Directory of Open Access Journals (Sweden)

    Mazgaj Piotr

    2016-01-01

    Full Text Available Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant design and operation. In the field of Loss Of Coolant Accident, evolutions of the regulations are discussed in various countries taking into account the very unlikely character of a double-ended guillotine break and questioning the necessity to study such an event with Design Basis Conditions assumptions. As a consequence, the consideration of intermediate size piping rupture becomes more and more important. The paper presents the modeling of the Test Facility ROSA-2/LSTF in the calculation code CATHARE 2.V2.5. OECD/NEA ROSA-2 Project Test 7 was conducted with the Large Scale Test Facility on June 14, 2012. The experiment simulated the thermal-hydraulic responses during a PWR 13% cold leg Intermediate Break Loss Of Coolant Accident (IBLOCA. The break was simulated by a cold leg upwardly mounted long break nozzle. The facility and the experiment conditions are modeled in CATHARE. The vessel is modeled by using a 3D module. A thermal-hydraulic analysis is conducted and the obtained results are subsequently compared with the experimental results from ROSA-2/LSTF Test 7. Evaluation of the differences between experimental and calculated results is discussed.

  9. Analysis of a large break loss of coolant accident at a pressurized water reactor

    International Nuclear Information System (INIS)

    The paper presents results of a demo analysis of a large-break loss-of-coolant accident (LB LOCA) at a PWR of the Zion NPP. The event, double ended cold leg break, is analyzed using the best estimate method, which has been verified based on the results of the L2-5 test experiment performed at the LOFT facility within the OECD 'Best Estimate Methods Uncertainty and Sensitivity Evaluation' programme. The key parameter of the event analyzed, i.e. the maximum cladding temperature, is represented by one-sided upper (0,95; 0,95) tolerance limit, in accordance with the license requirement. The uncertainty analysis is performed by using the non-parametric Wilks formula at the order of one to five, as well as by using parametric statistics assuming normal distribution of the output parameter, i.e. the maximum cladding temperature. The sensitivity of the maximum cladding temperature to the uncertain input parameters is evaluated by using Spearman's rank correlation coefficient and Pearson's correlation coefficient. The results of the sensitivity analysis determine the conservative RELAP5/MOD3.3 code models and the conservative initial and boundary conditions for a LB LOCA. The application of statistical methods is made possible owing to the availability of two hundred RELAP5/MOD3.3 code calculation runs with random values of 33 uncertain input parameters. (orig.)

  10. Analysis and development of the automated emergency algorithm to control primary to secondary LOCA for SUNPP safety upgrading

    International Nuclear Information System (INIS)

    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)

  11. Fuel behavior during a LOCA: LOFT experiments

    International Nuclear Information System (INIS)

    The LOFT experiments have provided the following fuel behavior information which appears to be valuable for improving the safety of PWR operation and resolving PWR licensing issues: (1) A generic unassisted core cooling event occurs during large-break LOCAs that dominates the cooling of the core before ECC reflood commences and potentially eliminates the possibility of flow channel blockage from prepressurized fuel rod swelling. (2) The large-break LOCA decompression forces do not disturb the normal control rod gravity drop and may not structually damage the fuel assemblies. (3) Large-break LOCA core cooling may also be enhanced by spacer grid and core counter flow delay of liquid escape from the core boundaries and liquid fallback from the upper plenum into the core region. (4) Lower fuel rod prepressurization may be possible in PWR fuel rods which would reduce flow channel blockage complications during LOCA's. (5) Uniform fuel rod cladding temperature indications during the large break LOCA's do not confirm expectations for the fuel rod cladding temperature variations that would inhibit development of flow channel blockages by ballooning of prepressurized fuel rods

  12. Analysis of axial fuel relocation based on gamma scan data from OECD Halden Reactor Project LOCA tests

    International Nuclear Information System (INIS)

    The on-going LOCA test program IFA-650 at the OECD Halden Reactor Project (HRP) conducts in-house gamma scanning as standard post-irradiation examination (PIE) procedure on the fuel rod. One of the primary objectives of the program is to investigate fuel relocation into the balloon region. A simple model called Gamma Transport Model was formulated for purpose of interpretation of fuel relocation based on the gamma scan data. Fuel relocation may have strong effect on the linear heat generation rate at the balloon due to, firstly, increase in linear fuel density, and secondly due to differences in burn-up and local heat generation rate at the periphery and bulk of the pellet. For this analysis, a pair of isotopes with very different FP yields for U and Pu isotopes is selected from the gamma scan spectrum. The intention is to use the difference in their ratio in the balloon region to qualitatively make conclusion on the fuel relocation. As a separate outcome, the same analysis can be applied to the ejected fuel region and draw conclusion on its origin (pellet rim or bulk). The Gamma Transport Model is validated against a special case from the Halden's LOCA test program and then applied for the analysis of selected tests. (author)

  13. Decay heat removal and operator's intervention during a very small LOCA

    International Nuclear Information System (INIS)

    Sample calculations were done for KORI-1 to develop a better understanding of what happens after very small LOCA (< or approx.0.05 ftsup(2)). For a water-side break with the break size larger than 0.006ftsup(2), fluid-loss through break exceeds the makeup. If the break sizeis larger than 0.008ftsup(2), decay heat can be completely removed through break. Based on these results, it was concluded that KORI-1 is fairly safe for the whole spectrum of sizes in very small LOCA. However, for the reactor with 9000MWe or 1200MWe, a certain spectrum of sizes in very small LOCA should be carefully considered. In the accident sequence the transition from natural circulation to pool boiling or from pool boiling to natural circulation may be troublesome to the operator or in the safety analysis. Operator's intervension was discussed; primary pump shutoff, HPI pump shutoff, break isolation, and opening relief valve. It was proved that continuous operation of HPI pumps after shutdown will not threaten the integrity of the primary system. (Author)

  14. Analysis of the bubble condenser structure of WWER-440 NPP under LOCA loading

    International Nuclear Information System (INIS)

    Two problems may arise in relation to the title topic: (1) problem with the uplift of the beams I 600 of the first floor, and (2) possible plastic collapse of the wall on the 12th floor. The problems were tacked by computer calculations. The FEM model of the bubble condenser was created in the ANSYS 6.0 environment and analyzed for the pressure loading defined for the LOCA accident in IAEA TECDOC 803. The model of the bubble condenser structure so created included all geometrical and material non-linearities. The duration of the pressure wave was 0.4 s, amplitude 30 kPa. The analyses revealed that a plastic collapse of the tank wall is not the most critical failure mode. Instead, weld connections appear to be the most critical parts of structure. The tank walls are very ductile and the results of the analyses are in agreement with the test simulating the LOCA accident. The tank walls suffered no damage during the tests

  15. Effect of Sump Clogging in Post-LOCA Long Term Cooling Process

    International Nuclear Information System (INIS)

    One of the important concerns on the safety analysis of nuclear power plant (NPP) is to confirm the adequacy of design provision and operational procedure related to the long term cooling (LTC) process following a loss-of- coolant accident (LOCA). The post-LOCA LTC capability of NPP is specified at the licensing requirements in most countries such as 10 CFR 50.46 in USA. Recently a potential to loss of recirculation due to containment recirculation sump (CRS) blockage have been issued in regulatory review on the LTC in Korea. In the course of LOCA, debris generated by the LOCA could accumulate and block the screen of the sump so that the flow to ECCS could be lost at the recirculation actuation signal (RAS) indicating the water of the refueling water storage tank (RWST) was emptied. To evaluate if this issue is significant in Korean NPP and to provide the guidance to further in-depth analysis, an analysis has been performed for the LTC process for a selected plant. A LTC behavior in RCS and core following a double ended hot leg guillotine (DEHLG) break was calculated with RELAP5/MOD3.3, a best estimate system thermal-hydraulic analysis code. The DEHLG break is selected that could generate the largest amount of debris. The significance of the CRS blockage was evaluated through blocking the ECCS source after RAS

  16. Evaluation of VVER-1200/V491 reactor pressure vessel integrity during large break LOCA along with SBO using MELCORE 1.8.6

    International Nuclear Information System (INIS)

    Integrity evaluation for the lower head of RPV during severe accident progress is important to Severe Accident Management Guidelines (SAMG). In this study, MELCOR 1.8.6 is used to evaluate the lower head integrity of RPV for VVER-1200 (V-491) reactor during simultaneous occurrence of LB LOCAs and SBO. The study figures out several parameters related to melt down progress such as: rupture position and rupture timing. The hydrogen generation and amount of corium inside the vessel are also investigated. The availability of the second stage hydro-accumulators (HA2) in the VVER-1200 (V-491) as additional design is assumed to evaluate the cooling capacity as well as to maintain the vessel integrity for long-term. (author)

  17. Numerical and experimental analysis of transient wave propagation through perforated plates for application to the simulation of LOCA in PWR

    International Nuclear Information System (INIS)

    Highlights: ► LOCA simulation: analyze effects of perforated plates on rarefaction waves propagation at reactor level. ► Localized impedance relations to account for singular head loss and acoustic delays. ► Experimental campaign to produce a reference solution for wave propagation through a thickorifice plate. ► Qualification of a reference CFD code to produce reflection/transmission solutions for any geometry. ► Calibration procedure for impedance relations using CFD reference solutions. - Abstract: Loss of coolant accident is characterized by a transient rarefaction wave propagating inside the primary loop after pipe break, resulting in fluid loading on internal structures, especially on the baffle surrounding the reactor core. In that case, loading comes from differences in rarefaction wave travel times, whether it propagates through the reactor core or through the by-pass between baffle and core barrel, yielding delays and pressure differences on both sides of baffle plates. Propagation is strongly influenced by geometric obstacles such as holes in baffle reinforcement plates, which cannot be represented in a numerical model of the whole primary loop and has then to be replaced by suitable impedance relations. A methodology to validate and calibrate such impedance models, based on specific experimental results and high-order CFD simulations, is thus proposed. Models are tested and integrated into EUROPLEXUS fast-transient fluid–structure dynamics software.

  18. Comparison report on OECD-CSNI LOCA standard problem N.7: analysis of a reflooding experiment

    International Nuclear Information System (INIS)

    The 7. CSNI Standard Problem on LOCA (Loss of coolant accident) is a separate effects experiment proposed by the French CEA. For the first time the subject was the reflooding phase of the accident. It is based on an experiment performed on the ERSEC loop located at the Grenoble Nuclear Center. This blind exercise was a test to assess the capability of various codes (RELAP 4 mod 6 and others codes) to predict the thermal behaviour during reflooding in a simple test section with given boundary and initial conditions, and also a way to check how the physics of thermal-hydraulic phenomena involved during reflooding are handled in these codes. Results from individual results from several countries such as Australia, Finland, France, Germany, Nederland, Japan, Sweden, Switzerland and the United States, are presented and compared

  19. A simple approach for pre-LOCA analysis of MTR type research reactor

    International Nuclear Information System (INIS)

    In this study, it is intended to analyse early phases of a protected loss of coolant accident (LOCA) for TR-2 research reactor at Istanbul, and to show applicability of the present model to the other similar types of research reactors. Even though, there has been substantial amount of experimental and numerical works concerning LOCA of research reactor in the literature, most of the works has been done for the latest phase of accident where the core was totally uncovered and being cooled by natural circulation of air. It is our aim to investigate the transient situation since the time when coolant is beginning to be lost throughout one or more of the main coolant pipes which where supposed to be broken guillotine-like to the time when the core is totally uncovered. The modelling of the problem was separated into two phases: in the first phase when the water level of the pool being decreased in a pre-estimated time-dependent way calculated by using modified Bernoulli equation, the conservation equations are solved by a usual implicit finite difference algorithm. The later phase, when water level reaches to the top level of fuel plates and begins to decrease until the bottom of the core, needs some modifications to the approach used for the first phase. Because, the coolants channels among fuel plates are filled with air when the level goes below, and the fuel plates are being cooled by air above the water level. This complexity is resolved using a moving boundary approach in the numerical solution. A Lagrange type interpolation approximation for the derivatives along with interface conditions is the neighborhood of the air-water interface was imported to the numerical algorithm. For the meshes which are not close to the interface above mentioned usual finite difference scheme to solve conservation equations both for air and water side. The analyse is performed for a nominal channel and for a hot channel

  20. ROSA-III experimental program for BWR LOCA/ECCS integral simulation tests

    International Nuclear Information System (INIS)

    This is the final report of the ROSA-III experimental program, in which the summary of integral simulation test results is described on thermal-hydraulic behavior during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) and on the effectiveness of the emergency core cooling system (ECCS). Also presented in the report are the assessment results of computer codes for the BWR LOCA analysis and of the similarity between ROSA-III test results and thermal-hydraulic phenomena during a BWR LOCA by using ROSA-III test data and code analysis results. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core consisting of four half-length bundles. Many test series were conducted between April 1978 and March 1983. The similarity between a ROSA-III test and a BWR LOCA concerning the fundamental thermal-hydraulic phenomena has been confirmed for major ROSA-III tests. The accident scenario has been well understood and defined for various break locations and break sizes. The effectiveness of the current BWR ECCS design has been well demonstrated. (author)

  1. A preliminary sensitivity study on LOCA fuel/coolant heat transfer analysis using the Theta1-B code

    International Nuclear Information System (INIS)

    A series of computer experiments has been performed, using the reactor thermal analysis code Theta1-B, to ascertain the significance and relative importance of the various heat transfer regimes in relation to the prediction of maximum fuel cladding temperature for the slowdown phase of a LOCA in a PWR system. Also considered were the significance of the choice of heat transfer correlation for a particular regime, the method of delineating between regimes, and core inlet coolant flow conditions. The most significant finding of this study was that, for a LOCA resulting from a double-ended rupture of a PWR inlet feeder, a nominal change in the coolant mass flow rate for the period 3 to 11 s, from essentially zero flow to a nominal 1.5% of the normal core flow, produced a marked decrease in maximum fuel cladding temperature (around 600 F). For this period the coolant was predicted to be superheated vapour at the position of maximum heat flux and the substantial change in performance is associated with the facts that (a) no correlation is included in the code to estimate heat transfer coefficients for single-phase stagnant coolant conditions, an arbitrary minimum value of the heat transfer coefficient being adopted; and (b) for low flow rates of single-phase coolant the code assumes that the Dittus Boelter heat transfer correlation is applicable. These place a significant limitation on use of the code for reliable prediction of maximum cladding temperature under such conditions, and highlights an area where further information is required

  2. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    International Nuclear Information System (INIS)

    It was postulated, in the cooling system of the core, a LOCA, where 431 m3 of soda almost instantaneously was lost. This inventory contained 1.87x1010 Bq/m3 of tritium, 2.22x107 Bq/m3 of cobalt,3.48x108 Bq/m3 of cesium and 3.44x1010 Bq/m3 of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x106 Bq/m3, 1,11x104 Bq/m3 and 1,85x103 Bq/m3) after 22 hours, respectively for 3H, 60Co, 131I and 137Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  3. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (LIMA/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m{sup 3} of soda almost instantaneously was lost. This inventory contained 1.87x10{sup 10} Bq/m{sup 3} of tritium, 2.22x10{sup 7} Bq/m{sup 3} of cobalt,3.48x10{sup 8} Bq/m{sup 3} of cesium and 3.44x10{sup 10} Bq/m{sup 3} of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10{sup 6} Bq/m{sup 3}, 1,11x10{sup 4} Bq/m{sup 3} and 1,85x10{sup 3} Bq/m{sup 3}) after 22 hours, respectively for {sup 3}H, {sup 60}Co, {sup 131}I and {sup 137}Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  4. Characteristic responses of core exit thermocouples during inadequate core cooling in small break LOCA experiments conducted at Large-Scale Test Facility (LSTF) of ROSA-IV program

    International Nuclear Information System (INIS)

    Characteristic responses of core exit thermocouples (CETs) for detection of an inadequate core cooling (ICC) were experimentally studied at a large-scale plant simulator for a pressurized water reactor (PWR). The ICC conditions were established by assuming a failure or delayed actuation of high pressure injection (HPI) system. The CET responses were studied in twenty-one experiments simulating different kinds of small break loss-of-coolant accident (SBLOCA) in the PWR. It is concluded that the CETs are useful for ICC monitoring during boil-off process. An empirical equation to estimate a delay time for ICC detection is obtained for the experiments with scaled break area less than 5%. On the other hand, the ICC was not detected in 10% cold leg break test due to water falling back from the hot legs

  5. Safety evaluation of small-break LOCA with various locations and sizes for SMART adopting fully passive safety system using MARS code

    International Nuclear Information System (INIS)

    Highlights: • A safety evaluation of SMART fully passive safety system was performed for a SBLOCA. • CMT was modeled with thermal-front model for the temperature gradient in a volume. • Limiting case was determined with sensitivity of various break locations and sizes. • The collapsed liquid level is maintained high enough above the top of the core. • It was proven that the core is not uncovered for 72 h after the SBLOCA with PSIS. - Abstract: A safety evaluation of SMART adopting a fully passive safety system was performed for a small-break loss of coolant accident (SBLOCA) with various break locations and sizes using the MARS code. It is necessary for SMART, adopting a fully passive safety system composed of passive safety injection system (PSIS), automatic depressurization system (ADS), and passive residual heat removal system (PRHRS), to satisfy the passive safety performance requirements, i.e., the capability to maintain safe shutdown conditions for a minimum of 72 h without AC power supply or operator action in the case of a design basis accident. A number of SBLOCAs of different locations and sizes were analyzed using the MARS code. The results of the break spectrum analyses showed that the collapsed liquid level inside the core support barrel is maintained high enough above the top of the core owing to the sufficient passive safety injection flow from the core makeup tank (CMT) and safety injection tank (SIT). Therefore, the core is not uncovered for 72 h after the break without AC power supply or operator action, resulting in a continuous decrease of fuel cladding temperature throughout the transient

  6. APT Blanket Thermal Analysis of Cavity Flood Cooling with a Beam Window Break; FINAL

    International Nuclear Information System (INIS)

    The cavity flood system is designed to be the primary safeguard for the integrity of the blanket modules and target assemblies during loss of coolant accidents, LOCA''s. In the unlikely event that the internal flow passages in a blanket module or a target assembly dryout, decay heat in the metal structures will be dissipated to the cavity flood system through the module or assembly walls. This study supplements the two previous studies by demonstrating that the cavity flood system can adequately cool the blanket modules when the cavity vessel beam window breaks

  7. A Study of Time Response for Safety-Related Operator Actions in Non-LOCA Safety Analysis

    International Nuclear Information System (INIS)

    The classification of initiating events for safety analysis report (SAR) chapter 15 is categorized into moderate frequency events (MF), infrequent events (IF), and limiting faults (LF) depending on the frequency of its occurrence. For the non-LOCA safety analysis with the purpose to get construction or operation license, however, it is assumed that the operator response action to mitigate the events starts at 30 minutes after the initiation of the transient regardless of the event categorization. Such an assumption of corresponding operator response time may have over conservatism with the MF and IF events and results in a decrease in the safety margin compared to its acceptance criteria. In this paper, the plant conditions (PC) are categorized with the definitions in SAR 15 and ANS 51.1. Then, the consequence of response for safety-related operator action time is determined based on the PC in ANSI 58.8. The operator response time for safety analysis regarding PC are reviewed and suggested. The clarifying alarm response procedure would be required for the guideline to reduce the operator response time when the alarms indicate the occurrence of the transient

  8. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

    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

  9. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

    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 design (Oconee) and a Westinghouse 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 availability, and main coolant pump trip on these times. 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 s for the Babcock and Wilcox and Westinghouse 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

  10. Effects of high temperature ECC injection on small and large break BWR LOCA simulation tests in ROSA-III program (RUNs 940 and 941)

    International Nuclear Information System (INIS)

    The ROSA-III program, of which principal results are summarized in a report of JAERI 1307, conducted small and large-break loss-of-coolant experiments (RUNs 940 and 941) with high water temperature of the emergency core cooling system (ECCS) are one of the parametric study with respect to the ECCS effect on core cooling. This report presents all the experiment results of these two tests and describes additional finding with respect to the hot ECC effects on core cooling phenomena. By comparing these two tests (water temperature of 393 K) with the standard ECC tests of RUNs 922 and 926 (water temperature of 313 K), it was found that the ECC subcooling variation had a small influence on the core cooling phenomena in 5 % small break tests but had larger influence on them in 200 % break tests. The ECC subcooling effects described in the previous report are reviewed and the temperature distribution in the pressure vessel is investigated for these four tests. (author)

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

    International Nuclear Information System (INIS)

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

  12. An application of CONTAIN code to hydrogen distribution analysis following a LOCA in the CANDU-6 containment

    International Nuclear Information System (INIS)

    As a feasibility study on the use of the CONTAIN code for the CANDU PHWR plant, the hydrogen concentration distribution in the short term following LOCA without ECC has been analyzed. lie purpose of this study is to verify the hydrogen mixing analysis in a CANDU-6 plant and the adequacy of the location of hydrogen ignitors. lie coolant mass and energy discharge rates obtained from the utility's CATHENA calculation results were modified for the CONTAIN analysis with the assumption of flashing at the containment pressure calculated by PRESCON2. Tle heat removal by the local air coolers were based on their rated cooling capacity. The time-varying flow rates were described to simulate the dousing characteristic. The results show that the calculated peak hydrogen concentrations are about equal to those based on the PRESCON2 model. No hydrogen bum occurs due to low hydrogen concentration. However in case of much enhanced beat removal by the local air coolers, high hydrogen concentrations induce hydrogen burns without significant pressure rise. It is shown that the hydrogen ignitors can play a major role in removing hydrogen. It is necessary to examine further the reason of over-estimation of the accident compartment temperature and the effect of discharge condition and hew sinks on the hydrogen concentrations Complete modeling of the CANDU-PHWR by the CONTAIN code requires that CANDU specific material and safety feature models should be available. (author)

  13. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    International Nuclear Information System (INIS)

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  14. Small break loss of coolant accident analysis for the international reactor innovative and secure (IRIS)

    International Nuclear Information System (INIS)

    The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. IRIS has been primarily focused on establishing a design with innovative safety characteristics. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. In IRIS, this concept is implemented through the safety by design approach, briefly discussed in this paper. Because of the safety by design approach, the number and complexity of the safety systems and required operator actions can be minimized in IRIS. The net result is a design with significantly reduced complexity and improved operability, and extensive plant simplifications to enhance construction. This paper focus is on the IRIS response to a small break loss of coolant accident (LOCA). Note that in IRIS large break LOCA events are eliminated by the use of an integral layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS integral configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than providing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. A brief overview of the IRIS safety concept, describing the simplified engineered

  15. A study on the aging degradation of ethylene-propylene-diene monomer (EPDM) under LOCA condition

    International Nuclear Information System (INIS)

    The aging degradation and lifetime assessment of a domestic class 1E Ethylene-Propylene-Diene-Monomer (EPDM), which is a popular insulating elastomer for electrical cables in the nuclear power plants, were studied for equipment qualification verification under the Loss of Coolant Accident (LOCA) conditions. The specimens were acceleratively aged, underwent a LOCA environment, as well as tested mechanically, thermo-gravimetrically, and spectroscopically according to the American Society of the Testing of Materials (ASTM) procedures. The tensile test results revealed that the elongation at break gradually decreased with an increasing aging temperature. The lifetime of EPDM aged isothermally at 140 .deg. C was 1,316 hours and reduced to 1,120 hours after experiencing the severe accident test. The activation energies of the elongation reduction were 1.10 ± 0.196 eV and 0.93 ± 0.191 eV before and after the LOCA condition, respectively. The TGA test results also showed that the activation energy of the aging decomposition decreased from 1.35 eV to 1.02 eV after undergoing the LOCA environment. Although the mechanical property changes were discernibly observed during the aging process, along with the LOCA simulation, the FT-IR analysis showed that the spectroscopic peaks and their intensities did not alter significantly. Therefore, it can be concluded that the degradation of the domestic class 1E EPDM due to aging can be tolerable, even in severe accident conditions such as LOCA, and thus it qualifies as a suitable insulating material for electrical cables in the nuclear power plants

  16. A study on the aging degradation of ethylene-propylene-diene monomer (EPDM) under LOCA condition

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Yong Dae; Lee, Hyun Seon; Kim, Yong Soo [Hanyang University, Seoul (Korea, Republic of); Song, Chi Sung [Korea Institute of Machinery and Materials, Daejeon (Korea, Republic of)

    2011-06-15

    The aging degradation and lifetime assessment of a domestic class 1E Ethylene-Propylene-Diene-Monomer (EPDM), which is a popular insulating elastomer for electrical cables in the nuclear power plants, were studied for equipment qualification verification under the Loss of Coolant Accident (LOCA) conditions. The specimens were acceleratively aged, underwent a LOCA environment, as well as tested mechanically, thermo-gravimetrically, and spectroscopically according to the American Society of the Testing of Materials (ASTM) procedures. The tensile test results revealed that the elongation at break gradually decreased with an increasing aging temperature. The lifetime of EPDM aged isothermally at 140 .deg. C was 1,316 hours and reduced to 1,120 hours after experiencing the severe accident test. The activation energies of the elongation reduction were 1.10 {+-} 0.196 eV and 0.93 {+-} 0.191 eV before and after the LOCA condition, respectively. The TGA test results also showed that the activation energy of the aging decomposition decreased from 1.35 eV to 1.02 eV after undergoing the LOCA environment. Although the mechanical property changes were discernibly observed during the aging process, along with the LOCA simulation, the FT-IR analysis showed that the spectroscopic peaks and their intensities did not alter significantly. Therefore, it can be concluded that the degradation of the domestic class 1E EPDM due to aging can be tolerable, even in severe accident conditions such as LOCA, and thus it qualifies as a suitable insulating material for electrical cables in the nuclear power plants

  17. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

    OpenAIRE

    Takeshi Takeda

    2014-01-01

    RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM) measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW) injection into the secondary-side of both steam generators (SGs) as the AM measures were taken 10 min after a safety injection signal. The primary depre...

  18. Modeling of the PACTEL facility and simulation of a small break LOCA experiment with the TRACE V5.0 code

    International Nuclear Information System (INIS)

    The applicability of the TRACE code to analyze VVER type PWRs is being studied in Finland. To this end the PACTEL integral test facility of Lappeenranta University of Technology (LUT), has been modeled using the TRACE V5.0 patch2 code. PACTEL is a full height model of VVER-440 type nuclear reactor located in Loviisa. A small break loss-of-coolant-accident (SBLOCA) experiment SBL-30 with multiple loops included in the test was chosen to be simulated using the model. The calculation results succeeded reasonably well in estimating the experiment propagation. (author)

  19. Design basis neutronics calculations for NRU-LOCA experiments

    International Nuclear Information System (INIS)

    The report describes the neutronics analysis for the LOCA simulation experiments in the NRU reactor. The experimental program will provide greater understanding of nuclear fuel assembly behavior during the heatup, reflood and quench sequence of a hypothetical LOCA. The decay heat and stored heat, which are the energy source in a LOCA will be simulated by fission heat provided by the NRU reactor. The reactor, the test and test operation are described

  20. Analysis of transient dry patch behavior on CANDU reactor calandria tubes in a LOCA with late stagnation and impaired ECI

    International Nuclear Information System (INIS)

    An analytical method to describe the behavior of transient dry patches on CANDU reactor calandria tubes has been developed. Dry patches may form following the sagging of a pressure tube onto a calandria tube in certain low-probability scenarios in which a loss-of-coolant accident occurs with subsequent failure or impairment of the emergency cooling injection function. Results of the analysis show that the dry patches will not grow beyond a few degrees on each side of the bottom of the calandria tube and will rewet within a few tens of seconds, with the values depending on the specific CANDU reactor design and the mechanism of dry patch formation and rewetting. Maximum local calandria tube temperatures reached during the transient will be about 5500C to 7000C. There will be no significant effects (0C) on fuel, sheath and maximum pressure tube temperatures. The analytical results provide confidence that pressure tube and calandria tube integrity will not be threatened by dry-patch formation in the LOCA scenarios studied

  1. An Analysis of Break,Break,Break Based on the Stylistic Theory

    Institute of Scientific and Technical Information of China (English)

    李瑶

    2014-01-01

    Break,Break,Break is a poem by Alfred Lord Tennyson, the Poet Laureate during the Queen Victoria's reign. This exquisite little poem is wel known for the poet’s grief-stricken feelings and heart-broken emotions over the premature death of his best friend, Arthur Henry Hal am. Most of the previous studies on this poem focus on the emotional level to consider it as an elegy, expressing sorrow and lamentation for the death of a particular person. However, in order to have a deep understanding in general, this paper analyzes the poem based on the stylistic theory, concerning on the phonological level and the grammatical level. It aims at helping the readers to cultivate a sense of appropriateness, to sharpen the understanding and appreciation of literary works and to achieve adaptation in translation.

  2. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2014-01-01

    Full Text Available RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW injection into the secondary-side of both steam generators (SGs as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses.

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

    International Nuclear Information System (INIS)

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

  4. RELAP5 analyses of OECD/NEA ROSA-2 project experiments on intermediate-break LOCAs at hot leg or cold leg

    International Nuclear Information System (INIS)

    Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution. (author)

  5. RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

    Science.gov (United States)

    Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

    Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

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

    International Nuclear Information System (INIS)

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

  7. Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

    Directory of Open Access Journals (Sweden)

    F. Reventós

    2012-01-01

    Full Text Available Integral test facilities (ITFs are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

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

    International Nuclear Information System (INIS)

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

  10. Semiscale liquid hold-up investigations: a comparison of results from small break LOCA tests performed in the Semiscale MOD-2A and MOD-2C facilities

    International Nuclear Information System (INIS)

    Results are compared from small break loss-of-coolant accident (SBLOCA) experiments performed in two different versions of the Semiscale facility. These experiments were designed to investigate the effect of downcomer to upperhead core bypass flow on transient severity. The first set of experiments, S-UT-6 and S-UT-8 with 4% and 1.1% bypass flows respectively, were performed in the Mod-2A facility. The second set of experiments, S-LH-1 and S-LH-2 with 0.9% and 3% bypass flows, were performed in the Mod-2C facility. The effect of the net head of fluid in the steam generator primary tubes (liquid hold-up) on the transient severity is examined as well as the general mechanism of core level depression. Both Semiscale Mods are volume-scaled representatives of a four-loop pressurized water reactor (PWR), which simulates most of the major features of a PWR. All experiments were performed at high temperature and pressure (595 K hot leg fluid temperature; 15.6 MPa pressure)

  11. Recalculation of simulated post-scram core power decay curve for use in ROSA-IV/LSTF experiments on PWR small-break LOCAs and transients

    International Nuclear Information System (INIS)

    Simulated post-scram core power decay curve for use in Large Scale Test Facility (LSTF) tests has been calculated on a best-estimate basis, particularly in two points, i.e. estimation of the delayed neutron fission power and consideration of the stored heat in a pressurized water reactor (PWR) fuel rod. The New Power Curve provides a LSTF heater rod with the heat transfer rate from a PWR fuel rod that was estimated for a typical pressure transient during a PWR small-break loss of coolant accident. This approach neglects conservatively the effect of stored heat release from the LSTF heater rod considering that there is large uncertainty in the thermal conductivity of outer insulator in the LSTF heater rod. When the New Power Curve is used as the LSTF core power curve, the heat transfer rate from a LSTF heater rod gives a little conservative values as compared with the heat transfer rate from a PWR fuel rod. (author)

  12. ROSA-IV LSTF 5% cold leg break analysis using the RELAP5/Mod2 code

    International Nuclear Information System (INIS)

    This paper presents the results of the OECD International Standard Problem calculation no. 26 (ISP-26), performed with the RELAP5/Mod2 code (frozen version 36.06 for the CRAY-X/MP) at the Paul Scherrer Institute (PSI) and analyses their comparison with the 5% cold leg break test (run SB-CL-18) conducted on the Large Scale Test Facility (LSTF) of the ROSA-IV program. The principal objective of the present calculation is the analysis of the simulation capability of the code in regards to the main phenomena occurring during a typical Small Break Loss of Coolant Accident (SB-LOCA), such as: natural circulation heat removal; liquid holdup in the up- and downflow sides of the steam generator U-tubes in combination with a reflux condenser mode heat removal, core upper plenum pressure build-up causing manometric liquid level unbalance between core and downcomer; loop seal clearing concurrent with core liquid depression leading to core heat-up; core level recovery after loop seal clearing; vessel inventory boil-off leading to second core uncovering; core reflooding due to accumulator injection intervention. 7 refs., 15 figs., 2 tabs

  13. Couplet finite element fluid-structure interaction dynamic analysis for reactor internals under seismic and LOCA conditions

    International Nuclear Information System (INIS)

    The influence of Fluid-Structure interaction phenomena on seismically or LOCA excited reactor internals is becoming increasingly important for an accurate behavior and nuclear safety considerations. The internals by virtue of their location are exposed to mass of surrounding fluid and excited in motion by severe transients such as LOCA or an earthquake. The response of reactor internals, including fuel assemblies, core mass, upper and lower grid assemblies, core support frame, plenum-assembly, control-rod drive system and thermal-shield, is related to the operational requirements of each subcomponent in assessing their integrity. These subcomponents, submerged in an excited fluid-mass, constitute intrinsically an overall vibrating system with individual and combined eigen frequencies. Each individual subcomponent is subjected to fluid-elastic interaction, which affects their vibrating characteristics. Numerical results have been obtained for PWR internals and fuel-assemblies in water of a typical 950 MWe plant. In order to completely assess the adequacy of each subcomponent, the overall effects of both seismic and LOCA conditions are evaluated and superimposed directly. Various effects of local non-linearities due to gaps, coulomb friction, impacting and sliding are taken into account. An overall dynamic response due to seismic and LOCA effects, has shown that maximum displacements and stress-resultants are well below the specified critical limits. (orig./GL)

  14. A feasibility study on the extended cycle from the point of view of Non-LOCA safety analysis

    International Nuclear Information System (INIS)

    Extended cycle operation has many advantages as compared with standard cycle operation (12 months) from the viewpoint of operators' exposure to radiation dose, man-power for maintenance and the production of nuclear waste. And it is more economic than the standard cycle operation. If the extended cycle operation is adopted in the CE type nuclear plant, the effective management and enhancement of operation capability could be expected. A feasibility study on the extended cycle operation should be performed. The purpose of this technical report is to perform safety analysis using the design data for Yonggwang Nuclear Plant (YGN) 3 and 4 extended cycle operation and to evaluate qualitatively and quantitatively the safety related design basis events to verify the satisfaction of the safety criteria. Boron dilution and steam line break accidents were found to be most influenced by the change of physics data due to the adaptation of the extended cycle operation. For boron dilution accident, source range monitor ratio of 3.07 for YGN 3 cycle 2 was decreased to 2.92. The 3D reactivity feedback effect due to the local heatup in the vicinity of stuck CEA was credited in the analysis of steam line break. No return-to-power occurred for the steam line break with offsite power available and return-to-power occurred for the steam line break with loss of offsite power. For the steam line break with loss of offsite power, the safety margin was preserved with respect to fuel performance (DNB and LHGR) despite the return-to-power. 6 tabs., 10 figs., 5 refs. (Author) .new

  15. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

    International Nuclear Information System (INIS)

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal (HR) system. These simulations were performed for the Preliminary Safety Analysis Report

  16. Post test analysis of ROSA-III double-ended break test RUN 901

    International Nuclear Information System (INIS)

    The ROSA-III test facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core. The purpose of ROSA-III experiment is to study the thermal-hydraulic behavior and the performance of the emergency core cooling system (ECCS) during a postulated loss-of-coolant accident (LOCA) and to provide the data base for the assessment and improvement of reactor safety analysis codes. RUN 901 was a first ROSA-III experiment with the fourth fuel assembly and assumed a 200 % double-ended break at the recirculation pump section line with full ECCS actuation. Post test analyses of RUN 901 were performed with the computer codes RELAP4J and RELAP5/MOD1/001. The system pressure response calculated with the two codes agreed well with the data. The agreement was also good for the core inlet flow behavior until the beginning of the lower plenum flashing (LPF). RELAP4J is a fast running code and calculated well the overall behavior of the mixture level in the core. However, the spray water was accumulated in the upper plenum (UP) due to the inability of the code to calculate counter current flow at the upper tie plate (UTP). While spray water was not accumulated in the upper plenum in the experiment. RELAP5 with an advanced two-phase flow model calculated well the rewetting of the fuel after the LPF and the top-down quench after the uncovery of the whole core. However, the incorporation of a CCFL model and/or the improvement in the interphase drag correlations are necessary to be able to calculate the mixture level behavior more accurately. An appropriate discharge coefficient is also necessary to calculate the break flow accurately with the RELAP5 characteristic analysis break flow model. (author)

  17. Review of the analysis of a failure to shutdown following a large LOCA in a Pickering NGS A reactor unit

    International Nuclear Information System (INIS)

    A review has been undertaken of the study performed by Ontario Hydro, with the assistance of Argonne National Laboratory, on the nature and consequences of a failure to shut down (LOSD) following a large LOCA in a Pickering A NGS reactor unit. Ontario Hydro analyzed the complete accident sequence, from the initiating LOCA and LOSD to containment behaviour and off-site radiological consequences. Argonne, working independently, analyzed only fuel behaviour, fuel channel response and failure, and moderator response. Both studies were in very close agreement. Off-site radiological consequences of the accident were not found to be worse than those of any severe dual-failure accident analyzed during the licensing process and thus to be within the regulatory limits from such accidents. Recommendations for further work are given

  18. Condensation in the cold leg as results of ECC water injection during A LOCA: modeling and validation

    International Nuclear Information System (INIS)

    During postulated LOCA events in pressurized water reactors, cold water is injected into cold legs by emergency core cooling system (ECCS). As the ECC water comes into contact with steam, the amount of condensation in the cold legs which results from mixing of the two phases is expected to have an effect on the thermal hydraulic behavior of the system. During boil off period and recovery period of a small break LOCA, the condensation in the cold leg is enhanced by the impingement of the ECC jet on the layer of liquid, when the flow in the cold leg is expected to be horizontal stratified. Consequently, the reactor coolant system (RCS) depressurization is accelerated, which in turn increases ECC flow rate and promotes accumulator injection. For a large break LOCA, the condensation process in the cold leg during refill period helps to reduce bypass flow at the top of downcomer, promoting ECC penetration. The condensation in the cold leg during reflood period is an important factor in determining the ECC bypass, the break flow rate, the downcomer and core water inventory, and the liquid subcooling in the downcomer, which in turn impacts the peak cladding temperature during reflood. A cold leg condensation model was considered for the new release of WCOBRA/TRAC-TF2 safety analysis code and presented in an authors' previous work. The model was further improved to better capture relevant data and a revised model was found to be in better agreement with such experimental data. The intent of this paper is to present the validation for the cold leg condensation model. The improved cold leg condensation model is assessed against various small break and large break LOCA separate effects tests such as COSI experiments, ROSA experiments and UPTF experiments. Those experiments cover a wide range of cold leg dimensions, system pressures, mass flow rates, and fluid properties. All the predicted condensation results match reasonably well with the experimental data. (author)

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

    International Nuclear Information System (INIS)

    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)

  20. Numerical analysis of the fusion of nuclear combustible rods under LOCA - type accidents

    International Nuclear Information System (INIS)

    The study of the melting of combustible rods is of great importance for the safety analysis of nuclear reactors. Due to the special characteristics of the problem, a sharp interface between the solid and liquid region does not exist, but appears a 'mushy' region in which the material is partially melted. The Finite Element Method is employed here, together with a regularized enthalpy formulation. Finally, the results obtained are presented and discussed. (Author)

  1. Simulation of LOCA within a German BWR containment with the coupled version of ATHLET-COCOSYS

    International Nuclear Information System (INIS)

    In this paper the effectiveness of the safety system in a generic Boiling Water Reactor (BWR) is investigated by simulating a loss-of-coolant accident (LOCA). To this end, a two-sided large break in a feed water line within the containment is simulated. Since the LOCA is characterized by significant interactions between the reactor cooling circuit and the containment, the simulation is performed using the coupled version of the thermo-hydraulic code ATHLET and the containment code COCOSYS. The analysis focuses on the calculated plant behavior from the accident initiation to the depressurization phase. The results show that the emergency core cooling system with only two operational pumps succeeds in cooling the fuel elements throughout the entire accident sequence. A comparison of the simulation results with a separate ATHLET simulation indicates slight deviations when the pressure in the reactor cooling circuit approaches the drywell pressure. (author)

  2. Validation of ORIGEN-S decay heat predictions for LOCA analysis

    International Nuclear Information System (INIS)

    Recent developments in the nuclear data libraries used by the ORIGEN-S isotope generation and depletion code have enabled the extension of the code to accurately predict the delayed energy release rates from nuclear decay (decay heat) at very short cooling times of interest to reactor accident analysis. Historically this time domain has required integral methods, such as the ANS-5.1 decay heat standard, because isotopic summation codes could not be reliably applied due to incomplete nuclear data. This paper describes work to validate ORIGEN-S against experimental measurements for decay times that extend down to about 1 second after fission. Benchmarks using measured gamma ray spectra following fission are also included because these results are important to predicting spatial energy deposition from delayed gamma energy release. (authors)

  3. One-dimensional system analysis code for reflood phase during LOCA

    International Nuclear Information System (INIS)

    A system code named REFLA-1D was developed by coupling the core thermo-hydrodynamic code and the primary system model for the analysis of the reflood phenomena. In order to assess the calculation method of this system code, the results of a base case test and parametric tests, which were run for the conditions of the base case test by varying only one parameter at a time, were compared with the results calculated with the system code. The calculation of the base case test showed a good agreement with the data for the core collapsed liquid level, the quench front elevation, and the heat transfer coefficient near the quench front. The calculation of the parametric test showed a good agreement with the data for the effect of the initial clad temperature and of the peak power, however, a good agreement was not obtained for the effect of system pressure. Further study of the two-phase flow modeling in the core and the quench front correlation against the pressure dependence is necessary for a better prediction of the system behaviors. (author)

  4. General Analysis of U-Spin Breaking in B Decays

    OpenAIRE

    Jung, Martin; Mannel, Thomas

    2009-01-01

    We analyse the breaking of U-spin on a group theoretical basis. Due to the simple behaviour of the weak effective hamiltonian under U-spin and the unique structure of the breaking terms such a group theoretical analysis leads to a manageable number of parameters. Several applications are discussed, including the decays B -> J/psi K and B -> D K.

  5. General analysis of U-spin breaking in B decays

    International Nuclear Information System (INIS)

    We analyze the breaking of U-spin on a group theoretical basis. Because of the simple behavior of the weak effective Hamiltonian under U-spin and the unique structure of the breaking terms such a group theoretical analysis leads to a manageable number of parameters. Several applications are discussed, including the decays B→J/ψK and B→DK.

  6. Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code; Analisis de un accidente LOCA en contencion de un reactor PWR-W con el codigo GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Perianez Alvarez, V.

    2013-07-01

    The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

  7. Analysis of chiral symmetry breaking mechanism

    International Nuclear Information System (INIS)

    The renormalization group invariant quark condensate μ is determinate both from the consistent equation for quark condensate in the chiral limit and from the Schwinger-Dyson (SD) equation improved by the intermediate range QCD force singular like δ (q) which is associated with the gluon condensate. The solutions of μ in these two equations are consistent. We also obtain the critical strong coupling constant αc above which chiral symmetry breaks in two approaches. The nonperturbative kernel of the SD equation makes αc smaller and μ bigger. An intuitive picture of the condensation above αc is discussed. In addition, with the help of the Slavnov-Taylor-Ward (STW) identity we derive the equations for the nonperturbative quark propagator from SD equation in the presence of the intermediate-range force is also responsible for dynamical chiral symmetry breaking. (author)

  8. Simulation of LOCA and ageing effect with containment liner mockup for analysis of liner-concrete interaction

    International Nuclear Information System (INIS)

    The investigation of the pre-stressed concrete wall behavior including the liner during LOCA conditions is important for the assessment of the structural integrity of the structure and the leak tightness of the liner. In the frame of the NUGENIA ACCEPPT project WP1 G4 'Structural interaction of liner with the concrete', a load test on a reactor containment liner mockup was carried out. The pre-stressed mockup represents a cylindrical part of the liner, embedded in the concrete wall, but without the wall curvature which is not test relevant. It correlates in material and geometrical properties to the EPR containment. The purpose of the test was to check the liners structural behavior and its integrity for Loss of Coolant Accident (LOCA) load combination considering pre-stressing forces and ageing effects due to creep and shrinkage including liner buckling. The test was carried out at the Karlsruhe Institute of Technology (KIT) in September 2013. This article presents the measurement technology, the results and the development of a calculation method for the embedded liner structure. It appears that the liner deformation results are exemplarily shown at the locations of the imperfections, where the liner buckling is anticipated. The measured liner surface strains ranged between +2 and -10 per thousand. The compressive strains are higher than the tensile strains due to the compressive membrane strains caused by pre-stressing and heating. Although the liner got plastic deformations, the liner strains are still far below the elongation at rupture, which indicates that the liner integrity is ensured. We can conclude that the liner mockup test proceeded as planned. The evaluation results show that the purpose of the liner mockup to simulate LOCA + ageing conditions and liner buckling has fully been achieved

  9. 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.

  10. Application of the best estimate plus uncertainty method to the small break LOCA with high pressure injection failure. Effect evaluation of the model uncertainty on the safety evaluation parameter

    International Nuclear Information System (INIS)

    By applying the BEPU (best estimate plus uncertainty) method, uncertainties of best estimate results can be estimated quantitatively, and excessive conservatism can be reasonably removed to obtain evaluation results with enhanced reliability. Application of the BEPU method is being made to analyses of 'low pressure injection by intentional depressurization of the steam generator secondary side' which is an accident management approach in a SBLOCA (small break loss-of-coolant accident) with high pressure injection failure. In the previous study, the applicability of the analysis code and the uncertainties of the parameters were evaluated. In this research, sensitivity analysis was performed for each model uncertainty separately and the influence of the model on the safety evaluation parameter was estimated. The evaluation result is used to confirm the validity of ranking in the PIRT (phenomena identification and ranking table), and to evaluate the result of the statistical analysis with combined model uncertainties. (author)

  11. Flow distribution due to multiple loop seal venting during a small break loss of coolant accident

    International Nuclear Information System (INIS)

    Calculations of a postulated small break Loss-of-Coolant-Accident (LOCA) have been performed using the NOTRUMP computer code for a typical Combustion Engineering pressurized water reactor (PWR). Evaluation Model initial conditions and assumptions were used in the analysis to examine a 0.1016 m (4.0 in.) equivalent diameter cold leg break. The behavior of the loop seals, and their effects on the overall system hydraulics is discussed

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

    International Nuclear Information System (INIS)

    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

  13. Analysis of an unmitigated large break loss of coolant accident (LBLOCA) with the non-mechanistic failure of passive cooling for the APT Spallation Target

    International Nuclear Information System (INIS)

    In order to support the Programmatic Environmental Impact Statement, an accident analysis has been performed for the Accelerator Production of Tritium (APT) Spallation-Induced Lithium Conversion (SILC) Source. This report presents a lumped-parameter analysis that predicts the thermal response of the source to a large-break LOCA. The accident scenario assumes the break to occur in the cold leg outside the source basin and the pipe break is immediately followed by the tripping of the proton beam and the activation of the source-basin flood system. The calculations were performed for a ''beyond-design-basis event'' which further assumes the failure of all other active cooling systems, and the failure to establish natural circulation in the unbroken loop. Calculations show that the source rods remain flooded in heavy water until 44 hours after the LOCA. At this time, the source rods begin to be uncovered and at 48 hours into the accident the source rods are completely boiled dry. The average source temperature reaches a maximum value of 303 C at 57 hours. Thereafter, the source rods begin to cool since the heat transfer to the basin water is sufficient to remove all the decay heat from the source. It is estimated that by this time a maximum of 27% of the lead inventory (6,558 kg) in the source rods can be expected to melt. This molten material, assuming that it can get out of the aluminum cladding, will fall to the D2O-filled bottom header, quench rapidly, and remain in a coolable state

  14. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

  17. Effect of spray on performance of the hydrogen mitigation system during LB-LOCA for CPR1000 NPP

    International Nuclear Information System (INIS)

    Highlights: → This paper presents the spray effect on HMS during LB-LOCA by using GASFLOW. → The positive and negative effects of spray are summarized. → And the combination of DIS and PAR system is suggested as reasonable countermeasures. → This research is an important work aimed at the study of spray and hydrogen mitigation. → The contents of this paper should become a required part of the safety analysis of Chinese NPPs. - Abstract: During the course of the hypothetical large break loss-of-coolant accident (LB-LOCA) in a nuclear power plant (NPP), hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel (RPV). It is then ejected from the break into the containment along with a large amount of steam. Management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen mitigation system (HMS) and spray system in CPR1000 NPP. The computational fluid dynamics (CFD) code GASFLOW is utilized in this study to analyze the spray effect on the performance of HMS during LB-LOCA. Results show that as a kind of HMS, deliberate igniter system (DIS) could initiate hydrogen combustion immediately after the flammability limit of the gas mixture has been reached. However, it will increase the temperature and pressure drastically. Operating the DIS under spray condition could result in hydrogen combustion being suppressed by suspended droplets inside the containment. Furthermore, the droplets could also mitigate local the temperature rise. Operation of a PAR system, another kind of HMS, consumes hydrogen steadily with a lower recombination rate which is not affected noticeably by the spray system. Numerical results indicate that the dual concept, namely the integrated application of DIS and PAR systems, is a constructive improvement for hydrogen safety under spray condition during LB-LOCA.

  18. A combined deterministic and probabilistic procedure for safety assessment of beyond design basis accidents in nuclear power plant: Application to ECCS performance assessment for design basis LOCA redefinition

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu, E-mail: littlewing@kins.re.kr [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Ahn, Seung-Hoon [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2013-07-15

    Highlights: • We propose the procedure for safety assessment of the beyond design basis accidents. • Deterministic (BEPU) and probabilistic (PSA) approaches are combined in the procedure. • We premised that LOCAs for any breaks larger than TBS would be regarded as BDBA. • Performance of ECCS is assessed against BDB LOCA for original design/design changes. • It is shown that the proposed methodology is applicable to safety assessment of BDBA. -- Abstract: The concept and assessment approach of nuclear safety in nuclear power plants (NPPs) have been evolved with the technological progress and the lessons learned from the major events. Recently, studies on the integrated approach of deterministic and probabilistic method have been done. In this study, a combined deterministic and probabilistic procedure (CDPP) is proposed for safety assessment of the beyond design basis accidents (BDBAs). In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. To verify applicability of the methodology, performance of the APR-1400 emergency core cooling system is assessed against large break loss of coolant accident (LOCA), under the premise that LOCAs for any breaks larger than transition break size would be regarded as BDBA. In addition, discussions are made for analysis results for allowable NPP changes of emergency diesel generator start time extension and power uprating. It is concluded that the proposed CDPP is applicable to safety assessment of BDBAs in NPPs without significant erosion of the safety margin.

  19. A combined deterministic and probabilistic procedure for safety assessment of beyond design basis accidents in nuclear power plant: Application to ECCS performance assessment for design basis LOCA redefinition

    International Nuclear Information System (INIS)

    Highlights: • We propose the procedure for safety assessment of the beyond design basis accidents. • Deterministic (BEPU) and probabilistic (PSA) approaches are combined in the procedure. • We premised that LOCAs for any breaks larger than TBS would be regarded as BDBA. • Performance of ECCS is assessed against BDB LOCA for original design/design changes. • It is shown that the proposed methodology is applicable to safety assessment of BDBA. -- Abstract: The concept and assessment approach of nuclear safety in nuclear power plants (NPPs) have been evolved with the technological progress and the lessons learned from the major events. Recently, studies on the integrated approach of deterministic and probabilistic method have been done. In this study, a combined deterministic and probabilistic procedure (CDPP) is proposed for safety assessment of the beyond design basis accidents (BDBAs). In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. To verify applicability of the methodology, performance of the APR-1400 emergency core cooling system is assessed against large break loss of coolant accident (LOCA), under the premise that LOCAs for any breaks larger than transition break size would be regarded as BDBA. In addition, discussions are made for analysis results for allowable NPP changes of emergency diesel generator start time extension and power uprating. It is concluded that the proposed CDPP is applicable to safety assessment of BDBAs in NPPs without significant erosion of the safety margin

  20. Steamline break analysis for 1000 MWe Kudankulam nuclear power plant

    International Nuclear Information System (INIS)

    Full text: Thermal hydraulic analysis has been carried out for the steamline break outside the containment of VVER-1000 reactor. 1000 MWe VVER is a light water moderated, light water-cooled pressurised water reactor with four primary coolant loops. The double-ended break in the non-isolable steamline was considered, as it will result in higher heat removal by the secondary circuit. This analysis has been carried out using thermal hydraulic code RELAP5/MOD 3.2. As a result of steam generator steamline rupture the steam begins out flowing from the secondary circuit, the secondary pressure decreases. Consequently the coolant temperature, coolant pressures in the primary circuit decrease. Due to negative coolant temperature reactivity, the increase of the reactor power can take place. The end of refuelling cycle is considered in the present analysis since the coolant feedbacks are the most negative and increase of reactor power is maximum. Also loss of power is not considered in the analysis as it results more coolant temperature and great rise in power takes place. The analysis predicts thermal hydraulic conditions following steamline break. Thermal hydraulic conditions like pressure, temperature, and flow at different locations in the PHT as well as in secondary side in the steam generator have been estimated during the transient. The analysis results have been discussed and compared with the acceptance criteriam

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

    International Nuclear Information System (INIS)

    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. Scaling effects concerning the analysis of small break experiments

    International Nuclear Information System (INIS)

    Some scaling effects related to the experimental facilities as well as to the analytical models used for the design and safety analysis of nuclear power plants are discussed or the basis of phenomena expected to occur during small-break loss - of - coolant accidents. The results of isolated small-break experiments should not be directly extrapolated to the safety analysis of commercial reactors, due to the scaling distortions inherent to the test facilities. With respect to the analytical models used to simulate thermohydraulic processes in experimental facilities, their eventual dependence relative to the system dimension should be examined in order to assess their applicability to the safety analysis of commercial power plants. (Author)

  3. Moderator analysis of the Wolsong Units 2/3/4 for the case of 35% RIH breaks with a loss of ECC injection using CFX-4.4

    International Nuclear Information System (INIS)

    A 3-D CFD model was developed for the transient simulation of the moderator circulation inside the Calandria vessel, using a commercial code CFX-4.4. To reduce the discretization errors in the reflector region, a set of butterfly-shaped grid structures were generated and used. The porous media approach was applied for the core region containing 380 Calandria tubes. A 3-dimensional hydraulic model for the porous core region has been developed based on the empirical pressure drop correlations. The transient moderator analysis was performed for the 35% RIH(Reactor Inlet Header) break with a loss of ECC(Emergency Core Cooling) injection, which gives the largest heat load to the moderator out of all the DBA's. During the initial 1200 seconds in the LOCA transient, the local subcooling near any PT/CT contact location does not go down below the minimum subcooling margin of 28oC, the experimentally derived threshold. It is concluded that film boiling on the Calandria tube surfaces would not occur during the selected LOCA scenario and thus the fuel channel integrity is maintained in the CANDU-6 reactor. (author)

  4. Behaviour of CANDU fuel during LOCA

    International Nuclear Information System (INIS)

    A large break loss-of-coolant accident (LOCA) in a CANDU nuclear reactor would result in a rapid increase of fuel and sheath temperatures. The temperature increase would, in turn, increase the gas pressure within the fuel and reduce the strength of the sheath material. Outside the fuel the loss of coolant from the primary heat transport system decreases pressure. The resulting pressure difference would cause deformation of the hot fuel sheath. Under certain circumstances, the deformation could be severe enough to fail the sheath thus releasing the fission products to the primary heat transport system. The computer code ELOCA-A is used to model the transient fuel behaviour following such an accident. ELOCA-A is a modified version of ELOCA.Mk 2 enabling us to consider the effects of axial variations in the microstructure of the sheath material caused by brazing of appendages to the sheath. The ELOCA-A code also features modelling of axial variations in neutron flux, pellet heat generation rate and heat transfer to the coolant. It predicts fuel pellet and sheath temperatures, sheath oxidation, sheath strain and probability of beryllium assisted cracking. A loss-of-coolant accident (LOCA) experiment was jointly sponsored by AECL and Ontario Hydro in the Power Burst Facility (PBF) at Idaho National Engineering Laboratories (INEL). This test was undertaken to provide an all-effects verification of the understanding of CANDU fuel behaviour during LOCA's. An extensive out-pile experimental program had provided single effects data which had been used for modelling such excursions. Integrated out-pile tests have confirmed our understanding of and accurate modelling of fuel under LOCA conditions. The integrated test in PBF provided the final proof that our understanding was complete and provided an experimental database for verification of transient fuel codes (4). The experiment was performed with modified CANDU fuel elements. The post-test measurements are compared with

  5. Analysis of the contribution and efficiency of the Santuario de la Naturaleza Yerba Loca, 33º S in protecting the regional vascular plant flora (Metropolitan and Fifth regions of Chile Análisis de la contribución y eficiencia del Santuario de la Naturaleza Yerba Loca, 33º S, en la protección de la flora vascular regional (regiones Metropolitana y Quinta de Chile

    Directory of Open Access Journals (Sweden)

    MARY T. K ARROYO

    2002-12-01

    Full Text Available Santuario de la Naturaleza Yerba Loca (SN Yerba Loca, Metropolitan Region (MR, 33º S, Chile is analyzed for its conservation value and efficiency in protecting native vascular plants in a regional context. The reserve's flora of 500 species and subtaxa was evaluated for species richness, endemism, range size and marginally distributed taxa, using species-area analysis, and tendencies in the floras of the MR (1.434 species and subtaxa and MR-Fifth regions (1,841 species and subtaxa to set the regional pattern. The reserve (0.7 % of MR land area and 0.3 % MR-Fifth land area contains 34 % of the MR and 27% of the MR-Fifth floras, and around 16-17 % of the mediterranean-climate area (regions IV-VIII flora of central Chile. Veech's Relative Richness Index (RRI revealed that SN Yerba Loca houses exaggerated richness in relation to its land area (28 % more species than expected from the regional model. However, endemism rates (35 % Continental Chile endemics, 22 % Mediterranean endemics, 3% MR-Vth endemics are statistically lower than in the MR (44 %, 29 %, 9 % and the MR-Vth (48 %, 31 %, 11 % floras, and SN Yerba Loca houses proportionately fewer MR endemics (2 % than the MR (6 %. Compared with the regional floras, the reserve contains statistically fewer marginally distributed species, and range size (median = five administrative regions is significantly larger. The reserve's outstanding species richness compensates for its low endemism rates bringing the absolute number of endemics to 92 % of the regional expectation. Corresponding values for marginally distributed species are 81 % (northern limits, 63% (southern limits and for median and shorter range taxa, 100 %. It is concluded that SN Yerba Loca is a highly efficient reserve from the point of view of vascular plant conservation, and represents an excellent conservation choice. SN Yerba Loca and MN El Morado (a second state protected area in the MR, conservatively, house 39 % of the native

  6. Tailings dam-break flow - Analysis of sediment transport

    Science.gov (United States)

    Aleixo, Rui; Altinakar, Mustafa

    2015-04-01

    A common solution to store mining debris is to build tailings dams near the mining site. These dams are usually built with local materials such as mining debris and are more vulnerable than concrete dams (Rico et al. 2008). of The tailings and the pond water generally contain heavy metals and various toxic chemicals used in ore extraction. Thus, the release of tailings due to a dam-break can have severe ecological consequences in the environment. A tailings dam-break has many similarities with a common dam-break flow. It is highly transient and can be severely descructive. However, a significant difference is that the released sediment-water mixture will behave as a non-Newtonian flow. Existing numerical models used to simulate dam-break flows do not represent correctly the non-Newtonian behavior of tailings under a dam-break flow and may lead to unrealistic and incorrect results. The need for experiments to extract both qualitative and quantitative information regarding these flows is therefore real and actual. The present paper explores an existing experimental data base presented in Aleixo et al. (2014a,b) to further characterize the sediment transport under conditions of a severe transient flow and to extract quantitative information regarding sediment flow rate, sediment velocity, sediment-sediment interactions a among others. Different features of the flow are also described and analyzed in detail. The analysis is made by means of imaging techniques such as Particle Image Velocimetry and Particle Tracking Velocimetry that allow extracting not only the velocity field but the Lagrangian description of the sediments as well. An analysis of the results is presented and the limitations of the presented experimental approach are discussed. References Rico, M., Benito, G., Salgueiro, AR, Diez-Herrero, A. and Pereira, H.G. (2008) Reported tailings dam failures: A review of the European incidents in the worldwide context , Journal of Hazardous Materials, 152, 846

  7. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    International Nuclear Information System (INIS)

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  8. MELCOR Comparative Analyses of Severe Accident of Medium LOCA for the NPP V2 Bohunice

    International Nuclear Information System (INIS)

    This paper presents the results of safety analysis of a medium LOCA (break size 100 mm in cold leg) for the V2 Bohunice nuclear power plant (VVER-440/V-213), and compares the results calculated by various computer codes (MELCOR, MAAP, RELAP/SCADAP). The analysis is performed within the SWISSLOVAK project by the safety analysis group at the Nuclear Regulatory Authority of the Slovak Republic. The medium LOCA accident is combined with station blackout scenario which leads to the core uncovery and meltdown of the reactor core. The core meltdown is followed by the core relocation to the lower plenum, heat up of the reactor pressure vessel lower head, failure of the lower head, and debris ejection into the reactor cavity. The time of key events calculated by various computer codes is similar. The start of core melt is predicted within 0.8 to 1.08 hours and the reactor pressure vessel lower head failure is predicted within 4.1 to 6.3 hours since the initiation of the accident. A substantial release of noble gases to the environment through the permanent containment leakage is calculated. The compartmentalization of the containment and the presence of the bubble condenser affect the release of the fission products. (author)

  9. Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

    International Nuclear Information System (INIS)

    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.

  10. Assessment of a large break loss of coolant accident scenario requiring operator action to initiate safety injection

    International Nuclear Information System (INIS)

    As part of the licensing basis for a nuclear power plant, the acceptability of the Emergency Core Cooling Systems (ECCS) following a postulated Loss-of-Coolant Accident (LOCA) as described in the Code of Federal Regulations (CFR), Title 10, Chapter 1, Part 50.46, must be verified. The LOCA analysis is performed with an acceptable ECCS Evaluation Model and results must show compliance with the 10 CFR 50.46 acceptance criteria. Westinghouse Electric Corporation performs Large and Small Break LOCA and LOCA-related analyses to support the licensing basis of various nuclear power plants and also performs evaluations against the licensing basis analyses as required. Occasionally, the need arises for the holder of an operating license of a nuclear power plant to submit a Licensee Event Report (LER) to the US Nuclear Regulatory Commission (USNRC) for any event of the type described in the Code of Federal Regulations, Title 10, Chapter 1, Part 50.73. To support the LER, a Justification for Past Operation (JPO) may be performed to assess the safety consequences and implications of the event based on previous operating conditions. This paper describes the work performed for the Large Break LOCA to assess the impact of an event discovered by Florida Power and Light and reported in LER-94-005-02. For this event, it was determined that under certain circumstances, operator action would have been required to initiate safety injection (SI), thus challenging the acceptability of the ECCS. This event was specifically addressed for the Large Break LOCA by using an advanced thermal hydraulic analysis methodology with realistic input assumptions

  11. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    International Nuclear Information System (INIS)

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken

  12. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken.

  13. Relationship between computed ANSI/ANS-5.1 and ORIGEN-S decay heat powers for BWR LOCA safety analysis

    International Nuclear Information System (INIS)

    The decay heat power fraction computed using ANSI/ANS-5.1-1979 with CASMO-4 decay heat parameters is compared with the decay heat power fraction computed using ANSI/ANS-5.1-1979 with ORIGEN-S decay heat parameters. The comparison indicates that the ORIGEN-based ANS-5.1 total decay power fraction appears very close to the CASMO-based ANS-5.1 total decay power fraction due to compensating effect between fission-product decay heat power fraction and U-239 and Np-239 decay heat power fraction, although the CASMO-4 fission fractions and U-238 neutron capture ratio are considered more accurate than the ORIGEN-S fission fractions and U-238 neutron capture ratio. Therefore, it seems acceptable to calculate the total decay heat fraction using ANSI/ANS-5.1-1979 with ORIGEN-S decay heat parameters. This result is useful, since ORIGEN-S /SCALE 5.1 are easier to run than CASMO-4. The decay heat power fraction computed using ANSI/ANS-5.1-1979 is also compared with the decay heat power fraction computed using ORIGEN-S directly. The comparison indicates that the ORIGEN-S total decay heat power fraction is much smaller than the corresponding ANS 5.1 total decay heat power fraction, which is due to the fact that the ORIGEN-S fission-product decay heat power fraction is much smaller than the corresponding ANS 5.1 fission-product decay heat power fraction. This demonstrates that the total decay heat power fraction calculated using ORIGEN-S directly is not conservative and that ANSI/ANS-5.1 must be used to calculate the total decay heat power fraction for LOCA safety analysis. (author)

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

    International Nuclear Information System (INIS)

    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

  15. Investigating a Partial LOCA in the IAEA Generic Research Reactor

    International Nuclear Information System (INIS)

    The behavior of the IAEA research reactor under partial LOCA (Loss of Coolant Accident conditions scats investigated. The reactor is a pool-type light water 10 MW research reactor employing MTR-type fuel elements. The extremely rare LOCA scenario is assumed to take place when a guillotine break in one of the Malta coolant loops occurs. Under these conditions the water level in the pool decreases, reaching a level that covers only part of the core. With no flow through the cooling channels, decay-heat will raise the temperature of the partly covered fuel elements. This study demonstrates that if water level remains at or above 1/4 of the fuel channel length, no damage to the core will occur. This has also been shown by an experiment performed at Livermore

  16. Analysis on Cooling Capacity of Passive Core Cooling System during LOCA Scenarios%非能动堆芯冷却系统LOCA下冷却能力分析

    Institute of Scientific and Technical Information of China (English)

    游曦鸣; 邵舸; 佟立丽; 曹学武

    2016-01-01

    The analysis model for advanced pressurized water reactor was established by using mechanism analytical code .The model included reactor coolant system ,engineer‐ing safety features and related secondary side pipes system .T he typical small break loss of coolant accident and large break loss of coolant accident were selected to analyze the accident progression .The water injection capacity and cooling capacity of passive residu‐al heat removal system (PRHRS ) ,core makeup tank (CM T ) ,accumulator (ACC ) , automatic depressurization system (ADS) and in‐containment reactor water storage tank (IRWST ) included in passive core cooling system (PXS ) were focused on for LOCA with different sizes and locations .The results show that the size and location of break have an influence on the accident progression .But the peak cladding surface temperature does not exceed 1 477 K and the reactor core is flooded underwater in all the accident conditions .The PXS can effectively remove reactor core decay heat and keep the reactor in the safe shutdow n situation to prevent severe accident .%本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统(PXS)中非能动余热排出系统(PRHRS)、堆芯补水箱(CMT)、安注箱(ACC)、自动卸压系统(ADS)和安全壳内置换料水箱(IRWST)等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1477 K ,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。

  17. High burnup effects on the burst behavior of Zr based alloy claddings under LOCA conditions

    International Nuclear Information System (INIS)

    A current loss of coolant accident (LOCA) criterion is based on the results obtained from non pressurized claddings specimens under simulated LOCA condition. However, integrity of fuel cladding can be significantly affected by ballooning and rupture that caused by pressure difference between inner and outer cladding during LOCA. Ballooning may cause the fuel relocation or fuel dispersal due to its rupture opening during accidents. In addition, wall thickness of cladding can be reduced and local regions near the rupture open would become heavily oxidized and hydride d. Therefore, integral test that can simulate whole process during LOCA should be carried out for comprehensive safety analysis. Although a number of researches have been conducted, most investigations of them were performed using as received cladding specimens. In this study, burst behavior of several kinds of zirconium based alloys was investigated by integral LOCA test and high burnup effects on the burst behavior of fuel cladding were also examined using H charged cladding sample

  18. A study on timing of rapid depressurization action during PWR vessel bottom break LOCA with HPI failure and AIS-gas inflow (ROSA-V/LSTF test SB-PV-06)

    International Nuclear Information System (INIS)

    A small break loss-of-coolant accident (SBLOCA) experiment (SB-PV-06) was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study the effects of initiation timing of rapid secondary depressurization action on core cooling as one of accident management (AM) measures for a pressurized water reactor (PWR) in case of high pressure injection (HPI) system failure and non-condensable gas inflow from the accumulator injection system (AIS). The break simulated rupture of 10 instrument tubes at the vessel bottom equivalent to 0.2% cold leg break. The rapid depressurization action was initiated after the vessel level below the primary loop nozzle was detected. The results were compared with those of two similar experiments of SB-PV-03 in which the action was initiated after core heat-up, and SB-PV-04 in which the earliest action was initiated by safety injection (SI) signal with 10 minutes delay resulting in adequate core cooling. It is clarified that the vessel level indication for start of the AM action is less effective on core cooling, while steam generator (SG) outlet plenum level indication for earlier AM action can be effective due to larger primary coolant mass as in the SB-PV-04 experiment. The report compares these experimental results to clarify the effects of the initiation timing of rapid secondary depressurization action on core cooling in addition to the precise results of the SB-PV-06 experiment. (author)

  19. Analysis of main roof breaking form and its mechanism during first weighting in longwall face

    Institute of Scientific and Technical Information of China (English)

    HUANG Qing-xiang

    2001-01-01

    By field observation and simulating test in shallow seam logwall mining, the asymmetry breaking of main roof is discovered during the first weighting. Based on simulating model test and theoretical analysis, the mechanism of main roof first breaking is revealed, and the asymmetry breaking parameter is determined at all.

  20. Application of fractional scaling methodology (FSM) to loss of coolant accidents (LOCA) Part 2. System level scaling for system depressurization

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgang Wulff [11 Hamilton Road, Setauket, NY 11733 (United States); Novak Zuber [703 New Mark Esplanade, Rockville, MD 20850 (United States); Upendra S Rohatgi [Brookhaven National Laboratory Bldg. 474B, Upton, NY 11973 (United States); Ivan Catton [UCLA - MAE, Box 951597, 48-121 Engr. IV Los Angeles, CA 90195-1597 (United States)

    2005-07-01

    Full text of publication follows: The Fractional Scaling Methodology (FSM) is demonstrated at the system level for the depressurization of nuclear reactor primary systems undergoing large- and small-break loss of coolant accidents in two integral test facilities of different size and shape, namely LOFT and Semiscale. As one of two global system characteristics, depressurization is common to all components in a multi-loop thermohydraulic system. Momentum transport as the second characteristic represents component interactions that facilitate fluid and energy exchange between components. The advantages are shown of holistic relative to reductionistic modeling and scaling approaches. The depressurization model accounting for all agents of system change as well as the full two-phase fluid system compliance is derived and scaled on the basis of FSM. The paper demonstrates the relation between pressure and volume displacement rates in analogy to generalized 'effort' and 'flow' in the interdisciplinary analysis of complex systems. It is demonstrated with experimental data that a properly scaled depressurization history is the same for Large-Break and Small-Break Loss of Coolant Accidents (LOCA) in two test facilities of different size. A single experiment or a single computer simulation, when properly scaled, is shown to suffice for all LOCAs of all break sizes and orientations in the primary system of a particular power plant (PWR or BWR) and of all related test facilities. FSM, when applied on the system, component and process levels, serves to synthesize the wealth of world-wide results from analyses and experiments into compact form for efficient storage, transfer and retrieval of information. This is demonstrated on the system level. Moreover, it is shown that during LOCAs the break flow is the dominant agent of change for break sizes between 2.5 and 200% of cold leg flow cross-sectional area, and that in general FSM ranks processes quantitatively

  1. REFLA-1D/MODE3: a computer code for reflood thermo-hydrodynamic analysis during PWR-LOCA

    International Nuclear Information System (INIS)

    This manual describes the REFLA-1D/MODE3 reflood system analysis code. This code can solve the core thermo-hydrodynamics under forced flooding conditions and gravity feed conditions in a system similar to FLECHT-SET Phase A. This manual describes the REFLA-1D/MODE3 models and provides application information required to utilize the code. (author)

  2. Hydrodynamics of AHWR gravity driven water pool under simulated LOCA conditions

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) employs a double containment concept with a large inventory of water within the Gravity Driven Water Pool (GDWP) located at a high elevation within the primary containment building. GDWP performs several important safety functions in a passive manner, and hence it is essential to understand the hydrodynamics that this pool will be subjected to in case of an accident such as LOCA. In this paper, a detailed thermal hydraulic analysis for AHWR containment transients is presented for postulated LOCA scenarios involving RIH break sizes ranging from 2% to 50%. The analysis is carried out using in-house containment thermal hydraulics code 'CONTRAN'. The blowdown mass and energy discharge data for each break size, along with the geometrical details of the AHWR containment forms the main input for the analysis. Apart from obtaining the pressure and temperature transients within the containment building, the focus of this work is on simulating the hydrodynamic phenomena of vent clearing and pool swell occurring in the GDWP. The variation of several key parameters such as primary containment V1 and V2 volume pressure, temperature and V1-V2 differential pressure with time, BOP rupture time, vent clearing velocity, effect of pool swell on the V2 air-space pressure, GDWP water level etc. are discussed in detail and important findings are highlighted. Further, the effect of neglecting the pool swell phenomenon on the containment transients is also clearly brought out by a comparative study. The numerical studies presented in this paper give insight into containment transients that would be useful to both the system designer as well as the regulator. (author)

  3. Analysis of different hypothetical recirculation line SB-LOCAs in the Muehleberg (KKM) BWR/4 by TRAC-BF1/v98.1

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.Th.; Coddington, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    In this work, we shall report on the analysis of a number of hypothetical recirculation line breaks at the KKM BWR/4 in Switzerland covering a spectrum of break sizes under the assumption of full ECCS (emergency core cooling system) availability, and also for one of these transients, by assuming limited ECCS, i.e. half and one third of the nominal LPCS (low pressure core sprays) availability. Furthermore, for the same transient, we shall analyse two additional sub-cases, assuming that two or three of the four valves that provide the ADS (automatic depressurization system) function remain closed. Finally, we investigated the differences in the predicted peak clad temperatures for two different code versions in which some of the terms of the 3D phasic momentum equations were up-winded or cell-length averaged. (authors)

  4. Analysis of different hypothetical recirculation line SB-LOCAs in the Muehleberg (KKM) BWR/4 by TRAC-BF1/v98.1

    International Nuclear Information System (INIS)

    In this work, we shall report on the analysis of a number of hypothetical recirculation line breaks at the KKM BWR/4 in Switzerland covering a spectrum of break sizes under the assumption of full ECCS (emergency core cooling system) availability, and also for one of these transients, by assuming limited ECCS, i.e. half and one third of the nominal LPCS (low pressure core sprays) availability. Furthermore, for the same transient, we shall analyse two additional sub-cases, assuming that two or three of the four valves that provide the ADS (automatic depressurization system) function remain closed. Finally, we investigated the differences in the predicted peak clad temperatures for two different code versions in which some of the terms of the 3D phasic momentum equations were up-winded or cell-length averaged. (authors)

  5. 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)

  6. Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

    International Nuclear Information System (INIS)

    The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

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

    International Nuclear Information System (INIS)

    The PWR safety studies involve an analysis of the consequences of a hypothetical rupture of a primary pipe. From the opening tune, the blowdown at the break causes the propagation of an acoustic wave through the whole primary circuit, as well as pipe whipping. The local pressure gaps due to the depressurization wave propagation may induce component recoils and internal structure movements. In parallel with the acoustic wave propagation, the circuit empties progressively first with a monophasic regime and later with a diphasic one. This paper presents a hydrodynamic simulation of the flows in the primary circuit of 4-loop PWR during a LOCA. The results concern the propagation of the depressurization acoustic wave along the circuit, coupled with the transient fluid flows. (authors)

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

    International Nuclear Information System (INIS)

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

  9. Bayesian Analysis of Dynamic Multivariate Models with Multiple Structural Breaks

    OpenAIRE

    Sugita, Katsuhiro

    2006-01-01

    This paper considers a vector autoregressive model or a vector error correction model with multiple structural breaks in any subset of parameters, using a Bayesian approach with Markov chain Monte Carlo simulation technique. The number of structural breaks is determined as a sort of model selection by the posterior odds. For a cointegrated model, cointegrating rank is also allowed to change with breaks. Bayesian approach by Strachan (Journal of Business and Economic Statistics 21 (2003) 185) ...

  10. Simulation of a main steam line double-ended guillotine break within the containment using the coupled code COCOSYS-ATHLET

    International Nuclear Information System (INIS)

    The coupled codes ATHLET and COCOSYS, i.e. the models for the reactor cooling circuit and the containment were applied to the NPP Brunsbuettel and Kruemmel for the case of a loss of coolant accident (LOCA) based on a double-ended guillotine break of the main steam line within the containment. The advantage of the coupled code application is the simulation of the long-term behavior in consequence of LOCAs. The calculation expenses are considered as disadvantage of the coupled model application. The coupled code analysis requires qualified data sets for separate analyses.

  11. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Application of PWR LOCA margin with the revised appendix K rule

    International Nuclear Information System (INIS)

    Today's focus for nuclear power plant utility owners is to improve plant performances such that the cost per kilowatthour is minimized with enhanced safety. This paper will discuss the impact of design and licensed margin on PWR plant performance, how these margins can be used to improve PWR performance, and how Westinghouse is addressing the regulatory design limits for large break and small break LOCA which impact core thermal design margin. (orig./GL)

  14. Numerical modelling dam break analysis for water supply project

    International Nuclear Information System (INIS)

    Dam provides many benefits to the society, but it can also cause extensive damage to downstream area when it fails. Dam failure can cause extensive damage to properties and loss of human life due to short warning time available. In general, dam spillway was designed to drain the maximum discharge from the dam during the Probable Maximum Flood (PMF). The spillway is functioned to prevent the dam from failure due to overtopping, which can lead to the dam failure. Dam failure will result in large volume of water travelling at very high velocity to the downstream area of the dam. It can cause extensive property damage, destruction of important facilities, and significant loss of human life along the way. Due to the potential of high hazard it poses to the downstream area, a dam break analysis is considered very essential. This paper focuses into the dam failure analysis for Kahang Dam by prediction of breach flow hydrographs and generation of inundation map at downstream area. From the PMF scenario simulation, the maximum inflow is 525.12 m3/s and peak discharge from the dam during dam failure is 6188m3/s. The results are able to provide information for preparation of Emergency Response Plan (PMF), in which appropriate steps can be taken by relevant authorities to avoid significant loss of human lives.

  15. RELAP5 analysis of PKL, main steam line break test

    International Nuclear Information System (INIS)

    Highlights: • RELAP5/MOD 3.2 code validation is performed by analyzing a main steam line break test in the PKL large scale test facility. • The RELAP5 model reproduces well the important events of the PKL test. • RELAP5 transient results show noticeable sensitivity to small differences in the initial conditions. • Accurate prediction of the coolant temperature is essential for the assessment of potential core re-criticality. - Abstract: PKL is a large scale test facility of the primary system owned by AREVA NP GmbH. It is used for extensive experimental investigations to study the integral behavior of Pressurized Water Reactor (PWR) plants under accident conditions. Since 2001, the test program is a part of an international cooperation project (SETH, followed by PKL1 and PKL2) set up by the OECD. The aim of the present work was to perform a short validation study of the thermo-hydraulics code RELAP5. A model of the PKL test facility has been developed, tested and applied to one of the experiments performed at the PKL. The chosen experiment was the test G3.1. In that experiment, a main steam line break occurs, causing a rapid depressurization of the affected steam generator. This leads to an increase of the heat transfer from the primary to the secondary side and thereby to a fast cool-down transient on the primary side. The main objective of this analysis was the qualification of the RELAP5 code results against heat transfer from the primary to the secondary side in both affected and intact loops, and temperatures in the primary system. The calculation results have been compared to the experimental results. It was concluded that the most important events during the test are reproduced relatively well by the model. The calculated coolant temperature in the core is higher than in the experiment. The minimum temperature is about 5% higher than measured. The secondary pressures in SG-1, 3, and 4 is in very good agreement with the experimental value, but in the

  16. An Analysis of Effect of Break-up Timing on the Necessity of a Feed-and-Bleed Operation in the case of TLOFW with Local

    International Nuclear Information System (INIS)

    A Feed-and-bleed (F and B) operation is a process to cool the reactor by the primary side directly. If adequate residual heat removal through the secondary side is not available, the heat can be removed from the RCS by F and B operation. A total loss of feedwater (TLOFW) accident is used to represent an accident involving the failure of cooling by the secondary cooling system. Even if the secondary cooling system fails, the RCS can be cooled by F and B transients when a loss of coolant accident (LOCA) with a TLOFW accident occurs. During an F and B transient, the RCS has a residual heat removal mechanism. If the break size is large, an F and B transient continuously occurs if the SIS is available. If the break size is small to sufficiently decrease the RCS pressure, the SIS cannot inject the coolant, causing the F and B transient to terminate. After the termination of the F and B transient, the residual heat cannot be removed, and the necessity of an F and B operation increases. The operators may hesitate to initiate F and B operation if a clear cue is not provided, since its initiation implies the radioactive coolant releases into the containment. Therefore, the necessity of F and B operation is needed to be identified. The factors affected the necessity of F and B operation are the availability of the safety injection system and safety depressurization system, water inventory in the primary and secondary cooling systems, break size in a loss-of-coolant accident, and time of accident occurrence. The necessity of F and B operation can be changed according to different timing of break-up despite same break size. Moreover, different timing of break-up makes the operators more complicated. To identify effect of timing of break-up, a thermohydraulic analysis was performed using the MARS code. This study is expected to provide a useful guideline to identify the necessity of an F and B operation under combined accident

  17. Fission Product Transport Models Adopted in REFPAC Code for LOCA Conditions in PWR and WWER NPPS

    International Nuclear Information System (INIS)

    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)

  18. Evaluation of the effect of break nozzle configuration in loss-of-coolant accident analysis

    International Nuclear Information System (INIS)

    The Semiscale Mod-1 test program has utilized two different break nozzle configurations in a test facility with identical initial and boundary conditions. An evaluation has been made to determine the effect these break nozzle configurations have on system thermal-hydraulic response during a 200% double-ended cold leg break loss-of-coolant accident simulation. The first nozzle had a convergent-divergent design; the second nozzle had a convergent design with an elongated constant area throat followed by a rapid expansion. Analysis of data from tests conducted with the two nozzles shows that the critical flow characteristics at the break plane were affected by the break nozzle geometry. Differences in break flow caused differences in the core inlet flow which in turn affected core heater rod thermal response. The results of this investigation show that the break flow behavior and the resulting core thermal response in the Semiscale experimental facility can be directly correlated

  19. CSNI LOCA standard problems

    International Nuclear Information System (INIS)

    In the 3. meeting of the CSNI Working Group on ECCS (August 1976), several countries expressed views on the scope of the Standard Problem programme and how often Standard Problems should be performed. Based on discussions expressed during the meeting and on some opinions, several recommendations have been made. They are presented here, divided into five sections: Objectives of CSNI Standard Problems; Proposals for New CSNI Standard Problems; Specifications for Standard Problem Analysis; Reporting of Results from CSNI Standard Problem Exercises; Calculated Results and Experimental Data Comparison Report. Section I discusses the purpose for doing Standard Problems, and the other four sections discuss the different documents that should be prepared in support of the programme. Each of these four sections attempts to outline specifications for the content of the documents

  20. CANDU 6 steam line break analysis with CATHENA

    International Nuclear Information System (INIS)

    Detailed thermalhydraulic simulations for CANDU 6 steam line break inside containment are performed to predict the response of the primary and secondary circuits. The analysis is performed using the thermalhydraulic computer code, CATHENA, with a coupled primary and secondary circuit model. A two-loop representation of the primary and secondary circuits is modelled. The secondary circuit model includes the feedwater line from the deaerator storage tank, multi-node steam generators and the steam line up to the turbine. Two cases were carried out using different assumptions for the efficiency of the steam separators. Case 1 assumes the efficiency of the steam separators becomes zero when the water level in the steam drum increases to the elevation of primary cyclones, or the outlet flow from the steam generator becomes higher than 150% of normal flow. Case 2 assumes the efficiency becomes zero only when the water level in the steam drum reaches the elevation of primary cyclones. The simulation results show that system responses are sensitive to the assumption for the efficiency of the steam separators and Case 1 gives higher discharge energy. Fuel cooling is assured, since primary circuit is cooled down sufficiently by the steam generators for both cases. 1 ref., 12 figs., 2 tabs

  1. Statistical analysis of the breaking processes of Ni nanowires

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Mochales, P [Departamento de Fisica de la Materia Condensada, Facultad de Ciencias, Universidad Autonoma de Madrid, c/ Francisco Tomas y Valiente 7, Campus de Cantoblanco, E-28049-Madrid (Spain); Paredes, R [Centro de Fisica, Instituto Venezolano de Investigaciones CientIficas, Apartado 20632, Caracas 1020A (Venezuela); Pelaez, S; Serena, P A [Instituto de Ciencia de Materiales de Madrid, Consejo Superior de Investigaciones CientIficas, c/ Sor Juana Ines de la Cruz 3, Campus de Cantoblanco, E-28049-Madrid (Spain)], E-mail: pedro.garciamochales@uam.es

    2008-06-04

    We have performed a massive statistical analysis on the breaking behaviour of Ni nanowires using molecular dynamic simulations. Three stretching directions, five initial nanowire sizes and two temperatures have been studied. We have constructed minimum cross-section histograms and analysed for the first time the role played by monomers and dimers. The shape of such histograms and the absolute number of monomers and dimers strongly depend on the stretching direction and the initial size of the nanowire. In particular, the statistical behaviour of the breakage final stages of narrow nanowires strongly differs from the behaviour obtained for large nanowires. We have analysed the structure around monomers and dimers. Their most probable local configurations differ from those usually appearing in static electron transport calculations. Their non-local environments show disordered regions along the nanowire if the stretching direction is [100] or [110]. Additionally, we have found that, at room temperature, [100] and [110] stretching directions favour the appearance of non-crystalline staggered pentagonal structures. These pentagonal Ni nanowires are reported in this work for the first time. This set of results suggests that experimental Ni conducting histograms could show a strong dependence on the orientation and temperature.

  2. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R.; Erixon, S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B. [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B. [RSA Technologies, Visat, CA (United States)

    1996-09-01

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs.

  3. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    International Nuclear Information System (INIS)

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs

  4. LOCA testing of damaged cables

    International Nuclear Information System (INIS)

    Experiments were conducted to assess the effects of dielectric withstand voltage 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 cables. During aging and LOCA testing, Okonite ethylene propylene rubber cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging. For Brand Rex crosslinked polyolefin cables, the results suggest that 8 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage levels necessary to detect when 8 mils of insulation remain are expected to be roughly 40 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. Although two Rockbestos silicone rubber cables failed during the accident test, the induced wall thickness did not seem to be the major cause of the failures. It appears likely that under less stressful thermal aging conditions, the cables would survive accident testing with as little as 4 mils or less of insulation remaining

  5. The Effect of Corporate Break-ups on Information Asymmetry: A Market Microstructure Analysis

    OpenAIRE

    Bardong, Florian; Bartram, Söhnke M.; Yadav, Pradeep K.

    2006-01-01

    This paper investigates the information environment during and after a corporate break-up utilizing direct measures of information asymmetry developed in the market microstructure literature. The analysis is based on all corporate break-ups in the United States in the period 1995-2005. The results document that information asymmetry declines significantly as a result of a break-up. However, this reduction takes place not at the time of its announcement or its completion, but after it has been...

  6. CFD analysis of Siphon Break in a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hong Beom; Seo, Kyoungwoo; Kim, Seong Hoon; Chi, Dae Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The results employing the homogenous model and inhomogeneous model with the SST turbulence model were compared. Homogeneous model has higher undershooting height than inhomogeneous model. And results with various free surface option show siphon break phenomena is not associated with free surface. Based on a numerical simulation, it was evaluated that a siphon break is dependent on the air-water flow mixture and interface length scale. In open pool type research reactor, reactor core is cooled by natural circulation after the primary cooling pump is turned off and the pool water is used as the ultimate heat sink. The pool water also behaves as a shielding barrier for many kinds of radio-nuclides from the reactor core and the spent fuel. Pool water is essential for nuclear safety. So guaranteeing the pool water inventory to be higher than the required minimum level is one of the most important tasks of a research reactor design. The lowest pool penetration of cooling pipes should be located above the reactor core against a cooling pipe break. However, system components outside the pool can be installed below the core level due to the component purpose such as the acceptance of a Net Positive Suction Head(NPSH) of a pump for downward core flow research reactor. So the pool water can be drained below the core through siphon effect and the core can't be cooled through natural circulation when a postulated pipe break occurs below the reactor core position. Therefore siphon breaker should be installed to limit the pool water drain. In this study, 3D numerical simulations are performed to be applicable to the siphon breaker design for a research reactor because undershooting(height between the end of siphon break line and the final pool water level) is expected for a large pipe break. ANSYS CFD is used to solve the Navier-Stokes equation with the turbulent model and two-phase model. Siphon breaker was designed to satisfy the minimum pool water level requirement during

  7. Simulation of a main steam line double-ended guillotine break within the containment using the coupled code COCOSYS-ATHLET; Simulation eines 2F-Bruches einer Frischdampfleitung innerhalb des SHB mit dem gekoppelten Code COCOSYS-ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Karsten [Becker Technologies, Eschborn (Germany); Gall, Uwe [Vattenfall Europe Nuclear Energy GmbH, Hamburg (Germany); Klein-Hessling, Walter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

    2009-07-01

    The coupled codes ATHLET and COCOSYS, i.e. the models for the reactor cooling circuit and the containment were applied to the NPP Brunsbuettel and Kruemmel for the case of a loss of coolant accident (LOCA) based on a double-ended guillotine break of the main steam line within the containment. The advantage of the coupled code application is the simulation of the long-term behavior in consequence of LOCAs. The calculation expenses are considered as disadvantage of the coupled model application. The coupled code analysis requires qualified data sets for separate analyses.

  8. A 25% double-ended LOCA in the PSB-VVER facility

    International Nuclear Information System (INIS)

    The paper presents the results of the experimental investigation in the PSB-VVER facility and post-test analysis with RELAP5/MOD3.3 of the thermal-hydraulic response of the PSB-VVER to a 25% double-ended Hot Leg Break (HLB). The test scenario included loss of off-site power concurrently with the scram-signal and the safety system operation as described in the reference VVER-1000 operational manual in the case of this type of accident assuming one diesel-generator failure. The key transient parameter trends as well as sequence of events and phenomena are given in the paper. RELAP5/MOD3.3 a post-test analysis has been performed using the experimental data gained as a base. The reasonable qualitative agreement between the key calculated and measured variables has been shown. The quantitative code accuracy evaluation has shown that the total average amplitude of the main parameters' deviations AAtot tot< 0.28 that corresponds to satisfactory quality of the VVER-1000 hot leg guillotine break LOCA modeling in the PSB-VVER. (author)

  9. Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

    Science.gov (United States)

    Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

    2016-04-01

    Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.

  10. Strain model and zircaloy-steam reaction model in the TUF code and their application in large LOCA

    International Nuclear Information System (INIS)

    As an integral part of a generic study of the Emergency Coolant Injection System effectiveness in Ontario Hydro reactors during a large break Loss of Coolant Accident (LOCA), the TUF (Two-Unequal-Fluids) code has been developed to enhance safety analysis capability. Recent enhancement to the TUF code includes the pressure tube transverse strain model and the zircaloy-steam reaction model. These models are employed to predict thermal-mechanical response of fuel channels and determine the thermal heat load to the moderator during postulated large LOCA scenarios. Presented in this paper are the description of the models, the cross-code comparison of the predictions between the TUF code and the SMARTT code and the discussion of parameters that may affect the pressure tube strain and the effect of pressure tube ballooning into contact with the calandria tube on the system response simulations. The modified TUF code is employed to quantify the extent of pressure tube ballooning and to calculate the thermal heat load to moderator. (author) 14 refs., 4 tabs., 16 figs

  11. Use of Utility Codes for Fuel Analysis during Off-Stagnation Feeder Break in CANDU

    International Nuclear Information System (INIS)

    Feeder break accident is regarded as one of the design basis accident in CANDU reactor which results in a fuel failure. For a particular range of inlet feeder break sizes, the flow in the channel is reduced sufficiently that the fuel and fuel channel integrity can be significantly affected to have damage in the affected channel, while the remainder of the core remains adequately cooled. The flow in the downstream channel can be more or less stagnated due to a balance between pressure at the break on the upstream side and the reverse driving pressure between the break and the downstream end. In the extreme, this can lead to rapid fuel heatup and fuel damage and failure of the fuel channel similar to that associated with a severe channel flow blockage. Such an inlet feeder break scenario is called a stagnation break. For an inlet feeder break which is slightly larger or smaller than that for the stagnation break case, the result is a channel flow which is low enough to result in fuel failure but high enough that the pressure tube remains intact. This event is identified as the off-stagnation break. In this report, the fuel analysis methodology and the usage of utility codes to evaluate the fission gas release during the off-stagnation feeder break are described. The accident was assumed to be occurred in the refurbished Wolsong unit 1 and the latest safety codes were used in the analysis. Fission product inventories during the steady state were calculated by using ELESTRES-IST 1.2 code. After starting the off-stagnation break, ELOCA code evaluated the timing of fuel failure and the following fission gas release due to the oxidation of the pellet are calculated by using several utility codes until the reactor trip. The calculated fission product releases are provided to the following dose assessment as a source term

  12. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

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

  13. Propagation Mechanism Analysis Before the Break Point Inside Tunnels

    OpenAIRE

    Guan, Ke; Zhong, Zhangdui; Bo, Ai; Briso Rodriguez, Cesar

    2011-01-01

    There is no unanimous consensus yet on the propagation mechanism before the break point inside tunnels. Some deem that the propagation mechanism follows the free space model, others argue that it should be described by the multimode waveguide model. Firstly, this paper analyzes the propagation loss in two mechanisms. Then, by conjunctively using the propagation theory and the three-dimensional solid geometry, a generic analytical model for the boundary between the free space mechanism and the...

  14. Analysis of Leak Before Break and Calculation Method of Critical Crack

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Now leak before break (LBB) technology is widely used in nuclear power plant design. It has a good development in foreign countries, but domestic research is relatively little. The study of crack propagation is core of LBB analysis. It

  15. RELAP4/MOD6/U4/J3: a JAERI improved version of RELAP4/MOD6 for transient thermal-hydraulic analysis of LWR including effects of BWR core spray

    International Nuclear Information System (INIS)

    The RELAP4/MOD6/U4/J3 code is the latest version of RELAP4/MOD6/Update4 improved in JAERI. The major improvements and modifications included in this version have been carried out aiming at small break LOCA analysis and BWR-LOCA analysis after core spray initiation. For example, a CCFL calculation model and a spray heat transfer model have been added for BWR-LOCA analysis. Using these models, through calculation from the beginning of blowdown to the end of reflood in BWR-LOCA was made practicable. Furthermore, the analyses of operational transients of LWR were facilitated greatly by an addition of a trip reset function. In this report, the description of the improvements and modifications included in this version, the input data description, and the results of two sample problems are contained. (author)

  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)

    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)

  17. LOCA features peculiar to an integral water cooled PWR

    International Nuclear Information System (INIS)

    LOCA initiated by a guillotine break of the pressurizer surge line has been considered in the paper. The failure of two emergency core cooling system (ECCS) trains was also postulated, that turns the considered accident sequence into a beyond the design basis (BDB) class. Basic design characteristics of the ABV reactor and the containment system are presented as well as the factors of much importance to the accident progression. SCDAP/RELAP5/MOD3.1 was used as the computer code for the simulation of reactor and containment system behavior in the course of the accident. Since a noncondensable driven pressurizing system was employed in the reactor design, the presence of dissolved nitrogen in the primary water was taken into account in calculations. The important feature of the simulated accident is the primary system refilling with the water of pressure suppression pool driven by the pressure difference between containment system compartments. (author)

  18. LOCA and RIA studies at JAERI

    International Nuclear Information System (INIS)

    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)

  19. Analysis on the Direct Vessel Injection Line Break Accident at APR+ Standard Design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngho; Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Kim, Kap [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    APR+ (Advanced Power Reactor +) is the newest design variation of APR1400. The main characteristics of APR+, compared with APR1400, are passive safety systems and dedicated systems for severe accident mitigation. APR+ is under review for standard design certification. In this study, thermal hydraulic analysis on the Direct Vessel Injection (DVI) line break accident postulated in APR+ design was performed. Comparisons of the major parameters which can represent the overall accident behavior during DVI line break accident, several discrepancies between this study and reference data were found and such discrepancies include actuation timing of SIPs and SITs, and also include parameter behaviors of break flow rate and PCT at the accident initiation. These differences were mainly from the different thermal hydraulic models in simulation codes. The behavioral differences for break flow as well as peak cladding temperatures will be examined further as a next step for this study.

  20. Analysis on a Nambu--Jona-Lasinio Model of Dynamical Supersymmetry Breaking

    CERN Document Server

    Cheng, Yifan; Faisel, Gaber; Kong, Otto C W

    2016-01-01

    This is a report on our newly proposed model of dynamical supersymmetry breaking with some details of the analysis involved. The model in the simplest version has only a chiral superfield (multiplet), with a strong four-superfield interaction in the K\\"ahler potential that induces a real two-superfield composite with vacuum condensate. The latter has supersymmetry breaking parts, which we show to bear nontrivial solution following basically a standard nonperturbative analysis for a Nambu--Jona-Lasinio type model on a superfield setting. The real composite superfield has a spin one component but is otherwise quite unconventional. We discuss also the parallel analysis for the effective theory with the composite. Plausible vacuum solutions are illustrated and analyzed. The supersymmetry breaking solutions have generated soft mass(es) for the scalar avoiding the vanishing supertrace condition for the squared-masses of the superfield components. We also present some analysis of the resulted low energy effective th...

  1. Economics of Garlic Production in Baran District of Rajasthan; Break Even Analysis

    OpenAIRE

    Lokesh Kumar Meena; Chandra Sen; Arun jhajharia; N. K. Raghuwanshi

    2013-01-01

    The study focuses on economic analysis of garlic production in the Baran District of Rajasthan. The study is carried out to determine break even analysis and constraints of garlic production in the study area. Break even analysis is carried out to arrive at that minimum level at which optimum conditions of cost and returns is equated that is no profit no loss point. In this study selected small, medium and large farmers will not be at loss even if their actual yield of garlic is decline by 56...

  2. Analysis of primary pipe break for Korea Advanced Liquid Metal Reactor

    International Nuclear Information System (INIS)

    A postulated break in the primary pump discharge pipe is analyzed to assure the inherent safety of Korea Advanced Liquid Metal Reactor, a pool-type liquid metal-cooled reactor generating the 392 MWth of power in the core. The main concern of the analysis is the amount of increase in the fuel and the coolant temperatures. The stabilization of the transient due to reactivity feedback is also important. In the present analysis it is assumed that one of the four pipes connecting the pump discharge to core inlet plenum is broken. The break is located at 3.7m below the pump outlet and the diameter of the break is 0.4 m. It is also assumed that the reactor is not scrammed after the initiation of the break, therefore, the pumps keep on running during the accident. The analysis is performed with SSC-K code, which is developed for the analysis of the transient system response of a pool-type reactor. As soon as the break occurs, the core flow decreases drastically to 65 % full flow in the base case. A more conservative case is also analyzed in which the core flow is reduced artificially to 50 % full flow. The reactor power stabilizes by the reactivity feedback effects in about 10 minutes. The increase of the fuel and coolant temperatures due to the sudden reduction of the core flow are also mitigated with a large margin to coolant saturation temperature. The gas expansion module plays an important role to provide the dominant reactivity feedback when the core flow is reduced less than 50% full power. It is convinced from these results that both a sufficient subcooling margin more than 400 K and a stable system response are maintained in the KALIMER design during the primary pipe break accident, which guarantees the inherent safety of KALIMER against a pipe break. (author)

  3. Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

    International Nuclear Information System (INIS)

    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)

  4. An analysis of oil production by OPEC countries: Persistence, breaks, and outliers

    International Nuclear Information System (INIS)

    This study examines the time series behaviour of oil production for OPEC member countries within a fractional integration modelling framework recognizing the potential for structural breaks and outliers. The analysis is undertaken using monthly data from January 1973 to October 2008 for 13 OPEC member countries. The results indicate there is mean reverting persistence in oil production with breaks identified in 10 out of the 13 countries examined. Thus, shocks affecting the structure of OPEC oil production will have persistent effects in the long run for all countries, and in some cases the effects are expected to be permanent. - Research Highlights: →Mean reverting persistence in oil production with breaks identified in 10 out of the 13 countries examined. → Standard analysis based on cointegration techniques and involving oil production should be examined in the more general context of fractional cointegraton. → Analysis of outliers did not alter the main conclusions of the study.

  5. Validation of the thermal hydraulic computer code S-RELAP5 for performing loss-of-coolant accident analysis (LOCA) in Pressurized Water Reactors (PWRs)

    International Nuclear Information System (INIS)

    Siemens Power Corporation (SPC) has developed S-RELAP5, a RELAP5/MOD2 based thermal hydraulic system code with main modifications and improvements relative to RELAP5/MOD2 concerning Multi-Dimensional Capability, Energy Equations, Numerical Solution of Hydrodynamic, Constitutive Models, Heat Transfer Models, Chocked Flow, and Counter-Current Flow Limiting. S-RELAP5 was exercised over a range of integral and separate effects tests in order to demonstrate that the code could predict the important phenomena associated with PWR LBLOCA. A methodology for calculation of statistical uncertainties has been developed and applied to analyses of hypothetical large break loss-of-coolant accidents (LBLOCA). To extend the application capability of S-RELAP5 to small break loss-of-coolant accidents problems (SBLOCA) an investigation program for appropriate experiments was launched and largely carried out. (author)

  6. Experiments on the high temperature graphite and steam reaction under LOCA condition

    International Nuclear Information System (INIS)

    The results of conceptual design activity of ITER (International Thermonuclear Experimental Reactor) gave graphite as a primarily candidate for the plasma facing materials in the first physics phase. In safety analysis of the fusion experimental reactor, release of radioactive tritium and activation product in the plasma-facing materials caused by a loss of coolant accident (LOCA) in a vacuum vessel is one of the severest accident scenarios. It is presumed that loss of coolant flow will generate an extremely high temperature spot, above 1000 degrees C, on plasma-facing components such as armor tiles of the first wall and divertor plates. Thus coolant pipe made of copper will break or melt, then coolant water will blow out as steam into the vacuum vessel. High temperature graphite will react violently with the steam pouring into the vacuum vessel. Then safety analysis necessitate that the reactivity of the graphite materials should be precisely evaluated. To obtain fundamental data for safety analyses considering loss of coolant inside of the vessel, the rate of reaction between high temperature graphite and steam has been experimentally measured between 1000 and 1600 degrees C. For experiments isotopic graphite and C/C composite ones were used. Reaction rate to a unit surface area was measured dividing the weight loss by inner surface area of tube type graphite specimen, and product gases were analyzed by gas-chromatography

  7. AREVA LOCA and non-LOCA realistic methodology development strategy based on CATHARE

    International Nuclear Information System (INIS)

    The CATHARE code developed since 1979 by AREVA, Cea, EDF and IRSN is one of the major thermal-hydraulic system codes worldwide. The paper gives an overview of CATHARE 2 Version 2.5 based realistic methodologies elaborated by AREVA for LOCA and non-LOCA and the underlying process (called DRM) applied for that purpose, the special features and improvements implemented in the code to handle additional needs and possible future requirements for industrial applications such as the effect of high Burn-up on fuel and cladding behaviour during LOCAs, coupling with core thermal-hydraulics, 3-dimensional core physics and instrumentation and control, capability to account for asymmetric reactor coolant system flow transients by means of dedicated vessel mixing matrices, second order numerical resolution scheme for boron front propagation for non-LOCA transients. (Author)

  8. A methodology for the estimation of release of fission products during LOCA with loss of ECCS

    International Nuclear Information System (INIS)

    A Loss of Coolant Accident (LOCA) in a nuclear reactor along with the failure of the Emergency Core Cooling System can cause sustained voiding of the core. In such a situation the core experiences very low flow which leads to poor heat removal from the reactor core. The heat to be removed from the core includes stored heat, heat generated due to metal water reaction at high temperatures, decay heat etc. The poor heat removal leads to heating of the fuel pins to high temperatures. The heating of fuel pins is further enhanced due to metal-water reaction at high temperatures. These high temperatures of the fuel pins can lead to fission product release, which is transported into the Primary Heat Transport (PHT) system and can enter the containment through the break. Analysis is involved due to the complexity of the system and the phenomena to be simulated. In this paper a multistage analysis methodology is presented that involves the development and application of a number of computer programs to model the various phenomena involved. The computer code PHTACT computes the activity release from the fuel as a function of fuel temperatures and cladding oxidation, its distribution into the PHT system and release into the containment. Computation of thermal hydraulic parameters during LOCA is done using the thermal hydraulic analysis code RELAP5. The detailed simulation of fuel pin temperatures is done using computer code HT/MOD4. The convective boundary conditions required for the code are obtained from RELAP5. Creep deformation is considered in the computation of dimensional changes of the coolant channel and estimation of flow blockage due to clad ballooning. The progression of various reaction layers due to high temperature reaction between fuel and clad and clad and steam is also computed, which affects the structural strength of the clad. Different approaches are possible and analysis can be carried out in different phases depending upon the complexities to be

  9. Analysis Of Feedwater Line Break Of APR1400 By MARS Code

    International Nuclear Information System (INIS)

    This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

  10. SAFER03 qualification against ROSA-III recirculation line break spectrum tests

    International Nuclear Information System (INIS)

    Analyses of the ROSA-III recirculation line break spectrum tests were performed by using the SAFER03 computer code to verify the predictive capability of the code for a BWR LOCA. The SAFER03 computer code is the revised version of SAFER02 and has more realistic models for BWR LOCA phenomena. From these analyses and the comparisons with test data, key parameters which are important to predict major behavior during BWR large and small LOCA transients have been clarified and the SAFER03 code has been assessed against recirculation line break test data obtained from ROSA-III facility

  11. 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)

    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

  12. Analysis of the small break loss of coolant accident in the VVER-1000/V446 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Altaha, S. Mahmoud; Mansouri, Masoud; Jahanfarnia, Gholamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2015-12-15

    In this paper, the analysis of a Small Break Loss of Coolant Accident (SBLOCA) in the VVER-1000/V446 nuclear power plant is presented. For a conservative analysis of the accident, the loss of power to the NPP and failure of one accumulator, and also of two emergency core cooling systems (ECCS) in loops 2 and 3 of the primary and secondary circuits are considered when SBLOCA has occurred. The RELAP5/MOD3.2 computer code has been used in performing the analyses. Two cases of accident scenarios as 25 mm and 100 mm breaks are analyzed. The results are in good agreement with those reported in the plant's FSAR. The results of liquid velocity show that in both cases, the flow of hot legs after the break is reversed, which provides the potential for reflux condensation phenomena. Furthermore, in the 25 mm break, the flow rate in the broken and intact side downcomer remains in the downward motion while in the 100 mm break, the broken and intact side flow rate changes to the reversed state alternatively.

  13. Analysis of the small break loss of coolant accident in the VVER-1000/V446 reactor

    International Nuclear Information System (INIS)

    In this paper, the analysis of a Small Break Loss of Coolant Accident (SBLOCA) in the VVER-1000/V446 nuclear power plant is presented. For a conservative analysis of the accident, the loss of power to the NPP and failure of one accumulator, and also of two emergency core cooling systems (ECCS) in loops 2 and 3 of the primary and secondary circuits are considered when SBLOCA has occurred. The RELAP5/MOD3.2 computer code has been used in performing the analyses. Two cases of accident scenarios as 25 mm and 100 mm breaks are analyzed. The results are in good agreement with those reported in the plant's FSAR. The results of liquid velocity show that in both cases, the flow of hot legs after the break is reversed, which provides the potential for reflux condensation phenomena. Furthermore, in the 25 mm break, the flow rate in the broken and intact side downcomer remains in the downward motion while in the 100 mm break, the broken and intact side flow rate changes to the reversed state alternatively.

  14. Break and trend analysis of EUMETSAT Climate Data Records

    Science.gov (United States)

    Doutriaux-Boucher, Marie; Zeder, Joel; Lattanzio, Alessio; Khlystova, Iryna; Graw, Kathrin

    2016-04-01

    EUMETSAT reprocessed imagery acquired by the Spinning Enhanced Visible and Infrared Imager (SEVIRI) on board Meteosat 8-9. The data covers the period from 2004 to 2012. Climate Data Records (CDRs) of atmospheric parameters such as Atmospheric Motion Vectors (AMV) as well as Clear and All Sky Radiances (CSR and ASR) have been generated. Such CDRs are mainly ingested by ECMWF to produce a reanalysis data. In addition, EUMETSAT produced a long CDR (1982-2004) of land surface albedo exploiting imagery acquired by the Meteosat Visible and Infrared Imager (MVIRI) on board Meteosat 2-7. Such CDR is key information in climate analysis and climate models. Extensive validation has been performed for the surface albedo record and a first validation of the winds and clear sky radiances have been done. All validation results demonstrated that the time series of all parameter appear homogeneous at first sight. Statistical science offers a variety of analyses methods that have been applied to further analyse the homogeneity of the CDRs. Many breakpoint analysis techniques depend on the comparison of two time series which incorporates the issue that both may have breakpoints. This paper will present a quantitative and statistical analysis of eventual breakpoints found in the MVIRI and SEVIRI CDRs that includes attribution of breakpoints to changes of instruments and other events in the data series compared. The value of different methods applied will be discussed with suggestions how to further develop this type of analysis for quality evaluation of CDRs.

  15. REWET, PWR LOCA accident experiments

    International Nuclear Information System (INIS)

    1 - Description of test facility: The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator. All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor. The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm. The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate. 2 - Description of test: Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal

  16. Energy analysis and break-even distance for transportation for biofuels in comparison to fossil fuels

    Science.gov (United States)

    In the present analysis various forms fuel from biomass and fossil sources, their mass and energy densities, and their break-even transportation distances to transport them effectively were analyzed. This study gives an insight on how many times more energy spent on transporting the fuels to differe...

  17. Modelling for great breaks accident analysis in the primary system of Angra 1 reactor

    International Nuclear Information System (INIS)

    An analysis is made for a break in the cold leg, of the guillotine type with discharge coefficient C sub(D)=1.0, for the Angra 1 reactor. The computer codes, geometrical models and options used are described. A comparison between the method used and the requirements in the Appendix K of 10 CRF 50 is done. (Author)

  18. Statistical models for the analysis of water distribution system pipe break data

    International Nuclear Information System (INIS)

    The deterioration of pipes leading to pipe breaks and leaks in urban water distribution systems is of concern to water utilities throughout the world. Pipe breaks and leaks may result in reduction in the water-carrying capacity of the pipes and contamination of water in the distribution systems. Water utilities incur large expenses in the replacement and rehabilitation of water mains, making it critical to evaluate the current and future condition of the system for maintenance decision-making. This paper compares different statistical regression models proposed in the literature for estimating the reliability of pipes in a water distribution system on the basis of short time histories. The goals of these models are to estimate the likelihood of pipe breaks in the future and determine the parameters that most affect the likelihood of pipe breaks. The data set used for the analysis comes from a major US city, and these data include approximately 85,000 pipe segments with nearly 2500 breaks from 2000 through 2005. The results show that the set of statistical models previously proposed for this problem do not provide good estimates with the test data set. However, logistic generalized linear models do provide good estimates of pipe reliability and can be useful for water utilities in planning pipe inspection and maintenance

  19. A study of 2-Dimensional effects in the core of a PWR during the refloading phase of a LOCA. Analysis of data of PERICLES experiments with the COBRA-NC code

    International Nuclear Information System (INIS)

    The project is embedded in the Shared Cost Action Programme (SCA) of the European Communities (CEC) on Reactor Safety, Research Area No. 4, concerning the analysis of experimental data on loss-of-coolant accidents and emergency core cooling. The PERICLES experiments, performed at CEA in Grenoble, had the objective to study multidimensional effects under well defined conditions concentrating on the inter-assembly character of reflood phenomena. The general aim of the present project is to analyse PERICLES experimental data in order to improve models in relevant system codes. Particular objectives of the project are - the critical evaluation of the experimental data of PERICLES Run 8; - the drawing of conclusions from the data with respect to physical and geometrical models for the multi-bundle reflood analysis; - the performance of one-and multi-dimensional computations with COBRA-NC; - the comparison of computational and experimental data; and - the development of conclusions and specifications of additional research needed. The analysis of the experimetal data of Run 8 was performed by a computer programme developed for postprocessing data of any PERICLES experiment. The postprocessor includes an automatic location of the quenchfront and displays it graphically with respect to time, vertical and horizontal directions. Furthermore, rod and fluid temperatures versus height, quenchtimes versus height, densities versus height, and temperatures, pressures, densities etc. versus time can be plotted. As far as computer simulations are concerned, it was one of the objectives of the present study to analyse in greater detail the multidimensional phenomena during the reflooding phase of a LOCA and to compare the numerical results with the experimental data. Such simulation may serve to adjust and improve existing computer codes as well as to validate the codes. Moreover, computer simulations are able to give information which are not available from experimental data; in the

  20. The effect of internals vent valves on reflood following a hypothetical PWR LOCA

    International Nuclear Information System (INIS)

    This paper presents an analysis of the effect of internals vent valves in alleviating the potential for core steam binding and reducing the conventional loss coefficient for the venting pipework during reflood following a hypothetical PWE LOCA. The RAP code was used to construct response surfaces for the time to quench at six-foot elevation for systems with and without the valves. (author)

  1. Break-Even Income Analysis of Pharmacy Graduates Compared to High School and College Graduates.

    Science.gov (United States)

    Chisholm-Burns, Marie A; Gatwood, Justin; Spivey, Christina A; Dickey, Susan E

    2016-04-25

    Objective. To project the net cumulative income break-even point between practicing pharmacists and those who enter the workforce directly after high school graduation or after obtaining a bachelor's degree. Methods. Markov modeling and break-even analysis were conducted. Estimated costs of education were used in calculating net early career earnings of high school graduates, bachelor's degree holders, pharmacists without residency training, and pharmacists with residency training. Results. Models indicate that over the first 10 years of a pharmacist's career, they accumulate net earnings of $716 345 to $1 064 840, depending on cost of obtaining the PharmD degree and career path followed. In the break-even analysis, all pharmacy career tracks surpassed net cumulative earnings of high school graduates by age 33 and bachelor's degree holders by age 34. Conclusion. Regardless of the chosen pharmacy career track and the typical cost of obtaining a PharmD degree, the model under study assumptions demonstrates that pharmacy education has a positive financial return on investment, with a projected break-even point of less than 10 years upon career entry. PMID:27170815

  2. Break-Even Income Analysis of Pharmacy Graduates Compared to High School and College Graduates

    Science.gov (United States)

    Gatwood, Justin; Spivey, Christina A.; Dickey, Susan E.

    2016-01-01

    Objective. To project the net cumulative income break-even point between practicing pharmacists and those who enter the workforce directly after high school graduation or after obtaining a bachelor’s degree. Methods. Markov modeling and break-even analysis were conducted. Estimated costs of education were used in calculating net early career earnings of high school graduates, bachelor’s degree holders, pharmacists without residency training, and pharmacists with residency training. Results. Models indicate that over the first 10 years of a pharmacist’s career, they accumulate net earnings of $716 345 to $1 064 840, depending on cost of obtaining the PharmD degree and career path followed. In the break-even analysis, all pharmacy career tracks surpassed net cumulative earnings of high school graduates by age 33 and bachelor’s degree holders by age 34. Conclusion. Regardless of the chosen pharmacy career track and the typical cost of obtaining a PharmD degree, the model under study assumptions demonstrates that pharmacy education has a positive financial return on investment, with a projected break-even point of less than 10 years upon career entry.

  3. Blind-blind prediction by RELAP5/MOD1 for a 0.1% very small cold-leg break experiment at ROSA-IV large-scale test facility

    International Nuclear Information System (INIS)

    The large-scale test facility (LSTF) of the Rig of Safety Assessment No. 4 (ROSA-IV) program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (LOCAs) and operational transients. The 0.1% very small cold-leg break experiment was conducted as the first integral experiment at the LSTF. The test provided a good opportunity to truly assess the state-of-the-art predictability of the safety analysis code RELAP5/MODI CY18 through a blind-blind prediction of the experiment since there was no prior experience in analyzing the experimental data with the code; furthermore, detailed operational characteristics of LSTF were not yet known. The LOCA transient was mitigated by high-pressure charging pump injection to the primary system and bleed and feed operation of the secondary system. The simulated reactor system was safely placed in hot standby condition by engineered safety features similar to those on a PWR. Natural circulation flow was established to effectively remove the decay heat generated in the core. No cladding surface temperature excursion was observed. The RELAP5 code showed good capability to predict thermal-hydraulic phenomena during the very small break LOCA transient. Although all the information needed for the analysis by the RELAP5 code was obtained solely from the engineering drawings for fabrication and the operational specifications, the code predicted key phenomena satisfactorily

  4. Fuel safety analysis following feeder break accident for refurbished Wolsong 1

    International Nuclear Information System (INIS)

    The objective of the fuel analysis for the postulated accident was to estimate the quantity and timing of a fission product release from fuels when a postulated single channel accident occurs in CANDU 6 reactors. In this study, a fuel safety analysis for the refurbished Wolsong 1 was carried out by using the latest IST (Industrial Standard Toolset) fuel code. The relevant accident scenario focused in this study was a feeder stagnation break accident. The amount of fission product inventory and its distribution during the normal operating conditions were calculated by using the latest ELESTRES-IST code. For a calculation of transient fission product release following the feeder stagnation break, it was assumed that all fuel sheaths in the channel were failed and the entire gap inventory was released instantaneously at the beginning of the accident. The additional releases from the grain boundary and in-grain bound inventories were estimated by applying the Gehl's release model. (author)

  5. Analysis of breaks in a non-equally load sharing system

    Czech Academy of Sciences Publication Activity Database

    Linka, A.; Volf, Petr; Tunák, M.

    Bordeaux : University Victor Segalen, 2008, s. 99-101. ISSN N. [ALT'2008-- 2nd International Conference on Accelerated Life Testing. Bordeaux (FR), 09.06.2008-11.06.2008] R&D Projects: GA MŠk(CZ) 1M06047 Institutional research plan: CEZ:AV0Z10750506 Keywords : reliability analysis * load sharing system * breaking strength Subject RIV: BB - Applied Statistics, Operational Research

  6. New Life Styles, New Barriers to Break Down : Rainbow Family Tourism, Service Description and Analysis

    OpenAIRE

    Algueró Durany, Mariona

    2012-01-01

    This thesis titled 'New life styles, new barriers to break down. Rainbow family tourism; service description and analysis', was written to offer the opportunity to get to know one of the latest services in the tourism industry, the rainbow family tourism or lesbian, gay, transgender and bisexual family tourism (LGTB family tourism). As the title says, the service has been described in detail, starting from basic concepts like tourism and family and continuing with the predecessor the LGTB tou...

  7. Nanofluid application in post SB-LOCA transient in VVER-1000 NPP

    International Nuclear Information System (INIS)

    This paper is the third in a series of four papers that the application of nanofluid as a coolant to improve heat transfer in a VVER-1000 nuclear reactor is investigated. In the first and second papers, neutronics and thermo-hydraulic behavior of nano-particles in normal operation mode of the reactor were reported. In this study, the effects of nanofluids in Small Break Loss of Coolant Accident (SB-LOCA) to complement the existing safety systems is investigated. During SB-LOCA transient, due to reduced mass flow rate and Reynolds number, flow boiling along with vapor formation around the fuel rods occur. The fuel assembly coolant channel of a VVER-1000 core is modeled using a CFD code and heat transfer coefficients, pressure drop and volume fraction distribution of phases are computed for water/Al2O3 nanofluid. We observe that with the escalation of heat transfer enhancement, due to reduction in void fraction, pressure drop along the channel is reduced. This new phenomenon of lower pressure drop along with heat transfer enhancement, would be a significant factor for the use of nanofluids during reactor SB-LOCA transients

  8. BEACON/MOD2A analysis of the Arkansas-1 reactor cavity during a hypothetical hot leg break

    International Nuclear Information System (INIS)

    As part of the evaluation of the new MOD2A version of the BEACON code, the Arkansas-1 reactor cavity was modeled during a hypothetical loss-of-coolant accident. Results of the BEACON analysis were compared with results obtained previously with the COMPARE containment code. Studies were also made investigating some of the BEACON interphasic, timestep control, and wall heat transfer options to assure that these models were working properly and to observe their effects on the results. Descriptions of the Arkansas-1 reactor cavity, initial assumptions during the hypothetical LOCA, and methods of modeling with BEACON are presented. Some of the problems encountered in accurately modeling the penetrations surrounding the hot and cold leg pipes are also discussed

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

    International Nuclear Information System (INIS)

    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)

  10. TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

    International Nuclear Information System (INIS)

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations

  11. High energy pipe break analysis for a Main Steam Line of a WWER NPP

    Energy Technology Data Exchange (ETDEWEB)

    Zaccari, N., E-mail: nicola.zaccari@enel.com; Caria, S., E-mail: sara.caria@enel.com; Rubeo, G., E-mail: giampiero.rubeo@enel.com

    2014-01-15

    Highlights: • Detailed methodological approach on the High Energy Pipe Break (HEPB) analysis for NPP. • HEPB level 3 analysis on the Main Steam Line of MO3 NPP was provided. • Reducing the construction or modification cost on the NPP due to a less conservative approach as a level 1 or 2 analysis. • Innovative LBB approach to treat the through-wall crack on the piping system. -- Abstract: The aim of this study is to show the application of the High Energy Pipe Break (HEPB) methodology to a level 3 analysis, applied to the study of the Main Steam Line (MSL) of a Mochovce (MO) NPP. The analysis highlighted the significant advantage, as compared to a standard level 1 analysis, in the minimization of extra supports and restraints. In the first part of the paper an introduction on the NUREG Standard Review Plan (STRP) and on the ANSI/ANS 58.2 used as a reference in this kind of analysis is provided. In the second part a description of the analyses (coupled mechanical and fluido-dynamic) used to analyze the HEPB is described. Finally an original approach developed at ENEL is presented for an improved evaluation of the leakage cracking by following a non standard procedure related to the LBB approach, coupled with the above-mentioned HEPB.

  12. Experiment on Density Gradient Driven Flow in Small Break Air Ingress Accident of VHTRs

    International Nuclear Information System (INIS)

    This study measures amount of air-ingress rates through a small hole in a circular pipe for various break conditions. The main parameters considered are break orientation, break size, main flow velocity, and density ratio. The main objectives are summarized below: □ Understanding on fundamental air-ingress phenomena in the small break accident □ Development of flow regime map for the small break air-ingress □ Development of air-ingress model for VHTR safety analysis code. A Very High Temperature Reactor (VHTR) is one of the six Gen-IV reactor concepts which is adapting carbon layered TRISO-fuel, graphite-moderator, and helium-coolant. In spite of its inherent safety concept, the VHTR could be detrimental if a LOCA type accident occurs, which is followed by a pipe break. After the break, the air in the cavity starts to ingress into the reactor by either local density-gradient driven flow or molecular diffusion. The main concern of this accident is that it could eventually lead to structural degradation or release of the toxic and explosive gasses (CO) by oxidation of graphite. Previously, majority of the air-ingress studies have been focused on the large size break accident, which is called a double-ended-guillotine-break (DEGB). However, in this study, more focus in put on the small break (or leakage) accident, which is more realistic and probable in the VHTRs. According to the previous studies, the phenomena in the small break accident appear to be much more complicated than those in the DEGB, but little studies have been conducted and reported so far

  13. Heat transfer effectiveness for cooling of Canadian SCWR fuel assembly under the LOCA/LOECC scenario

    International Nuclear Information System (INIS)

    Highlights: • A two-dimensional heat conduction model is developed for SCTRAN code. • A radiation heat-transfer model has been incorporated into the SCTRAN code. • Cladding temperature of fuel rods stays below melting point under LOCA/LOECC. • The heat transfer effectiveness of Canadian SCWR under LOCA/LOECC is demonstrated. • Sensitivity analyses on the predicted MCST have been performed. - Abstract: The effectiveness of heat transfer from the fuel rods to the moderator in a Canadian, pressure-tube type, Super-Critical Water-cooled Reactor (SCWR) has been assessed for the postulated loss-of-coolant accident (LOCA) coupled with the total loss of emergency core cooling (LOECC) event using the safety analysis code “SCTRAN”. A radiation heat-transfer model, including a two-dimensional heat conduction solution scheme, has been incorporated into the SCTRAN code for this assessment. The assessment result shows that the combined radiation heat transfer from the fuel rods to the fuel channel and natural convective heat transfer from the fuel rods to the steam are effective in removing decay heat for the LOCA/LOECC scenario. Maximum cladding temperatures of fuel rods in inner and outer rings of the fuel assembly in the highest-power channel are predicted at 1278 °C and 1192 °C respectively, which are lower than the melting point of the modified stainless-steel cladding. Sensitivity analyses of several analytical parameters on the predicted maximum cladding temperatures have been performed

  14. TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.

  15. Structural analysis of steam generator internals following feed water main steam line break: DLF approach

    International Nuclear Information System (INIS)

    In order to evaluate the possible release of radioactivity in extreme events, some postulated accidents are analysed and studied during the design stage of Steam Generator (SG). Among the various accidents postulated, the most important are Feed Water Line Break (FWLB) and Main Steam Line Break (MSLB). This report concerns with dynamic structural analysis of SG internals following FWLB/MSLB. The pressure/drag-force time histories considered were corresponding to the conditions leading to the accident of maximum potential. The SG internals were analysed using two approaches of structural dynamics. In first approach simplified DLF method was adopted. This method yields an upper bound values of stresses and deflection. In the second approach time history analysis by Mode Superposition Technique was adopted. This approach gives more realistic results. The structure was qualified as per ASME B and PV Code SecIII NB. It was concluded that in all the components except perforated flow distribution plate, the stress values based on elastic analysis are within the limits specified by ASME Code. In case of perforated flow distribution plate during the MSLB transient the stress values based on elastic analysis are higher than the ASME Code limits. Therefore, its limit load analysis had to be done. Finally, the collapse pressure evaluated using limit load analysis was shown to be within the limits of ASME B and PV Code SecIII Nb. (author). 31 refs., 94 figs., 16 tabs

  16. Korean Consortium's preliminary research for enhancing a probabilistic fracture mechanics code, PRO-LOCA

    International Nuclear Information System (INIS)

    The Battelle developed a probabilistic fracture mechanics code called PRO-LOCA, which can be used as a tool for evaluating the pipe break frequency. It is being further developed through the international co-operative research program, PARTRIDGE. KINS, KHNP-CRI, and KEPCO-E&C are participating in the PARTIRDGE program by composing a Korean Consortium. The members of Korean Consortium performed benchmark analyses using the beta version of PRO-LOCA 4.0 to evaluate the effect of variables such as simulation methods, crack features, loading conditions, and inspection models on the failure probabilities. The benchmark analyses showed that the PRO-LOCA can provide a trend consistent with the expected crack growth and pipe failure behavior. Especially, the availability of the stress intensity factor and crack opening displacement for non-idealized through-wall cracks was proven from this study. This new solution for non-idealized through-wall cracks had been developed by the Korean Consortium and it was newly included in PRO-LOCA 4.0. However, further improvement is needed to address the problems such as the instability of adaptive sampling method and the unexpected trend of failure probabilities at the early stage of crack growth

  17. Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

    International Nuclear Information System (INIS)

    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

  18. The Loca Project: Locative media and pervasive surveillance.

    OpenAIRE

    Hemment, Drew; Raento, Mika; Humphries, Theo; Evans, John

    2006-01-01

    A discussion of artwork Loca: Set To Discoverable, awarded an Honorary Mention, Prix Ars Electronica 2008. Loca was an artist-led project on grass-roots, pervasive surveillance using mobile phones. The premier full presentation of at ISEA2006 and ZeroOne in August 2006 combined art installation, software engineering, activism, pervasive design, hardware hacking, SMS poetry, sticker art and ambient performance.

  19. Development of A Conservative Method for A Feedwater Pipe Break Analysis of An Integral Type Reactor

    International Nuclear Information System (INIS)

    The development of advanced small and medium sized nuclear power plants for multipurpose appears before the footlights, and some of them are ready for construction. The SMART, which is an integral pressurized water reactor is one of those advanced types of small sized nuclear reactors. The basic design of SMART was completed at the Korea Atomic Energy Research Institute. A new phase in order to test and verify the SMART design is currently underway in Korea. The results of these tests and verifications will be fed back into the SMART design for a further improvement of the safety and reliability. The integral type reactor can be mitigated design basis events by a reactor protection system, or engineered safety features. The consequences of design basis events must be less than the established acceptance limits and provide an acceptable margin to protect the health and safety. The design basis events are divided into general categories corresponding to their effect on a plant. One of these categories is a decrease in a heat removal by the secondary system. There are a turbine trip, a main steam isolation valve closure, a loss of the primary component cooling system, and a feedwater pipe break for the decrease in the heat removal by the secondary system. The feedwater pipe break accident is one of the most important accidents in the safety of the integral type reactor. Decrease in the feedwater supply to the steam generators causes a decrease in the heat extraction rate from the reactor coolant system, resulting in an increase of a primary coolant temperature and a pressure, and the nuclear power decreases due to a reactivity feedback. Performed sensitivity analysis to find parameters affecting seriously in the integral reactor's feedwater pipe break accident. According to these parametric analysis results, a power level, an initial system pressure, a moderator reactivity coefficient and a break size are major parameters for the maximum system pressure. The detailed

  20. Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

    International Nuclear Information System (INIS)

    This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

  1. Best effort analysis of critical large loss-of-coolant accident in Darlington NGS

    International Nuclear Information System (INIS)

    A best-effort analysis of Emergency Coolant Injection System (ECIS) effectiveness has been performed for a critical large break loss of coolant accident (LOCA) in Darlington NGS. This analysis, and various sensitivity analyses were performed using the best-effort version of the TUF two-fluid thermal-hydraulics code. The objective of this project is to develop analytical tools and analysis methodology to quantify, within reasonable bounds of certainty, the effectiveness of the ECIS in Ontario Hydro nuclear generating stations to limit activity releases from fuel in the event of a large break LOCA. As part of Best Effort ECIS effectiveness methodology, and the pilot application of this methodology to the analysis of Large LOCA for Darlington NGS, the TUF code has been developed to: quantify the degree of blowdown cooling in a multiple parallel channel reactor core; establish the minimum moderator subcooling required to ensure that fuel channel integrity is maintained, and determine the maximum time that the moderator is required to act as a heat sink; quantify the effectiveness of the ECIS to limit the extent of fuel and fuel channel heatup. The methodology described in this paper, together with enhancements to account for the effects of fuel string relocation, higher void reactivity uncertainty allowance and flux tilt on the initial overpower transient, has been implemented in the Generic Safety Report analysis to update the Large LOCA Safety Report sections for the Bruce and Pickering NGS. (author). 9 refs., 12 figs

  2. A kinetic analysis of strand breaks on large DNA induced by cigarette smoke extract

    Science.gov (United States)

    Kurita, Hirofumi; Takata, Tatsuya; Yasuda, Hachiro; Takashima, Kazunori; Mizuno, Akira

    2010-06-01

    We report a kinetic analysis of strand breakages on large DNA molecules induced by cigarette smoke extract (CSE), an extract of soluble cigarette smoke components. Previously, this DNA damage was analyzed by agarose gel electrophoresis, whereas we used fluorescence to kinetically analyze damage to individual DNA molecules. CSE caused a marked change in length of DNA molecules. The rate of CSE-induced double-strand breakage on large random-coiled DNA molecules was determined using a simple theoretical model, allowing the facile estimation of the rate of double-strand breaks on large DNA molecules.

  3. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  4. Phase-shift analysis of pd elastic scattering below break-up threshold

    International Nuclear Information System (INIS)

    A phase-shift analysis was performed for pd elastic scattering based on measurements of differential cross sections and proton and deuteron analyzing powers for energies below the break-up threshold. The angular momenta were restricted to l <= 3; j-splitting and channel-spin mixing of the P-phases and the tensor coupling between the S- and D-phases were taken into account. The phase shifts were parameterized by the effective-range formalism and the corresponding parameters were directly deduced from the data. The results are compared with Faddeev calculations in which the Coulomb interaction is treated exactly or as a two-body approximation. (orig.)

  5. The Breaking Bad Constellation. Analysis of the Newly Found Complementarity between Television and Internet

    Directory of Open Access Journals (Sweden)

    Sarah SEPULCHRE

    2011-01-01

    Full Text Available The hypothesis developed in this paper is that television and Internet are complementary. Both media collaborate in order to propose genuine transmedia narratives. These news adaptations are not identical to movie or novel adaptations, notably because they are simultaneous, interactive and multi-genres. The analysis of Breaking Bad will be presented in the second part of this communication. In the first one, concepts of “remediation” and “convergence”, which constitute the framework of our demonstration, are clarified.

  6. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  7. Evaluation of direct vessel injection design with pressurized thermal shock analysis

    International Nuclear Information System (INIS)

    The purpose of this paper is to evaluate the direct vessel injection design from a pressurized thermal shock(PTS) viewpoint for the Combustion Engineering System 80+. A break of the main steam line from zero power and a 0.05 ft2 small break loss-of-coolant accident(LOCA) from full power were selected as the potential PTS events. In order to investigate the stratification effects in the reactor downcomer region, the fluid mixing analysis was performed using the COMMIX-IB code for steam line break and using the REMIX code for 0.05 ft2 small break LOCA. The stress distributions within the reactor vessel walls experiencing the pressure and the temperature transients were calculated using the OCA-P code for both events. The results of the analysis showed that a small break LOCA without decay heat presented the greatest challenge to the vessel, however, there is no crack initiation through end-of-life of the vessel with consideration of decay heat. (Author)

  8. Fission product transport in the reactor coolant system for a spectrum of interfacing system LOCA scenarios

    International Nuclear Information System (INIS)

    One of the most important potential severe accident sequences for any pressurized water reactor (PWR) is a loss of coolant accident (LOCA), or V-sequence, in one of the interfacing systems. As initially described in the reactor safety study WASH-1400, interfacing system LOCAs involved the failure of check valves in emergency core cooling systems (ECCS), but could also involve the residual heat removal (RHR) systems. The check valves protect the low-pressure portions of these systems from the high pressures of the reactor coolant system (RCS) to which they are connected to provide cold leg injection. A consequent break in the low-pressure piping outside the containment may result in core damage and a direct pathway for fission products to be transported from the core, through the RCS and ECCS or RHR to the auxiliary building, from which they can escape to the environment. This paper addresses the retention and transport of fission products (specifically, CsI) in the RCS in V-sequence scenarios. It summarizes some of the major differences between models resulting from the latest version of the industry degraded core rulemaking (IDCOR) MAAP Computer Program, MAAP 3.0B. Discussed are the differences in: fission product transport and retention in small, medium, and large ECCS pipe breaks, as well as the effect of ECCS and auxiliary feedwater (AFW) system operation and fission product retention in the various regions of the RCS as calculated by MAAP 3.0B and the STCP

  9. Statistical analysis of LOCA, FY 75 report

    International Nuclear Information System (INIS)

    The propagation of probability distributions through a complicated function to form the resulting distribution is considered. Seven techniques are discussed and the Monte Carlo and Response Surface methods are evaluated on two test problems. The results indicate reasonably accurate results can be obtained and that the Response Surface is the most efficient technique

  10. Analysis of integral experiments at ISB-VVER-1000 with the thermohydraulic code ATHLET

    International Nuclear Information System (INIS)

    Post test calculations of LOCA experiments on the integral test facility ISB-VVER with the thermalhydraulic code ATHLET have been implemented. Fulfillment of the work has shown not only capabilities of the code to simulate VVER specifics but revealed peculiarities of the test facility itself. The ISB-VVER test facility models VVER-1000 reactor in the power-volumetric scale l:3000 with full pressure and nominal power. The thermalhydraulic code ATHLET has been developed by the German 'Gesellschaft fuer Anlagen- und Reaktorsicherheit' for PWR LOCA and transients analysis. Three experiments have been analyzed: 2.4% break in the upper plenum, 11% break in the upper plenum, 2.4% break in the downcomer with operation of HPIS. The calculations demonstrated rather good agreement of experimental and numerical data (pressures, temperatures, pressure differences, flow rates, void fractions). (author)

  11. Two phase flow in a PWR vessel downcomer during a refill phase of LOCA

    International Nuclear Information System (INIS)

    The refill stage of the Hypothetical Loss of Coolant Accident (LOCA), of a Pressurized Water Reactor (PWR), has been the subject of considerable analysis and experimental study even through it is a very unlikely event. The Large Break Loss of Coolant Accident has been studied extensively in PWR systems to assess the effectiveness of the emergency core cooling system to maintain the fuel rods within safe temperature level. A particular phase of the transient, known as the Refill phase may be reached when the emergency core coolant, inject via the cold legs could be prevented from completely entering the core due to opposing flow of steam in the annular downcomer of the PWR vessel. Under certain conditions the upward flowing steam can hold-up the liquid and by-pass it around the downcomer annulus to the break. Experiments were performed at close to atmospheric pressure in a polycarbonate 1/10th - scale model of a typical four loop Westinghouse pressurized Water Reactor. The objective of the study was to identify hydraulic conditions and flow regimes that exist during the refill stage of a LOCA in the PWR downcomer, and to study the effects of various liquid inlet conditions (non-uniform liquid injection) into the downcomer on the flooding characteristics. The tests were performed using air and water at the working fluids to simulate the refill process under thermal equilibrium conditions. The effect of hot leg penetrations (blockages) on the flooding characteristics was also investigated and studied. The results were plotted using the Wallis J* parameter, and the mean annular circumference w (was used as the characteristic dimension). The flow regimes types and their spatial distribution in the PWR vessel downcomer for air-water counter-current flow were mapped, presented and discussed. The controlling mechanisms for the flooding were postulated and discussed. The flow pattern maps associated with various liquid modes were constructed. In addition the corresponding

  12. Nurses' perspectives on breaking bad news to patients and their families: a qualitative content analysis.

    Science.gov (United States)

    Abbaszadeh, Abbas; Ehsani, Seyyedeh Roghayeh; Begjani, Jamal; Kaji, Mohammad Akbari; Dopolani, Fatemeh Nemati; Nejati, Amir; Mohammadnejad, Esmaeil

    2014-01-01

    Breaking bad news is quite often not done in an effective manner in clinical settings due to the medical staff lacking the skills necessary for speaking to patients and their families. Bad news is faced with similar reactions on the part of the news receiver in all cultures and nations. The purpose of this study was to explore the perspectives of Iranian nurses on breaking bad news to patients and their families. In this research, a qualitative approach was adopted. In-depth and semi-structured interviews were conducted with 19 nurses who had at least one year work experience in the ward, and content analysis was performed to analyze the data. Five major categories emerged from data analysis, including effective communication with patients and their families, preparing the ground for delivering bad news, minimizing the negativity associated with the disease, passing the duty to physicians, and helping patients and their families make logical treatment decisions. The results of this study show that according to the participants, it is the physicians' duty to give bad news, but nurses play an important role in delivering bad news to patients and their companions and should therefore be trained in clinical and communicative skills to be able to give bad news in an appropriate and effective manner. PMID:25512837

  13. Nurses’ perspectives on breaking bad news to patients and their families: a qualitative content analysis

    Science.gov (United States)

    Abbaszadeh, Abbas; Ehsani, Seyyedeh Roghayeh; begjani, Jamal; Kaji, Mohammad Akbari; Dopolani, Fatemeh Nemati; Nejati, Amir; Mohammadnejad, Esmaeil

    2014-01-01

    Breaking bad news is quite often not done in an effective manner in clinical settings due to the medical staff lacking the skills necessary for speaking to patients and their families. Bad news is faced with similar reactions on the part of the news receiver in all cultures and nations. The purpose of this study was to explore the perspectives of Iranian nurses on breaking bad news to patients and their families. In this research, a qualitative approach was adopted. In-depth and semi-structured interviews were conducted with 19 nurses who had at least one year work experience in the ward, and content analysis was performed to analyze the data. Five major categories emerged from data analysis, including effective communication with patients and their families, preparing the ground for delivering bad news, minimizing the negativity associated with the disease, passing the duty to physicians, and helping patients and their families make logical treatment decisions. The results of this study show that according to the participants, it is the physicians’ duty to give bad news, but nurses play an important role in delivering bad news to patients and their companions and should therefore be trained in clinical and communicative skills to be able to give bad news in an appropriate and effective manner. PMID:25512837

  14. Poisson structure and stability analysis of a coupled system arising from the supersymmetric breaking of Super KdV

    CERN Document Server

    Restuccia, A

    2014-01-01

    The Poisson structure of a coupled system arising from a supersymmetric breaking of N=1 Super KdV equations is obtained. The supersymmetric breaking is implemented by introducing a Clifford algebra instead of a Grassmann algebra. The Poisson structure follows from the Dirac brackets obtained by the constraint analysis of the hamiltonian of the system. The coupled system has multisolitonic solutions. We show that the one soliton solutions are Liapunov stable.

  15. Poisson structure and stability analysis of a coupled system arising from the supersymmetric breaking of Super KdV

    Science.gov (United States)

    Sotomayor, Adrián; Restuccia, Alvaro

    2013-11-01

    The Poisson structure of a coupled system arising from a supersymmetric breaking of N=1 Super KdV equations is obtained. The supersymmetric breaking is implemented by introducing a Clifford algebra instead of a Grassmann algebra. The Poisson structure follows from the Dirac brackets obtained by the constraint analysis of the hamiltonian of the system. The coupled system has multisolitonic solutions. We show that the one soliton solutions are Liapunov stable.

  16. Comparative analysis of several sediment transport formulations applied to dam-break flows over erodible beds

    Science.gov (United States)

    Cea, Luis; Bladé, Ernest; Corestein, Georgina; Fraga, Ignacio; Espinal, Marc; Puertas, Jerónimo

    2014-05-01

    Transitory flows generated by dam failures have a great sediment transport capacity, which induces important morphological changes on the river topography. Several studies have been published regarding the coupling between the sediment transport and hydrodynamic equations in dam-break applications, in order to correctly model their mutual interaction. Most of these models solve the depth-averaged shallow water equations to compute the water depth and velocity. On the other hand, a wide variety of sediment transport formulations have been arbitrarily used to compute the topography evolution. These are based on semi-empirical equations which have been calibrated under stationary and uniform conditions very different from those achieved in dam-break flows. Soares-Frazao et al. (2012) proposed a Benchmark test consisting of a dam-break over a mobile bed, in which several teams of modellers participated using different numerical models, and concluded that the key issue which still needs to be investigated in morphological modelling of dam-break flows is the link between the solid transport and the hydrodynamic variables. This paper presents a comparative analysis of different sediment transport formulations applied to dam-break flows over mobile beds. All the formulations analysed are commonly used in morphological studies in rivers, and include the formulas of Meyer-Peter & Müller (1948), Wong-Parker (2003), Einstein-Brown (1950), van Rijn (1984), Engelund-Hansen (1967), Ackers-White (1973), Yang (1973), and a Meyer-Peter & Müller type formula but with ad-hoc coefficients. The relevance of corrections on the sediment flux direction and magnitude due to the bed slope and the non-equilibrium hypothesis is also analysed. All the formulations have been implemented in the numerical model Iber (Bladé et al. (2014)), which solves the depth-averaged shallow water equations coupled to the Exner equation to evaluate the bed evolution. Two different test cases have been

  17. Performance analysis of single structural break test with an empirical study on efficient market hypothesis"

    OpenAIRE

    Yıldız, İzzet

    2005-01-01

    Cataloged from PDF version of article. In this thesis, performance of the single structural break tests is examined. Since it has proved superiority of Sequential F test on other single break tests, it is chosen as single break test. Monte Carlo simulation is run for different scenarios and performances of the test with respect to estimating break points, and parameters, and rejecting or accepting the joint null hypothesis is observed. For all cases small sample bias is obse...

  18. CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

    International Nuclear Information System (INIS)

    CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

  19. Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

    International Nuclear Information System (INIS)

    Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10−7 in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10−2 in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10−7, compared to those of commercial PWRs ∼10−6–10−1. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of PASS fuel rods during LOCA is

  20. Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2015-10-15

    Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of

  1. Lattice QCD analysis for relation between quark confinement and chiral symmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Doi, Takahiro M.; Suganuma, Hideo [Department of Physics, Graduate School of Science, Kyoto University, Kitashirakawa-oiwake, Sakyo, Kyoto 606-8502 (Japan); Iritani, Takumi [Yukawa Institute for Theoretical Physics, Kyoto University, Kitashirakawa-Oiwake, Sakyo, Kyoto 606-8502 (Japan)

    2016-01-22

    The Polyakov loop and the Dirac modes are connected via a simple analytical relation on the temporally odd-number lattice, where the temporal lattice size is odd with the normal (nontwisted) periodic boundary condition. Using this relation, we investigate the relation between quark confinement and chiral symmetry breaking in QCD. In this paper, we discuss the properties of this analytical relation and numerically investigate each Dirac-mode contribution to the Polyakov loop in both confinement and deconfinement phases at the quenched level. This relation indicates that low-lying Dirac modes have little contribution to the Polyakov loop, and we numerically confirmed this fact. From our analysis, it is suggested that there is no direct one-to-one corresponding between quark confinement and chiral symmetry breaking in QCD. Also, in the confinement phase, we numerically find that there is a new “positive/negative symmetry” in the Dirac-mode matrix elements of link-variable operator which appear in the relation and the Polyakov loop becomes zero because of this symmetry. In the deconfinement phase, this symmetry is broken and the Polyakov loop is non-zero.

  2. Lattice QCD analysis for relation between quark confinement and chiral symmetry breaking

    Science.gov (United States)

    Doi, Takahiro M.; Suganuma, Hideo; Iritani, Takumi

    2016-01-01

    The Polyakov loop and the Dirac modes are connected via a simple analytical relation on the temporally odd-number lattice, where the temporal lattice size is odd with the normal (nontwisted) periodic boundary condition. Using this relation, we investigate the relation between quark confinement and chiral symmetry breaking in QCD. In this paper, we discuss the properties of this analytical relation and numerically investigate each Dirac-mode contribution to the Polyakov loop in both confinement and deconfinement phases at the quenched level. This relation indicates that low-lying Dirac modes have little contribution to the Polyakov loop, and we numerically confirmed this fact. From our analysis, it is suggested that there is no direct one-to-one corresponding between quark confinement and chiral symmetry breaking in QCD. Also, in the confinement phase, we numerically find that there is a new "positive/negative symmetry" in the Dirac-mode matrix elements of link-variable operator which appear in the relation and the Polyakov loop becomes zero because of this symmetry. In the deconfinement phase, this symmetry is broken and the Polyakov loop is non-zero.

  3. Lattice QCD analysis for relation between quark confinement and chiral symmetry breaking

    International Nuclear Information System (INIS)

    The Polyakov loop and the Dirac modes are connected via a simple analytical relation on the temporally odd-number lattice, where the temporal lattice size is odd with the normal (nontwisted) periodic boundary condition. Using this relation, we investigate the relation between quark confinement and chiral symmetry breaking in QCD. In this paper, we discuss the properties of this analytical relation and numerically investigate each Dirac-mode contribution to the Polyakov loop in both confinement and deconfinement phases at the quenched level. This relation indicates that low-lying Dirac modes have little contribution to the Polyakov loop, and we numerically confirmed this fact. From our analysis, it is suggested that there is no direct one-to-one corresponding between quark confinement and chiral symmetry breaking in QCD. Also, in the confinement phase, we numerically find that there is a new “positive/negative symmetry” in the Dirac-mode matrix elements of link-variable operator which appear in the relation and the Polyakov loop becomes zero because of this symmetry. In the deconfinement phase, this symmetry is broken and the Polyakov loop is non-zero

  4. The reactor core behaviour in case of small break loss of coolant accident combined with total blackout

    International Nuclear Information System (INIS)

    After the Fukushima accident an extreme event beyond design basis is shown to be possible. The detailed analyses of an extended station blackout, where all the onsite and offsite power is failed, became very important. A large number of analyses were done in all countries operating nuclear reactors. An analysis of small break loss of coolant accident combined with total blackout is presented in this work. The operator actions in this case are very important in order to extend the time before irreversible damage to the core is done. The analysis is performed using RELAP5/Mod 3.3 for VVER‑1000 type reactor. The main conclusions are that the current emergency operating procedures are adequate to manage station blackout with small break loss of coolant accident (SBLOCA) sequence. Key words: LOCA, Safety Analyses Report, Blackout, Severe Accident

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

    International Nuclear Information System (INIS)

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

  6. Assessment of CONTAIN and MELCOR for performing LOCA and LOVA analyses in ITER

    International Nuclear Information System (INIS)

    This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe accident thermal hydraulic analyses. This code will require the least modification to be appropriate for calculating thermal hydraulic behavior in ITER relevant conditions that include vacuum, cryogenics, ITER temperatures, and the presence of a liquid metal test module. The assessment of the aerosol transport models in these codes concludes that several modifications would have to be made to CONTAIN and/or MELCOR to make them applicable to the aerosol transport part of severe accident analysis in ITER

  7. AEEW comments on the NNC/CEGB LOCA code validation report RX 440-A

    International Nuclear Information System (INIS)

    Comments are made on the NNC/CEGB report PWR/RX 440-A, Review of Validation for the ECCS Evaluation Model Codes, by K.T. Routledge et al, 1982. This set out to review methods and models used in the LOCA safety case for Sizewell B. These methods are embodied in the Evaluation Model Computer codes SATAN-VI, WREFLOOD, WFLASH, LOCTA-IV and COCO. The main application of these codes is the determination of peak clad temperature and overall containment pressure. The comments represent the views of a group which has been involved for a number of years in the development and application of Best-Estimate methods for LOCA analysis. It is the judgement of this group that, overall, the EM methods can be used to make an acceptable safety case, but there are a number of points of detail still to be resolved. (U.K.)

  8. An analysis on boron dilution events during SBLOCA for the KNGR

    International Nuclear Information System (INIS)

    An analysis on boron dilution events during small break loss of coolant accident (LOCA) for Korea Next Generation Reactor (KNGR) was performed using Computational Fluid Dynamic (CFD) computer program FLUENT code. The maximum size of the water slug was determined based on the source of un borated water slug and the possible flow paths. Axisymmetric computational fluid dynamic analysis model is applied for conservative scoping analysis of un borated water slug mixing with recirculation water of the reactor system following small break LOCA assuming one Reactor Coolant Pump (RCP) restart. The computation grid was determined through the sensitivity study on the grid size, which calculates the most conservative results, and the preliminary calculation for boron mixing was performed using the grid. (Author). 17 refs., 3 tabs., 26 figs

  9. Preliminary Safety Analysis of Korea HCSB Test Blanket Module

    International Nuclear Information System (INIS)

    A Helium Cooled Solid Breeder (HCSB) blanket has been considered as one of the promising blanket for the fusion power demonstration plant. Therefore HCSB Test Blanket Module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Korea has developed the HCSB TBM with some features such as graphite reflector and simplified flow passage. The objective of this study was to evaluate the thermal and structural integrity of the HCSB TBM under the hypothetical accidental conditions such as cooling pipe break in TBM. The safety analysis was performed under conservative conditions based on the TBM design, which can be assumed by the similarity of the safety analysis of the ITER shielding blanket. Transient analysis model was used to calculate the temperature distribution for Loss of Coolant Accident (LOCA). Simplified analysis conditions were a) simultaneous plasma shutdown and LOCA b) LOCA and then after FW temperature reaches 1150 deg. plasma shutdown. Helium circuit behavior during the different LOCA scenarios was also evaluated. Finally the design modifications based on the analysis result and the related R-and-D of the HCSB blanket design for the application in a DEMO reactor were mentioned. (author)

  10. The analysis of single-strand breaks in E. coli using a curve-fitting procedure

    International Nuclear Information System (INIS)

    A curve-fitting technique has been used to analyse the activity profiles occurring when the tritium labelled DNA of irradiated E. coli is sedimented in an alkaline sucrose gradient. The fitted curve is a modification of an expression derived elsewhere, which assumes that DNA of unit length undergoes scissions at random positions on the DNA strand. The theoretical curves fitted the data well, even after considerable repair or excision of the radiation damage had been effected by cellular enzymes. The profile resulting from sedimentation of unirradiated DNA could be fitted with the same theoretical expression provided that the velocity of centrifugation was not too high. The number of single-strand breaks derived from the analysis was almost unaffected by the presence of extraneous activity at the top or bottom of the sucrose gradient. The method is therefore more reliable than many other methods used to analyse sedimentation data. (author)

  11. Estimation of the porosity of wind breaks by using GIS-based ortho-image analysis

    Science.gov (United States)

    Mohammadian Behbahani, Ali; Hikel, Harald; Fister, Wolfgang; Heckrath, Goswin; Kuhn, Nikolaus J.

    2013-04-01

    The optimal design of windbreaks is very important to reduce wind erosion on farmlands and to combat soil degradation. Main parameters that must be considered when designing windbreaks are: height, width, orientation, porosity (density), distance between barrier rows, and length. There are two types of windbreaks, living (natural) and non-living (artificial). For tree shelterbelts (living windbreak) some of these parameters are related to inherent characteristics of the plants. For example, the height of a windbreak depends on the type of the plant, its growing conditions and the age of the plant. Porosity of windbreaks is considered to be one of the most important factors that controls wind erosion. It is expressed as the ratio between pore space and the space occupied by tree stems, branches, twigs and leaves. For the assessment of porosity it is necessary to convert the three-dimensional plant structure to a two-dimensional model of its shape or plant silhouette, because a direct measurement in the field is very inefficient, time consuming, and therefore impractical. To solve this issue, different approaches have been introduced to estimate the porosity of wind breaks, e.g. optical or aerodynamic porosity. In this study, the porosity of wind break networks was assessed for agricultural land in north Jutland, Denmark. The objective of this study was to develop a GIS-based Ortho-Image Analysis (OIA) method to estimate the porosity of windbreaks. The images of the windbreaks have three visible (RGB) bands and were taken in autumn 2012. The pixel size of 0.5 m is sufficient to visually distinguish the tree rows from their surrounding background. The identification of trees was done using grayscale images, where the dark trees strongly contrast to the bright sky in the background. The preliminary results indicate that the GIS based Ortho-Image analysis can be used as a quick, accurate, and reliable method to estimate the porosity of wind break networks. It can

  12. Debris transport evaluation during LOCA blow-down using CFD methodology for OPR-1000 plant

    International Nuclear Information System (INIS)

    In response to GSI-191, 'Potential of PWR Sump Blockage Post-LOCA', NEI and the industry formed the PWR Sump Performance Task Force. The primary purpose of Task Force was to creation of a methodology document that could be used as guideline for PWR operators to address the issue. The NEI methodology document provides basic guidance on approach and various methods available. But some additional information be required in order to apply to specific plants, such as OPR-1000, and APR-1400 plant. According to the baseline evaluation of NEI 04-07, debris transport logic chart was composed of 4 transport phases. The present work aim to evaluate debris transport during LOCA blow-down, the first transport phase, based on CFD analysis. The target plant is Ulchin 3 and 4 which is OPR-1000 plant. Flow pattern strongly affects shape of containment, and disposition of components, such as steam generators, RCPs, and pipes, etc. The present work takes advantage of 3D CAD model so that real geometry of OPR-1000 plant is used. The analysis results give a clear figure about flow pattern in containment during LOCA blow-down, and fraction of debris transport to upper containment, which is one of major safety issues. (author)

  13. The substantiation of embrittlement criterion of E110 alloy under LOCA conditions

    International Nuclear Information System (INIS)

    temperatures 800, 900 0C E110 oxidation is significant slowly than Zircaloy-4 oxidation; 2. At temperature 800 0C the oxidation kinetics of E100 and M5 practically have not difference; 3. At temperature 900 0C E110 oxidation is significant slowly than M5 oxidation (may be it is M5 break away oxidation?); 4. At temperatures 800, 900 0C the break away (linear) oxidation of E110 alloy was not observed in contrast to Zircaloy-4 (and M5?); 5. At temperature 1000 0C all named alloys are inclined to break away oxidation but the time of break away effect beginning of E110 alloy is greater than such of M5 and Zircaloy-4 especially; 6. At temperatures 1100, 1200 0C the oxidation kinetics of E110, M5 and Zircaloy-4 practically have not difference; 7. At temperature range from 1000 to 1200 C the dependences of residual ductility from exposure time of E110, M5 and Zircaloy-4 are identical with the exception of Zircaloy-4 embrittlement at 1200 0C after 17 150 seconds. But, at temperatures 1000 and 1100 0C we have not data about M5 and Zircaloy-4 residual ductility after long term oxidation to compare with E110 alloy. The results of investigations of E110 alloy (corrosion tests with continuous record of weigh gain, the results of integral LOCA tests, the researches of irradiated cladding behavior, the results of visual survey of oxidized claddings and their metallographic investigations, results of compression tests) allow to make the next conclusions: 1. Oxide films on surfaces of claddings are black and shining, visual spots and peelings are absent. The oxide films are tightly adhered to the metal and have no cracks. The oxidation kinetics was parabolic, linear oxidation was not observed. But it is one exception, at temperature 1000 0C the break away effect was observed after oxidation during 14000 seconds. 2. All oxidized samples (ECR value more then 18% at temperatures range from 1000 to 1200 0C or exposure time more then 64000 s at temperatures range from 800 to 900 0C) have saved

  14. Virginia Power's generic main steam-line-break DNBR [departure from nucleate boiling ratio] analysis

    International Nuclear Information System (INIS)

    Virginia Power operates four nuclear reactors, two units each at the Surry and North Anna Power stations. The original operating licenses were based on acceptable analysis results of the accidents in the final safety analysis report (FSAR). The assumptions of these analyses must be verified on a reload basis. Included in these FSAR accidents is the main steam-line-break (MSLB) event. The plant FSARs describe the MSLB analyses, which is summarized as follows. The plant is assumed to be at hot zero power at end of life, when the moderator temperature coefficient (MTC) is most negative. The MSLB rapidly cools the secondary side, followed by a primary cooldown in one loop. The surge of cold water into the core, coupled with the negative MTC, results in high local power factors, which in turn can result in a violation of the departure from nucleate boiling ratio (DNBR) limit. The three-dimensional power distribution is calculated at several key state points. These distributions are then subjected to core thermal-hydraulic analysis by the COBRA code. The W-3 correlation is used to calculate the state-point DNBRs. Both the physics and the DNBR calculations have been repeated on a reload basis. As a result, Virginia Power has accumulated a reasonably large data base of MSLB DNBRs for both Surry and North Anna. Virginia Power now uses the power peaking factors criterion to verify that the MSLB analysis remains bounding on a reload basis

  15. Analysis of heavy-ion-induced DNA strand breaks in plasmid pUC18

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Plasmid DNA was irradiated or implanted by mixed particle field(CR) or lithium-ion-beam to detect strand breaks.The primary results showed that mixed particle field could induce single and double strand breaks with positive linear-dose-effects;most of sequence changes induced by CR were point mutant.Lithium-ion-beam could induce strand breaks also,but it was only at dose of 20Gy.

  16. Containment integrity analysis for the Westinghouse Advanced AP600

    International Nuclear Information System (INIS)

    Since 1987, Westinghouse has been performing containment cooling analyses in support of the advanced AP600 plant design. This analysis effort was intended to verify the feasibility of the passive containment cooling system (PCCS) features of the AP600 design, which is being jointly developed by Westinghouse, Burns and Roe Company, and Avondale Industries. To support this goal, the response of the AP600 containment has been analyzed for a large-break loss-of-coolant-accident (LOCA) and for a large steam line break (SLB). These cases were chosen based on the characteristic mass and energy releases each scenario would impose on the containment, with LOCA releases taxing the long-term heat-removal aspects while the SLB release typically defines the limiting short-term response. The transient results indicate that the PCCS design is feasible and capable of removing sufficient heat to limit containment pressure to within acceptable limits

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

    International Nuclear Information System (INIS)

    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

  18. 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)

    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

  19. Measuring the area of tear film break-up by image analysis software

    Science.gov (United States)

    Pena-Verdeal, Hugo; García-Resúa, Carlos; Ramos, Lucía.; Mosquera, Antonio; Yebra-Pimentel, Eva; Giráldez, María. Jesús

    2013-11-01

    Tear film breakup time (BUT) test only examines the first break in the tear film, but subsequent tear film events are not monitored. We present a method of measuring the area of breakup after the appearance of the first breakup by using open source software. Furthermore, the speed of the rupture was determined. 84 subjects participated in the study. 2 μl volume of 2% sodium fluorescein was instilled using a micropipette. The subject was seated behind a slit-lamp using a cobalt blue filter together with a Wratten 12 yellow filter. Then, the tear film was recorded by a camera attached to the slit lamp. 4 frames of each video was extracted, the first rupture (BUT_0), breakup after 1 second (BUT_1), rupture after 2 seconds (BUT_2) and breakup before the last blink (BUT_F). Open source software of measurement based on Java (NIH ImageJ) was used to measure the number of pixels in areas of breakup. These areas were divided by the area of exposed cornea to obtain the percentage of ruptures. Instantaneous breakup speed was calculated for second 1 as the difference between BUT_1 - BUT_0, whereas instant speed for second 2 was BUT_2 - BUT_1. Mean area of breakup obtained was: BUT_0 = 0.26%, BUT_1 = 0.48%, BUT_2 = 0.79% and BUT_F = 1.61%. Break speed was 0.22 area/sec for second 1 and 0.31 area/sec for second 2, showing a statistical difference between them (p = 0.007). Post BUT analysis may be easily monitoring with the aid of this software.

  20. Application of MCNP for predicting power excursion during LOCA in Atucha-2 PHWR

    International Nuclear Information System (INIS)

    Highlights: • Evaluation of moderator physical variables using different level of spatial resolution is relevant for the selected scenario. • Analysis based in high-order method beyond the level actual capability of system codes used for safety analysis. • Prove the feasibility in coupling a Monte Carlo neutron transport code and a computational fluid dynamics code. • Results prove the conservatism of inserted reactivity using the reference system code. - Abstract: Atucha-2 is a Siemens-designed pressurized heavy water reactor in the Republic of Argentina. The correct prediction of the negative reactivity introduced in the moderator by an Emergency Boron Shutdown System (EBSS) is of great relevance for the correct safety evaluation of a double-ended guillotine large break LOCA scenario. During such event the EBSS is in charge to compensate the insertion of positive reactivity, caused by the void generated in the coolant channels by a sharp system pressure drop, in order to avoid severe core damage. The correct simulation of such event implies the minimization of the so called “numeric boron self-shielding effect” or the over-estimation of the inserted negative reactivity caused by the adoption of relatively large numerical meshes. In fact, because during the first phases of the injection, a very high concentrated boron solution is introduced in a small volume of the moderator tank, non-conservative reactivity estimation can be calculated if a “numeric boron dilution” is resulting by the adoption of large thermal-hydraulic and neutronic meshes. A methodology based on Monte Carlo transport code MCNP5 has been developed in order to predict power and reactivity excursions, representing a boron distribution in the moderator with different spatial resolutions. In such a way, it was possible to investigate the negative reactivity over-estimation due to the “boron self-shielding effect”. This investigation is generally not possible by system codes used

  1. Physical data generation methodology for return-to-power steam line break analysis

    International Nuclear Information System (INIS)

    Current methodology to generate physics data for steamline break accident analysis of CE-type nuclear plant such as Yonggwang Unit 3 is valid only if the core reactivity does not reach the criticality after shutdown. Therefore, the methodology requires tremendous amount of net scram worth, specially at the end of the cycle when moderator temperature coefficient is most negative. Therefore, we need a new methodology to obtain reasonably conservation physics data, when the reactor returns to power condition. Current methodology used ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Current methodology uses ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Return-to-power reactivity credit is produced to assist the reactivity table generated by closed channel model. Other data includes hot channel axial power shape, peaking factor and maximum quality for DNBR analysis. It also includes pin census for radiological consequence analysis. 48 figs., 22 tabs., 18 refs. (Author) .new

  2. PBF-LOCA test series test LOC-11 test result report

    International Nuclear Information System (INIS)

    This report presents the results of Loss-of-Coolant (LOC) Test LOC-11, the first test of the Loss-of-Coolant Accident (LOCA) Test Series conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The primary objective of the test was to evaluate the behavior of pressurized water reactor (PWR) fuel under LOCA conditions similar to those postulated during a simulated double-ended cold leg break in a PWR. Test LOC-11 consisted of four, separately shrouded, fresh fuel rods of PWR design, with initial plenum pressure as a variable. Maximum cladding temperatures of up to 10700K (corresponding to high ductility α-phase Zircaloy) were sought during Test LOC-11. The fuel rods were exposed to a series of three blowdowns from different power and coolant conditions. The final blowdown resulted in the maximum measured cladding temperature of 10340K. Upon disassembly of the test train the rods were found to be uniformly covered with a dark grey oxide. Posttest results indicated slight cladding circumferential swelling of the pressurized rods and slight collapse of the relatively unpressurized rods. The results are compared with the posttest analyses to aid in understanding the coolant thermal-hydraulic behavior and fuel rod behavior

  3. A simple analytical scaling method for a scaled-down test facility simulating SB-LOCAs in a passive PWR

    International Nuclear Information System (INIS)

    A Simple analytical scaling method is developed for a scaled-down test facility simulating SB-LOCAs in a passive PWR. The whole scenario of a SB-LOCA is divided into two phases on the basis of the pressure trend ; depressurization phase and pot-boiling phase. The pressure and the core mixture level are selected as the most critical parameters to be preserved between the prototype and the scaled-down model. In each phase the high important phenomena having the influence on the critical parameters are identified and the scaling parameters governing the high important phenomena are generated by the present method. To validate the model used, Marviken CFT and 336 rod bundle experiment are simulated. The models overpredict both the pressure and two phase mixture level, but it shows agreement at least qualitatively with experimental results. In order to validate whether the scaled-down model well represents the important phenomena, we simulate the nondimensional pressure response of a cold-leg 4-inch break transient for AP-600 and the scaled-down model. The results of the present method are in excellent agreement with those of AP-600. It can be concluded that the present method is suitable for scaling the test facility simulating SB-LOCAs in a passive PWR

  4. Breaking Bat

    Science.gov (United States)

    Aguilar, Isaac-Cesar; Kagan, David

    2013-01-01

    The sight of a broken bat in Major League Baseball can produce anything from a humorous dribbler in the infield to a frightening pointed projectile headed for the stands. Bats usually break at the weakest point, typically in the handle. Breaking happens because the wood gets bent beyond the breaking point due to the wave sent down the bat created…

  5. Validation Cases of CATHARE 2 for VVER-1000 Main Steam Line Break Analysis

    Science.gov (United States)

    Kolev, Nikolay P.; Sabotinov, Luben; Petrov, Nikolay; Nikonov, Sergey; Donov, Jordan

    Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plant transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation.

  6. A Sensitivity Analysis of a Pipe Break Accident in a Preliminary Specific Design of the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Jeong, Jae Ho; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is a pool type sodium cooled fast reactor with a thermal power of 392.1 MW which has been developed in accord with an enhanced safety, an efficient utilization of uranium resources and a reduction of a high level waste volume in the Korea Atomic Energy Research Institute (KAERI) since 2012 under a National Nuclear R and D Program. The PGSFR has an inherent safety characteristic owing to the design to have a negative power reactivity coefficient during all operation modes and it has a passive safety characteristic due to the design of a passive decay heat removal circuit. In order to assess the inherent safety features of the PGSFR, a safety analysis was performed for a pipe break accident with MARS-LMR. And, the sensitivity studies were also performed to find the most conservative condition. As a result, the PGSFR was appropriately tripped by a high power to PHTS flow ratio using the method of extracting the PHTS flow rate from the pressure drop. The air flow rate was the most sensitive variable in the sensitivity analysis. Therefore, it is important to know the accurate uncertainty of the air flow rate in the AHX.

  7. Simulation and analysis of a main steam line transient with isolation valves closure and subsequent pipe break

    International Nuclear Information System (INIS)

    Simulation and analysis of a real main steam line break transient at the coal fired 300 MW Drmno Thermal Power Plant have been performed by the computer code TEA-01. The methods and procedures used could be applied to a nuclear power plant. 9 refs., 6 figs

  8. Thermalhydraulic analysis of Candu 6 100% reactor outlet header break using RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Dupleac, D.; Prisecaru, I.; Ghitescu, P. [Power Plant Engineering Faculty, Politehnica University, Bucharest (Romania); Negut, G. [Institute for Nuclear Research, Pitesti (Romania)

    2007-07-01

    One of the postulated large break losses of coolant accident (LBLOCA) in a Candu reactor is the Reactor Outlet Header (ROH) break. After such an event, the normal coolant flow in the channels downstream the break is disrupted and late stagnation occurs after the pump head is degraded and before emergency core cooling (ECC) injection takes place. Given that the fuel decay and stored energy decreased significantly by that time, the heatup of the fuel sheath is much lower than in the case of 35% Reactor Inlet Header (RIH) break. However, the coolant pressure is much lower than the corresponding one at the 35% RIH break.The combination of a high fuel clad temperature and coolant low-pressure lead to more fuel failure events. Thus, the 100% ROH break has the highest potential for radioactivity release. The paper presents the thermal hydraulic analyses of a 100% reactor outlet header break. The study is done with RELAP5/SCDAP mod 3.4 and the results were compared with those of CATHENA. RELAP5 predicts a slightly faster inventory loss, an extended flow stagnation period and a higher clad temperature.

  9. Evaluation of Leak and LOCA Probabilities in RCS piping of Domestic NPPs under Fatigue and Stress Corrosion Cracking Conditions

    International Nuclear Information System (INIS)

    10CFR 50.46 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors' requires the LB-LOCA evaluation in reactor coolant system (RCS) piping. Based on risk informed regulation (RIR) option 3, re-definition of LB-LOCA is discussed in USA. The technical basis of re-definition of LB-LOCA is the extremely low probability of piping rupture in RCS piping. However, the probability of the piping rupture can not be determined from the field data because no rupture event in RCS piping has been reported. However, expert elicitation and probabilistic approach can be used to estimate the rupture frequency of RCS piping. Probabilistic approach can give us relative information on the rupture probability. Another advantage of the probabilistic approach is to inform the effect of variables by parametric analysis. In this study, leak and LOCA probabilities in RCS piping of domestic nuclear power plants (NPPs) under fatigue and stress corrosion cracking are evaluated by using the probabilistic piping integrity evaluation program (P-PIE) developed in this study

  10. Simulation code of reactor core heatup during LOCA

    International Nuclear Information System (INIS)

    The computer code SCORCH-B2 is for analysis of the core heatup during a loss-of-coolant accident of BWR. The purpose of the code is to ensure that the BWR core keeps its coolable geometry throughout the postulated LOCA transient. In other words, it is to evaluate the peak cladding temperature and the maximum oxide thickness. The code examines only thermal behavior of the fuel assembly. Nuclear behavior of the core and hydrodynamic behavior of the coolant are given as input. Features of the code are as follows : Heat generation and trasmission are considered only on the horizontal plane at a certain elevation of a fuel assembly. The fuel rods are classified into a small number of groups. All the calculations are performed for representative fuel rods of groups. Radiative heat exchange is considered between fuel rods and the channelbox. View factors are re-computed when there occurs deformation of the cladding. The view factor is given as a function of the shadow areas or radiating and irradiated bodies by parallel rays. When the cladding temperature reaches a certain value given as input, all the rods in this group instantaneously balloon and rupture. After the deformation, surface area, gap conductance, view factors, etc. are changed and oxidation reaction is considered also on the cladding inner surface. (auth.)

  11. Thermal–hydraulic analysis and code assessment for ATLAS 6-inch cold leg break (SBLOCA) test using MARS-KS

    International Nuclear Information System (INIS)

    Highlights: • The thermal–hydraulic analysis using MARS-KS was conducted for 6-inch cold leg break test of ATLAS. • The assessment of the code capability to simulate the transient thermal–hydraulic behavior for SBLOCA was performed. • Additional discussions including comparison with RELAP5 were made for the cause of core level dip at SIT injection. • MARS-KS code has good capabilities to simulate cold leg break SBLOCA. • However, the condensation model needs to be improved to predict more accurate results. - Abstract: The thermal–hydraulic analysis using MARS-KS code was performed for 6-inch cold leg break test of ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation), which was the second domestic standard problem. The calculation results were compared with experimental data to assess the code capability to simulate the transient thermal–hydraulic behavior for small break loss of coolant accident (SBLOCA). The sequence of events, except for the location of loop seal clearing (LSC) and safety injection tank (SIT) injection time was predicted well. The loop seals of 1A and 2B intermediate legs were cleared at 398 s in the experiment, while that of 1A was only cleared in the calculation at the same time. The prediction showed good agreement with the experimental data for pressurizer pressure and break mass flow rate. The sudden decrease and increase of water level at the LSC time were predicted qualitatively. After LSC, there was significant water level dip at SIT injection time which was not seen in the experiment. In addition, sensitivity study to investigate the cause of core level dip at SIT injection time was performed and discussions were made for it. In conclusion, MARS-KS code has good capabilities to simulate cold leg break SBLOCA, however, including interfacial heat and mass transfer, especially condensation model needs to be improved to predict more accurate results

  12. Numerical analysis on coal-breaking process under high pressure water jet

    Energy Technology Data Exchange (ETDEWEB)

    Jin-hua Chen; Yun-pei Liang; Guo-qiang Cheng [Shandong University of Science and Technology, Qingdao (China)

    2009-09-15

    Based on the theory of nonlinear dynamic finite element, a control equation of coal and water jet was acquired in the coal breaking process under a water jet. A calculation model of coal breaking under a water jet was established; the fluid-structure coupling of water jet and coal was implemented by penalty function and convection calculation. The dynamic process of coal breaking under a water jet was simulated and analyzed by combining the united fracture criteria of the maximum tensile strain and the maximal shear strain in the two cases of damage to coal and damage failure to coal. 5 refs., 5 figs., 2 tabs.

  13. Numerical analysis on coal-breaking process under high pressure water jet

    Institute of Scientific and Technical Information of China (English)

    CHEN Jin-hua; LIANG Yun-pei; CHENG Guo-qiang

    2009-01-01

    Based on the theory of nonlinear dynamic finite element, the control equation of coal and water jet was acquired in the coal breaking process under a water jet. The calcu-lation model of coal breaking under a water jet was established; the fluid-structure cou-pling of water jet and coal was implemented by penalty function and convection calculation. The dynamic process of coal breaking under a water jet was simulated and analyzed by combining the united fracture criteria of the maximum tensile strain and the maximal shear strain in the two cases of damage to coal and damage failure to coal.

  14. Pipe rupture test results; 4 inch pipe whip test under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Pipe whip tests or jet discharge tests have been performed in JAERI, which simulate the instantaneous circumferential guillotine break of primary coolant piping in nuclear power plants. This report summarizes the results of 4 inch pipe whip tests (RUN 5407, 5501, 5504, 5603), under BWR LOCA conditions, which were performed from 1979 to 1981. The test pressure was 6.8 MPa and test temperature 2850C. In these tests clearance was kept constant at the value of 100 mm and overhang length were 250 mm, 400 mm, 550 mm and 1000 mm, respectively. The main purpose of these tests was to investigate the effect of overhang length on pipe whip behavior. From the tests results, a shorter overhang length is recommended to minimize the deformation of the pipe and restraints. (author)

  15. Multiple-pathway analysis of double-strand break repair mutations in Drosophila.

    Directory of Open Access Journals (Sweden)

    Dena M Johnson-Schlitz

    2007-04-01

    Full Text Available The analysis of double-strand break (DSB repair is complicated by the existence of several pathways utilizing a large number of genes. Moreover, many of these genes have been shown to have multiple roles in DSB repair. To address this complexity we used a repair reporter construct designed to measure multiple repair outcomes simultaneously. This approach provides estimates of the relative usage of several DSB repair pathways in the premeiotic male germline of Drosophila. We applied this system to mutations at each of 11 repair loci plus various double mutants and altered dosage genotypes. Most of the mutants were found to suppress one of the pathways with a compensating increase in one or more of the others. Perhaps surprisingly, none of the single mutants suppressed more than one pathway, but they varied widely in how the suppression was compensated. We found several cases in which two or more loci were similar in which pathway was suppressed while differing in how this suppression was compensated. Taken as a whole, the data suggest that the choice of which repair pathway is used for a given DSB occurs by a two-stage "decision circuit" in which the DSB is first placed into one of two pools from which a specific pathway is then selected.

  16. Tank Riser Pit Decontamination System (Pit Viper) Return on Investment and Break-Even Analysis

    International Nuclear Information System (INIS)

    This study assessed the cost benefit of Pit Viper deployment for 80 tank farm pits between October 1, 2003 and September 30, 2012 under the technical baseline for applicable double-shell tank (DST) and single-shell tank (SST) projects. After this assessment had been completed, the U.S. Department of Energy (DOE) Richland Operations Office (RL) and Office of River Protection (ORP) published the Hanford Performance Management Plan (August 2003), which accelerated the schedule for SST retrieval. Then, DOE/CH2M HILL contract modification M064 (October 2002) and The Integrated Mission Acceleration Plan (March 2003) further accelerated SST retrieval and closure schedules. Twenty-six to 40 tanks must be retrieved by 2006. Thus the schedule for SST pit entries is accelerated and the number of SST pit entries is increased. This study estimates the return on investment (ROI) and the number of pits where Pit Viper deployment would break even or save money over current manual practices. The results of the analysis indicate a positive return on the federal investment for deployment of the Pit Viper provided it is used on a sufficient number of pits

  17. Analysis of nanosecond breaking of a high-density current in SOS diodes

    Science.gov (United States)

    Grekhov, I. V.; Lyublinskii, A. G.; Smirnova, I. A.

    2015-11-01

    Effect of a sharp (nanosecond) breaking of the reverse current with a density on the order of 103-104 A/cm2 in a silicon diode upon switching from direct to reverse bias voltage (so-called silicon opening switch, or SOS effect) is widely used in nanosecond technologies of gigawatt powers. For detailed analysis of the SOS effect, we constructed a special setup with small stray inductance, which makes it possible to test single SOS diodes with a working area of 1-2 mm2 in a wide range of current densities. Our experiments show, in particular, that the numerical model of the SOS effect developed at the Institute of Electrophysics, Ural Branch, Russian Academy of Sciences successfully described the experimental results. It is also shown that the charge extracted from the diode structure by the reverse current exceeds the charge introduced by a direct current pulse by not more than 10%, indicating a relatively small role of ionization processes. The possibility to carry out experiments on single samples with a small surface area allows us to study the SOS effect and considerably facilitates investigations aimed at the perfection of the design of SOS diodes.

  18. Theoretical analysis and experimental study of oxygen transfer under regular and non-breaking waves

    Institute of Scientific and Technical Information of China (English)

    尹则高; 梁丙臣; 王乐

    2013-01-01

    The dissolved oxygen concentration is an important index of water quality, and the atmosphere is one of the important sources of the dissolved oxygen. In this paper, the mass conservation law and the dimensional analysis method are employed to study the oxygen transfer under regular and non-breaking waves, and a unified oxygen transfer coefficient equation is obtained with consi-deration of the effect of kinetic energy and wave period. An oxygen transfer experiment for the intermediate depth water wave is per-formed to measure the wave parameters and the dissolved oxygen concentration. The experimental data and the least squares method are used to determine the constant in the oxygen transfer coefficient equation. The experimental data and the previous reported data are also used to further validate the oxygen transfer coefficient, and the agreement is satisfactory. The unified equation shows that the oxygen transfer coefficient increases with the increase of a parameter coupled with the wave height and the wave length, but it de-creases with the increase of the wave period, which has a much greater influence on the oxygen transfer coefficient than the coupled parameter.

  19. A French guideline for defect assessment at elevated temperature and leak before break analysis

    Energy Technology Data Exchange (ETDEWEB)

    Drubay, B.; Chapuliot, St.; Lacire, M.H.; Marie, St. [CEA Saclay, Lab. d' Ingegrite des Structures et Normalisation, LISN, 91 - Gif sur Yvette (France); Deschanels, H. [FRAMATOME/Novatome, 69 - Lyon (France); Cambefort, P. [Electricite de France (EDF/SEPTEN), 69 - Lyon (France)

    2001-07-01

    A large program is performed in France in order to develop, for the design and operating FBR (fast breeder reactor) plants, defect assessment procedures and Leak-Before-Break methods (L.B.B.). The main objective of this A16 guide is to propose analytical solutions at elevated temperature coherent with those proposed at low temperature by the RSE-M. The main items developed in this A16 guide for laboratory specimen, plates, pipes and elbows are the following: evaluation of ductile crack initiation and crack propagation based on the J parameter and material characteristics as J{sub R}-{delta}a curve or J{sub i}/G{sub fr}. Algorithms to evaluate the maximum endurable load under increasing load for through wall cracks or surface cracks are also proposed; determination of fatigue or creep-fatigue crack initiation based on the {sigma} approach calculating stress and strain at a characteristic distance d from the crack tip; evaluation of fatigue crack growth based on da/dN-{delta}K{sub eff} relationship with a {delta}K{sub eff} derived from a simplified estimation of {delta}J for the cyclic load; evaluation of creep-fatigue crack growth adding the fatigue crack growth and the creep crack growth during the hold time derived from a simplified evaluation of C{sup *}; Leak-Before-Break procedure. The fracture mechanic parameters determined in the A16 guide (K{sub 1}, J, C{sup *}) are derived from handbooks and formula in accordance with those proposed in the RSE-M document for in service inspection. Those are: the K{sub I} handbook for a large panel of surface and through-wall defect in plates, pipes and elbows; elastic stress and reference stress formula; analytical Js and Cs{sup *} formulations for mechanical and through thickness thermal load. The main part of the formula and assessment methodologies proposed in the A16 guide are included in a software, called MJSAM, developed under the MS Windows environment in support of the document. This allows a simple application of

  20. A French guideline for defect assessment at elevated temperature and leak before break analysis

    International Nuclear Information System (INIS)

    A large program is performed in France in order to develop, for the design and operating FBR (fast breeder reactor) plants, defect assessment procedures and Leak-Before-Break methods (L.B.B.). The main objective of this A16 guide is to propose analytical solutions at elevated temperature coherent with those proposed at low temperature by the RSE-M. The main items developed in this A16 guide for laboratory specimen, plates, pipes and elbows are the following: evaluation of ductile crack initiation and crack propagation based on the J parameter and material characteristics as JR-Δa curve or Ji/Gfr. Algorithms to evaluate the maximum endurable load under increasing load for through wall cracks or surface cracks are also proposed; determination of fatigue or creep-fatigue crack initiation based on the σ approach calculating stress and strain at a characteristic distance d from the crack tip; evaluation of fatigue crack growth based on da/dN-ΔKeff relationship with a ΔKeff derived from a simplified estimation of ΔJ for the cyclic load; evaluation of creep-fatigue crack growth adding the fatigue crack growth and the creep crack growth during the hold time derived from a simplified evaluation of C*; Leak-Before-Break procedure. The fracture mechanic parameters determined in the A16 guide (K1, J, C*) are derived from handbooks and formula in accordance with those proposed in the RSE-M document for in service inspection. Those are: the KI handbook for a large panel of surface and through-wall defect in plates, pipes and elbows; elastic stress and reference stress formula; analytical Js and Cs* formulations for mechanical and through thickness thermal load. The main part of the formula and assessment methodologies proposed in the A16 guide are included in a software, called MJSAM, developed under the MS Windows environment in support of the document. This allows a simple application of the analysis proposed in the document. (authors)

  1. TRAC-PF1/MOD1 US/Japanese PWR conservative LOCA prediction

    Energy Technology Data Exchange (ETDEWEB)

    Gruen, G E; Fisher, J E

    1987-11-01

    This report documents the results of a 200%, double-ended, cold-leg-break, loss-of-coolant-accident (LOCA) calculation using the TRAC-PF1/MOD1 computer code. The reactor system represented a typical United States/Japanese pressurized water reactor with a 15 x 15 fuel bundle arrangement 12-ft long, four loops, and cold-leg Emergency Core Cooling (ECC) Systems. Conservation boundary and initial conditions were used. Reactor power was 102% of the 3250 MWt rated power, decay heat was set to 120% of American Nuclear Society Standard 5.1, highest core lifetime values for power peaking and fuel stored energy were used, and the LOCA occurred simultaneously with a loss of offsite power. Best estimate assumptions were used for the break flow model, fuel rod heat transfer and metal-water reaction correlations, and steady-state fuel temperature profiles. A flow blockage model, having the capability to account for the effects of cladding ballooning or rupturing, was not used. Except for these best estimate assumptions, the boundary and initial conditions were consistent with those used in licensing calculations. Maximum fuel rod temperatures were 1380 K (2020/sup 0/F) and 1040 K (1410/sup 0/F) on the hottest evaluation model rod and hottest best estimate rod, respectively. The high reported values or fuel cladding temperature were a direct consequence of the conservative boundary and initial conditions used for the calculation, primarily the 2% overpower condition, the core decay heat assumption, and the degraded ECCS. The calculation demonstrated successful core reflooding before 1478 K (2200/sup 0/F) cladding temperature was exceeded on any fuel rod. 7 refs., 47 figs., 5 tabs.

  2. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    International Nuclear Information System (INIS)

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

  3. Experimental simulation of LOCA in a PWHR : analytical study of similarity of thermal response between fuel rod simulators and nuclear fuel rods under reflood conditions

    International Nuclear Information System (INIS)

    For the safety analysis of a nuclear reactor, the similarity of the thermal response of an electrically heated Fuel Rod Simulator (FRS), mostly used in Loss-of-Coolant-Accident (LOCA) experiments, to that of a nuclear fuel rod is of great significance. The present analysis describes the characteristics and the similarity of thermal response fuel rods under reflood conditions of LOCA. The analysis has shown that the thermal response of a nuclear fuel rod can be well simulated by the use of an electrically heated FRS. (author). 7 refs., 12 figs

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

    International Nuclear Information System (INIS)

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

  5. Modelling of WWER fuel rod during LOCA conditions using FEM code ANSYS

    International Nuclear Information System (INIS)

    The report presents the results of the computer simulation of the IFA-650.6 experiment, the sixth test in Halden LOCA test project series, performed in May 18, 2007 with a pre-irradiated WWER-440 fuel with maximum burnup of 56 MWd/kgU. The thermo-mechanical analysis was fulfilled with the license finite element ANSYS code package.The calculation was carried out with the 2D axisymmetric and 3D problem definitions. Analysis of the calculational results shows that the ANSYS code can adequately simulate thermo-mechanical behavior of cladding under IFA-650.6 test conditions. (authors)

  6. Analysis of proteins phosphorylated at the early response of DNA double-strand breaks

    International Nuclear Information System (INIS)

    The serine/threonine protein kinase, ATM, recognizes DNA double-strand breaks (DSBs) at a very early stage, and plays a central role in transmitting damage signals to the cell cycle control machinery. To further understand the early events in DSBs recognition, we are undertaking two approaches: 1) identification of novel substrates which are phosphorylated by ATM using phospho-specific antibody, and 2) analysis of phosphorylation profile employing the proteomics-based 2-DE gel electrophoresis technique. There is a well-known consensus sequence for phosphorylation by ATM characterized by serine followed by glutamine (SQ), and their surrounding amino acids. Using one of the consensus sequences, we synthesized a phosphorylated peptide for immunizing rabbit and generated a phospho-specific antibody. This novel antibody recognized phosphorylated proteins in mouse fibroblast cultures after 5 min of exposure to ionizing radiation. This high and constant level of phosphorylation persisted until the 2nd hour and started decreasing to basal levels from the fourth hour. This antibody also detected phospho-proteins in mouse fibroblast cultures exposed to chemical reagents that induce DSBs such as NCS, bleomycin and etoposide, and is thought to be specific for DNA DSBs-related proteins. Furthermore, this antibody did not recognize phospho-proteins in ATM-deficient cells, suggesting that this phosphorylation is dependent on ATM kinase or downstream kinases controlled by ATM. As a second approach, we identified several spots that are phosphorylated in ATM wild type cells but not in ATM-deficient cells by 2-DE proteomic analysis. We are conducting further studies to generate comprehensive, ATM-dependent phosphorylation profiles as a novel approach to understand early events in DNA damage recognition by ATM

  7. Pressurized Thermal Shock Analysis for OPR1000 Pressure Vessel

    International Nuclear Information System (INIS)

    The study provides a brief understanding of the analysis procedure and techniques using ANSYS, such as the acceptance criteria, selection and categorization of events, thermal analysis, structural analysis including fracture mechanics assessment, crack propagation and evaluation of material properties. PTS may result from instrumentation and control malfunction, inadvertent steam dump, and postulated accidents such as smallbreak (SB) LOCA, large-break (LB) LOCA, main steam line break (MSLB), feedwater line breaks and steam generator overfill. In this study our main focus is to consider only the LB LOCA due to a cold leg break of the Optimized Power Reactor 1000 MWe (OPR1000). Consideration is given as well to the emergency core cooling system (ECCS) specific sequence with the operating parameters like pressure, temperature and time sequences. The static structural and thermal analysis to investigate the effects of PTS on RPV is the main motivation of this study. Specific surface crack effects and its propagation is also considered to measure the integrity of the RPV. This study describes the procedure for pressurized thermal shock analysis due to a loss of coolant accidental condition and emergency core cooling system operation for reactor pressure vessel.. Different accidental events that cause pressurized thermal shock to nuclear RPV that can also be analyzed in the same way. Considering the limitations of low speed computer only the static analysis is conducted. The modified LBLOCA phases and simplified geometry can is utilized to analyze the effect of PTS on RPV for general understanding not for specific specialized purpose. However, by integrating the disciplines of thermal and structural analysis, and fracture mechanics analysis a clearer understanding of the total aspect of the PTS problem has resulted. By adopting the CFD, thermal hydraulics, uncertainties and risk analysis for different type of accidental conditions, events and sequences with proper

  8. Transient analysis of primary heat transport system of Cirus reactor following 200% longitudinal break of lower header

    International Nuclear Information System (INIS)

    Full text: Cirus, India's second research reactor has been in operation at the Bhabha Atomic Research Centre, Trombay since 1960. It is a natural uranium fuelled, light water cooled, heavy water moderated and graphite reflected reactor of 40 MW power. This paper presents the details of an analysis carried out to predict the transient behaviour of primary heat transport (PHT) system of Cirus reactor in case of 200% longitudinal break of lower header (core outlet header). The availability of a large capacity emergency water storage tank (EWST) to provide core cooling following the break has been considered in the analysis. The analysis is carried out using the thermal hydraulic code RELAP4/MOD6. The results show that the maximum clad surface temperature is within the maximum specified limit and the fuel-clad integrity is maintained throughout the transient

  9. Estimation of Siphon Break in a Research Reactor using CFD analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hong Beom; Seo, Kyoungwoo; Kim, Seong Hoon; Chi, Dae Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The pool water also behaves as a shielding barrier for many kinds of radio-nuclides from the reactor core and the spent fuel. Pool water is essential for nuclear safety. So guaranteeing the pool water inventory to be higher than the required minimum level is one of the most important tasks of a research reactor design. The lowest pool penetration of cooling pipes should be located above the reactor core against a cooling pipe break. However, system components outside the pool can be installed below the core level due to the component purpose such as the acceptance of a Net Positive Suction Head(NPSH) of a pump for downward core flow research reactor. So the pool water can be drained below the core through siphon effect and the core can't be cooled through natural circulation when a postulated pipe break occurs below the reactor core position. ANSYS CFD is used to solve the Navier-Stokes equation with the turbulent model and two-phase model. Siphon breaker was designed to satisfy the minimum pool water level requirement during pipe break in a research reactor, and it is necessary to analyze siphon break phenomena. The results employing the various rupture size and siphon break line size were compared. Undershooting height increased with increasing rupture size and decreasing siphon break line size. In larger main pipe size, trend was similar and undershooting height increased. With loss coefficient near outlet region, undershooting height decreased. Based on a numerical simulation, it was evaluated that various parameters had effect on siphon break performance and it would help design of primary cooling system.

  10. Differentiation and analysis on rock breaking characteristics of TBM disc cutter at different rock temperatures

    Institute of Scientific and Technical Information of China (English)

    谭青; 张桂菊; 夏毅敏; 李建芳

    2015-01-01

    In order to study rock breaking characteristics of tunnel boring machine (TBM) disc cutter at different rock temperatures, thermodynamic rock breaking mathematical model of TBM disc cutter was established on the basis of rock temperature change by using particle flow code theory and the influence law of interaction mechanism between disc cutter and rock was also numerically simulated. Furthermore, by using the linear cutting experiment platform, rock breaking process of TBM disc cutter at different rock temperatures was well verified by the experiments. Finally, rock breaking characteristics of TBM disc cutter were differentiated and analyzed from microscale perspective. The results indicate the follows. 1) When rock temperature increases, the mechanical properties of rock such as hardness, and strength, were greatly reduced, simultaneously the microcracks rapidly grow with the cracks number increasing, which leads to rock breaking load decreasing and improves rock breaking efficiency for TBM disc cutter. 2) The higher the rock temperature, the lower the rock internal stress. The stress distribution rules coincide with the Buzin Neske stress circle rules: the maximum stress value is below the cutting edge region and then gradually decreases radiant around; stress distribution is symmetrical and the total stress of rock becomes smaller. 3) The higher the rock temperature is, the more the numbers of micro, tensile and shear cracks produced are by rock as well as the easier the rock intrusion, along with shear failure mode mainly showing. 4) With rock temperature increasing, the resistance intrusive coefficients of rock and intrusion power decrease obviously, so the specific energy consumption that TBM disc cutter achieves leaping broken also decreases subsequently. 5) The acoustic emission frequency remarkably increases along with the temperature increasing, which improves the rock breaking efficiency.

  11. Investigation into the in-box LOCA consequence and structural integrity of the KO HCCR TBM in ITER

    International Nuclear Information System (INIS)

    Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) and its auxiliary system in ITER. In parallel with its design, safety analysis has performed including accident analysis with the selected reference accidents. Among them, the effect of in-box LOCA to the structural integrity of the TBM was investigated. From the transient analysis of the GAMMA-FR on the in-box LOCA, it is found that the pressure of the internal TBM can be increased up to 8 MPa with the same pressure of the operating coolant through the Tritium Extraction System (TES) and He purge lines in the TBM. Structural analysis with ANSYS code for TBM was performed with this condition and it is confirmed that the TBM can endure and it does not affect the ITER machine by the failure

  12. Application of the Best Estimate Plus Uncertainty method to the small break LOCA with high pressure injection failure. Uncertainty quantification of the RELAP5 model related to fuel clad oxidation, decay heat, fuel clad deformation and SG U-tube condensation

    International Nuclear Information System (INIS)

    By applying the Best Estimate Plus Uncertainty (BEPU) method, the uncertainties of best estimate results can be estimated quantitatively, and as a consequence, excessive conservatism can be reasonably removed to obtain evaluation results with enhanced reliability. Application of the BEPU method is being made to analyses of the 'low pressure injection by intentional depressurization of the steam generator secondary side' which is in accident management approach in a SBLOCA (small break loss-of-coolant accident) with HPI (high pressure injection) failure. Among the 24th important phenomena extracted by making the PIRT (phenomena identification and ranking table), the phenomena which have the great influence on the PCT (peak clad temperature) and whose model uncertainties are able to be quantified without test analyses are 'fuel clad oxidization', 'decay heat', 'fuel clad deformation', and 'SG U-tube condensation'. In this research, the uncertainties of the correlation of the RELAP5/MOD3.2 model about these important phenomena were quantified. By comparison of the calculation results of the correlation and test results or arrangement with the error of decay heat standard data, the distribution of the RELAP5 model uncertainties were determined, that is to say, the distribution of the RELAP5 model uncertainties were quantised. (author)

  13. Prediction of Leak Flow Rate Using FNNs in Severe LOCA Circumstances

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Kim, Ju Hyun; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of); Hur, Seop; Kim, Chang Hwoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Leak flow rate is a function of break size, differential pressure ( i.e., difference between internal and external reactor vessel pressure), temperature, and so on. Specially, the leak flow rate is strongly dependent on the break size and the differential pressure, but the break size is not measured and the integrity of pressure sensors is not assured in severe circumstances. In this study, a fuzzy neural network (FNN) model is proposed to predict the leak flow rate out of break, which has a direct impact on the important times (time approaching the core exit temperature that exceeds 1200 .deg. F, core uncover time, reactor vessel failure time, etc.). Since FNN is a data-based model, it requires data to develop and verify itself. However, because actual severe accident data do not exist to the best of our knowledge, it is essential to obtain the data required in the proposed model using numerical simulations. These data were obtained by simulating severe accident scenarios for the optimized power reactor 1000 (OPR 1000) using MAAP4 code. In this study, FNN model was developed to predict the leak flow rate in severe post-LOCA circumstances.. The training data were selected from among all the acquired data using an SC method to train the proposed FNN model with more informative data. The developed FNN model predicted the leak flow rate using the time elapsed after reactor shutdown and the predicted break size, and its validity was verified in the basis of the simulation data of OPR1000 using MAAP4 code.

  14. Main Steam Line Break Analysis for the Fully Passive Safety System of SMART

    International Nuclear Information System (INIS)

    The standard design approval of SMART (System-integrated Modular Advanced ReacTor) developed by KAERI and KEPCO consortium was issued on July 4, 2012. Although SMART has enhanced safety compared to the conventional reactor, there is a demand to meet the 'passive safety performance requirements' after the Fukushima accident. The passive safety performance requirements are the capabilities to maintain the plant at a safe shutdown condition for a minimum of 72 hours without AC power supply or operator action in case of design basis accident (DBA). To satisfy the requirements, KAERI is developing a safety enhanced SMART by adopting a passive safety injection system. The passive safety injection system developed for SMART is a gravity-driven injection system, which consists of four trains, each of which includes a pressure balance line, core makeup tank (CMT), safety injection tank (SIT) and injection line. The CMT plays an important role to inject borated water into the RCS to prevent or dissolve the return to power (re-criticality) condition during the event of increase in heat removal by the secondary system. The main steam line break accident (MSLB) is the most limiting accident for an increase in heat removal by the secondary system. In this study, the safety analysis results of MSLBs at hot full power condition and at hot zero power condition in view of re-criticality are given. The MSLB accident has been analyzed for the SMART adopting fully passive safety system in the aspect of re-criticality. The results show that the core remains subcritical condition throughout the transient due to the borated water injected by the CMT. As further works, many kinds of analyses and sensitivity studies should be performed for the design establishment and improvement of the fully passive system of SMART

  15. Integral experiment and RELAP5 analysis for DVI line break SBLOCA in APR1400

    International Nuclear Information System (INIS)

    The thermal-hydraulic phenomena of Direct Vessel Injection (DVI) line Small-Break Loss-of-Coolant Accident (SBLOCA) in the pressurized water reactor, APR1400, were investigated. To understand the thermal-hydraulic phenomena during the SBLOCA transient, the reduced-height and reduced-pressure integral test loop, SNUF (Seoul National University Facility), was constructed according to the energy scaling methodology. The methodology conserves the mass inventory and energy of the system in the same time scale as the prototype. From the RELAP5 analysis, the energy scaling methodology was confirmed to show the reasonable transient when ideally scaled-down SNUF model was compared to the prototype model. In order to overcome the limitation of power in actual SNUF, the modified-power curve was utilized without simulating the forced flow by pump, so that those corrections did not affect the major phenomena during transient. Geometric distortion of actual SNUF also did not strongly disturb the thermal-hydraulic behaviors, especially occurrence of the downcomer seal clearing. In the experiments according to the conditions determined by energy scaling methodology, the phenomenon of downcomer seal clearing had a dominant role in decrease of the system pressure and increase of the coolant level of core. It occurred when the steam injected from cold legs penetrated the coolant in upper downcomer toward the broken DVI line. The experimental results was used to validate the calculation capability of RELAP5, especially for the downcomer seal clearing phenomenon, and to estimate the scale-up capability of RELAP5 code according to the scaling methodology. (authors)

  16. LOCA with consequential or delayed LOOP accidents: Unique issues, plant vulnerability, and CDF contributions

    International Nuclear Information System (INIS)

    A loss-of-coolant accident (LOCA) can cause a loss-of-offsite power (LOOP) wherein the LOOP is usually delayed by few seconds or longer. Such an accident is called LOCA with consequential LOOP, or LOCA with delayed LOOP (here, abbreviated as LOCA/LOOP). This paper analyzes the unique conditions that are associated with a LOCA/LOOP, presents a model, and quantifies its contribution to core damage frequency (CDF). The results show that the CDF contribution can be a dominant contributor to risk for certain plant designs, although boiling water reactors (BWRs) are less vulnerable than pressurized water reactors (PWRs)

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

    International Nuclear Information System (INIS)

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

  18. Thermal hydraulic analysis of main-steam-line-break accidents as potential initiators for reactor vessel pressurized thermal shock

    International Nuclear Information System (INIS)

    Results are presented from two thermal hydraulic analysis of postulated main-steam-line breaks for the Oconee nuclear power plant. One calculation assumes runaway feedwater supply, whereas normal feedwater management is used in the other. The analyses were performed with the TRAC-PD2 code. The objective was to provide primary coolant temperature and pressure histories to assist in evaluating possible reactor-vessel pressurized thermal-shock concerns

  19. Practical illustration of the traditional vers. alternative LOCA embrittlement criteria

    International Nuclear Information System (INIS)

    Evaluation of LOCA time behaviour is usually based on traditional embrittlement criterion, represented by the equivalent cladding reacted (ECR) limit 17 % (18 %) at the peak cladding temperature below 1204 0C (1200 0C). From different existing correlations for evaluation the ECR, the correlations of Baker-Just, Cathcart and VNIINM (Bibilashvili) are discussed here. Results, obtained by these correlations, are illustrated for typical and atypical LOCA courses analysed for the WWER 440 plant. An approach to assess these correlations from the viewpoint of violation of the observed criterion is presented. This approach is based on determination of the temperature vers. time of exposition, when the criterion limit is reached. Reasons leading to necessity of alternative criterion proposal are summarised. This criterion for LOCA events evaluation, including corresponding correlation, is proposed on the basis of the long-term experimental research of cladding materials at UJP Praha. The computational results, obtained according to this alternative criterion, are illustrated for the same courses of LOCA events as for traditional criteria and traditional correlations. Proposed criterion is also confronted with the other discussed criteria in accordance with mentioned approach presented in this paper. The characteristic experimental results and key findings are summarised. They substantiate and support the proposed alternative criterion. An advantage of the criterion is its independence on ECR, on hydrogen and oxygen content and on oxidation history, and its applicability to current Zr-based alloy cladding materials as well. This applicability is kept while preserving the simplicity of the criterion using. (author)

  20. Analysis of the water mass supplied to CAN GV's before the Steam Line Break

    International Nuclear Information System (INIS)

    When the theoretical Steam Line Break scenario is analyzed, in a PWR type reactor, the vapour generator level begins to descend. If additionally, conservatively, a fail in the control is supposed, the hypothetical conditions, for this study and this paper, appear.

  1. Precision Agriculture Equipment Ownership versus Custom Hire: A Break-even Land Area Analysis

    OpenAIRE

    Gandonou, Jean-Marc; Dillon, Carl; Shearer, Scott; Stombaugh, Tim

    2006-01-01

    Identifying the least-cost strategy of obtaining a technology is important. This study determined the break-even cropped area necessary to economically justify the purchase of Precision Agriculture (PA) equipment versus the custom hiring of the PA services. The results suggest that a commercial Kentucky grain farmers would purchase the PA equipment.

  2. Breaking through the Advertising Clutter: A Qualitative Analysis of Broken Stereotypes in Print and Television Advertisements.

    Science.gov (United States)

    Larson, Charles U.

    As a result of the overwhelming amount of print and electronic advertisements which compete for consumer attention, advertisers must find effective methods to get through the ad clutter and capture their audience's interest. Several tactics can accomplish this strategy, including the tactic of breaking or reversing audience expectations or…

  3. Thermal hydraulic analysis of the AHWR—The Indian thorium fuelled innovative nuclear reactor

    International Nuclear Information System (INIS)

    Highlights: • Advanced heavy water reactor. • Thermal hydraulics. • Safety analysis. • RELAP5. -- Abstract: Analysis has been carried out for simulating loss-of-coolant accident (LOCA) at inlet header in a natural circulation type reactor developed as the advanced heavy water reactor (AHWR).The paper will cover a case of LOCA due to 200% break at inlet header which is double ended rupture. The maximum clad surface temperature has been predicted in different cases by using the thermal hydraulic safety code RELAP5/Mod4.0. The proposed reactor is a 920 MWth vertical pressure tube type, boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is heat removal through natural circulation of primary coolant (at all allowed power levels) with no primary coolant pumps. This reactor is equipped with emergency core cooling system (ECCS) and isolation condensers (ICs) to remove decay heat during LOCA. This ECCS provides cooling to fuel in passive mode during first fifteen minutes of LOCA and it is achieved by high pressure injection from advanced accumulator. Cooling is continued for Later for three days by the gravity driven water pool (GDWP). This paper investigates the impact of high pressure injection in this cooling process

  4. Fuel Behaviour at High During RIA and LOCA Accidents

    International Nuclear Information System (INIS)

    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. Analysis of Hydrogen Source Term and Effectiveness of Hydrogen Control in Thousand Megawatt PWR Severe Accident%百万千瓦级压水堆严重事故下氢气源项及氢气空制有效性分析

    Institute of Scientific and Technical Information of China (English)

    邹杰; 佟立丽; 曹学武; 顾健; 薛峻峰; 江宇; 郝禄禄; 仇苏辰; 刘力

    2013-01-01

    针对百万千瓦级压水堆核电厂大型干式安全壳在严重事故情况下的氢气风险控制,建立了一体化事故分析模型,分别对大破口失水事故(LB-LOCA)、中破口失水事故(MB-LOCA)、小破口失水事故(SB-LOCA)、全厂断电事故(SBO)、蒸汽发生器(SG)传热管破裂事故(SGTR)以及主蒸汽管道破裂事故(MSLB)进行事故进程计算以及氢气源项分析.相对于其他事故序列,LB-LOCA下堆芯快速熔化,锆-水反应产生氢气的速率快,可以作为安全壳内氢气风险控制有效性分析的代表性事故序列.分析表明,严重事故情况下在安全壳中安装一定数量的非能动氢气复合器(PARs)能够有效去除安全壳中的氢气,消除氢气燃烧或爆炸的风险,保持安全壳的完整性.%The integrated severe accident analysis model of 100 MW PWR NPP is built to analyze the hydrogen risk under severe accidents.Large break loss of coolant accident (LB-LOCA),medium break loss of coolant accident (MB-LOCA),small break loss of coolant accident (SB-LOCA),station blackout (SBO),steam generator tube rupture (SGTR) and main steam line break (MSLB) are chosen as typical severe accident sequences to analyze the hydrogen source.Considering the hydrogen quantity of 100% zirconium water reaction,the LB-LOCA is selected as a representative sequence to evaluate the hydrogen mitigation system.The results show that a certain number of PARs could remove hydrogen and oxygen effectively,and protect the containment integrity against hydrogen deflagration or detonation.

  7. Decay Heat Removal Performance Analysis of AP1000 Startup Feed Water during Non-LOCA Accident%AP1000启动给水在非LOCA事故下的衰变热排出性能分析

    Institute of Scientific and Technical Information of China (English)

    吴昊; 甘泉; 罗琪; 肖三平; 刘妍; 陈树山

    2015-01-01

    为验证三代核电AP1000核电厂在非LOCA事故工况下,启动给水补给性能是否满足衰变热排出的纵深防御准则,保守认为事故发生后,反应堆停堆,厂用电及外电网丧失,主给水丧失,凝汽器热阱丧失,蒸汽发生器背压为安全阀最低整定压力,蒸汽发生器与启动给水泵均为单列可用.首先,验证凝结水储箱处于最低液位时,启动给水的最低补给能力能否满足不小于118.1 m3/h的准则要求;其次,论证事故后由于备用交流电源加载滞后而导致启动给水延后140 s投运,蒸汽发生器依靠自身缓冲水装量能否带走衰变热而不触发专设安全系统;再次,论证140 s后启动给水最低补给流量,能否稳定蒸汽发生器液位并使其回升;最后,验证凝结水储箱纵深防御水装量能否满足启动给水24 h连续补给的准则要求.本文通过对启动给水最低补给流量、蒸汽发生器缓冲水装量、启动给水液位控制,以及凝结水储箱水装量的保守计算分析,验证了AP1000启动给水在非失水事故(Non-LOCA)事故下衰变热排出功能设计的可靠性以及与纵深防御准则的一致性.

  8. The Relationship among Ethanol, Sugar and Oil Prices in Brazil: Cointegration Analysis with Structural Breaks

    OpenAIRE

    Chen, Bo; Saghaian, Sayed

    2015-01-01

    Ethanol has gradually gain momentum in the world’s energy market in recent decades with Brazil the largest producers. The issue of price linkage among ethanol, sugar and oil is particular interesting and important in the context of Brazilian sugarcane sector. By accounting for the possible structural breaks in the data, we investigate the price linkage of the three commodities and discover that prices are not cointegrated in the first sub-periods but cointegrated in the second sub-period. Als...

  9. Analysis of primary loop small-break loss-of-coolant accident

    International Nuclear Information System (INIS)

    On the basis of a typical model of the primary loop small-break loss-of-coolant accident, the transient variation of the thermo hydraulics parameters at a loss-of-coolant accident are calculated by RETRAN-02 code. The physical process and the relevant measures for protection under the incident condition are analyzed. The calculation result shows that the reactor has a favorable capacity to resist the accident. (authors)

  10. The purchasing power parity in emerging Europe: Empirical results based on two-break analysis

    Directory of Open Access Journals (Sweden)

    Mladenović Zorica

    2013-01-01

    Full Text Available The purpose of the paper is to evaluate the validity of purchasing power parity (PPP for eight countries from the Emerging Europe: Hungary, Czech Republic, Poland, Romania, Lithuania, Latvia, Serbia and Turkey. Monthly data for euro and U.S. dollar based real exchange rate time series are considered covering the period: January, 2000 - August, 2011. Given significant changes in these economies in this sample it seems plausible to assume that real exchange time series are characterized by more than one time structural break. In order to endogenously determine the number and type of breaks while testing for the presence of unit roots we applied the Lee-Strazicich approach. For two euro based real exchange rate time series (in Hungary and Turkey the unit root hypothesis has been rejected. For the U.S. dollar based real exchange rate time series in Poland, Romania and Turkey the presence of unit root has been rejected. To assess the adjustment dynamics of those real exchange rates that were detected to be stationary with two breaks, the impulse response function is calculated and half-life is estimated. Our overall conclusion is that the persistence of real exchange rate in Emerging Europe is still substantially high. The lack of strong empirical support for PPP suggests that careful policy actions are needed in this region to prevent serious exchange rate misalignment.

  11. R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron Simon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tu, Lei [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize

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

    International Nuclear Information System (INIS)

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

  13. SU-C-BRE-07: Sensitivity Analysis of the Threshold Energy for the Creation of Strand Breaks and of Single and Double Strand Break Clustering Conditions

    International Nuclear Information System (INIS)

    Purpose: To analyse the sensitivity of the creation of strand breaks (SB) to the threshold energy (Eth) and thresholding method and to quantify the impact of clustering conditions on single strand break (SSB) and double strand break (DSB) yields. Methods: Monte Carlo simulations using Geant4-DNA were conducted for electron tracks of 280 eV to 220 keV in a geometrical DNA model composed of nucleosomes of 396 phospho-diester groups (PDGs) each. A strand break was created inside a PDG when the sum of all energy deposits (method 1) or energy transfers (method 2) was higher than Eth or when at least one interaction deposited (method 3) or transferred (method 4) an energy higher than Eth. SBs were then clustered into SSBs and DSBs using clustering scoring criteria from the literature and compared to our own. Results: The total number of SBs decreases as Eth is increased. In addition, thresholding on the energy transfers (methods 2 and 4) produces a higher SB count than when thresholding on energy deposits (methods 1 and 3). Method 2 produces a step-like function and should be avoided when attempting to optimize Eth. When SBs are grouped into damage patterns, clustering conditions can underestimated SSBs by up to 18 % and DSBs can be overestimated by up to 12 % compared to our own implementation. Conclusion: We show that two often underreported simulation parameters have a non-negligible effect on overall DNA damage yields. First more SBs are counted when using energy transfers to the PDG rather than energy deposits. Also, SBs grouped according to different clustering conditions can influence reported SSB and DSB by as much as 20%. Careful handling of these parameters is required when trying to compare DNA damage yields from different authors. Research funding from the governments of Canada and Quebec. PP acknowledges partial support by the CREATE Medical Physics Research Training Network grant of the Natural Sciences and Engineering Research Council (Grant number: 432290)

  14. Analysis of breaks in pipe connections to the hot leg and to the loop seal in the primary system of Ringhals 2 PWR

    International Nuclear Information System (INIS)

    Analysis has been made of seven different cases of breaks in pipes connected to the hot leg and to the loop seal in Ringhals 2 PWR. The pipes, in which guillotine breaks are postulated, have nominal diameters ranging from 1 to 14 inches. The purpose of the analysis is to supplement the basis for a review of the inspection procedures for the safety of pressure vessels prescribed by SKI. A similar analysis already exists concerning breaks in the cold leg connections. The analysis has been made using the thermal hydraulic computer code RELAPS/MOD2 and with best estimate assumptions. This means that normal operating conditions have been adopted for the input to the calculations. However, the capacity of the safety injection system was assumed to be reduced by having one pump not operating each of the low pressure and high pressure safety injection system. The results of the analysis are presented in tables and as computer plots. The analysis shows that the consequences with respect to increased fuel rod and cladding temperatures are quite harmless. Only the two cases with the largest break sizes lead to critical heat flux (CHF) and increased temperatures for the hottest rods in the core. The peak cladding temperature was 636 degrees C for the 12 inch break. In both cases rewetting occurred within 25 s of accident initiation. In the cases with breaks in connections of 6 inch diameter and smaller the RELAP5 calculations indicated a substantial margin to CHF throughout the transient. (authors)

  15. UMCP 2 x 4 loop observations regarding the behavior of an integral system during SB-LOCA

    International Nuclear Information System (INIS)

    The test program at the University of Maryland, College Park (UMCP) 2x4 facility has conducted several series of small break-loss of coolant accidents (SB-LOCA) experiments. The accumulated data base is now sufficiently extensive that it is feasible to advance from the description of specific SB-LOCA transients to more generalized observations. The generality of the observations can be confirmed by comparison with the extensive data base generated by the once through integral system (OTIS) and multiloop integral system test (MIST) programs and selected test results provided by experiments at the SRI integral facility. A necessary initial step in the generalization of the extensive combined data base is a classification of the possible transient types and the identification of the observed flow modes. Several classification schemes have been employed. One divides the transient characteristics into the inherent response of the integral system itself and the modification of this response imposed by boundary conditions. Another scheme utilizes the observation that SB-LOCA transients can be divided into two dynamically different operational modes. These are quasi-steady state modes, and transition modes. The later can occur between two sequential quasi-steady state modes, but can also occur repeatedly for operational states which exhibit a cyclical character. It has been shown that the most dependable parameter for correlating the operational characteristics is the inventory of the primary system. The cyclical and oscillatory operational modes deserve special consideration. The causes and characteristics of oscillations vary. Some are generic and are observed in all of the integral system test (IST) facilities, a few are facility specific and can be related to specific atypicalities

  16. Çokal Dam-break model and flood risk analysis (Çanakkale

    Directory of Open Access Journals (Sweden)

    Hasan Özdemir

    2011-09-01

    Full Text Available The source of the hazard which is the main factor of disasters can be made by naturally or man. These are generally independent sources, but sometimes reason of one hazard such as flood can be both naturally and man-made. Certainly, all examples in our country and the world show that, the fail of constructed structures on the rivers (e.g. dam, embankment produce a very large amount of water and damage more than the normal river floods. In this study, based on cofferdam of Çokal Dam breaching which occurred in 16 November 2007 on the Kavak River (Çanakkale, 1D modeling of probable Çokal Dam break take in the account tectonic properties of the area and analyzing of the flood risk have been done. For these purposes, Digital Elevation Model (DEM gathered contours from 1:25000 scaled topographic maps and GPS points, high and medium resolution satellite images, hydrological soil data gathered from soil maps scaled 1:25000, precipitation and discharge data in 30 years, technical properties of structures on the Kavak River and field measurement have been used as a database. All these data is processed and analyzed using Geographic Information System (GIS, Hec-GeoRAS and HEC-RAS hydraulic models and hydrologic model. Hence, Çokal Dam break modeling based on cofferdam breach modeling reveals that probable flood after the dam-break will affect Evreşe Plain and the people which get livelihood from the plain. General probable lost in agricultural product after probable flood reaches TL 12 million. Thus, as a result of human interventions to the nature will cause great harm to himself again.   

  17. Çokal Dam-break model and flood risk analysis (Çanakkale

    Directory of Open Access Journals (Sweden)

    Hasan Özdemir

    2011-09-01

    Full Text Available The source of the hazard which is the main factor of disasters can be made by naturally or man. These are generally independent sources, but sometimes reason of one hazard such as flood can be both naturally and man-made. Certainly, all examples in our country and the world show that, the fail of constructed structures on the rivers (e.g. dam, embankment produce a very large amount of water and damage more than the normal river floods. In this study, based on cofferdam of Çokal Dam breaching which occurred in 16 November 2007 on the Kavak River (Çanakkale, 1D modeling of probable Çokal Dam break take in the account tectonic properties of the area and analyzing of the flood risk have been done. For these purposes, Digital Elevation Model (DEM gathered contours from 1:25000 scaled topographic maps and GPS points, high and medium resolution satellite images, hydrological soil data gathered from soil maps scaled 1:25000, precipitation and discharge data in 30 years, technical properties of structures on the Kavak River and field measurement have been used as a database. All these data is processed and analyzed using Geographic Information System (GIS, Hec-GeoRAS and HEC-RAS hydraulic models and hydrologic model. Hence, Çokal Dam break modeling based on cofferdam breach modeling reveals that probable flood after the dam-break will affect Evreşe Plain and the people which get livelihood from the plain. General probable lost in agricultural product after probable flood reaches TL 12 million. Thus, as a result of human interventions to the nature will cause great harm to himself again.

  18. STRUCTURAL BREAKS, COINTEGRATION, AND CAUSALITY BY VECM ANALYSIS OF CRUDE OIL AND FOOD PRICE

    Directory of Open Access Journals (Sweden)

    Aynur Pala

    2013-01-01

    Full Text Available This papers investigated form of the linkage beetwen crude oil price index and food price index, using Johansen Cointegration test, and Granger Causality by VECM. Empirical results for monthly data from 1990:01 to 2011:08 indicated that evidence for breaks after 2008:08 and 2008:11. We find a clear long-run relationship between these series for the full and sub sample. Cointegration regression coefficient is negative at the 1990:01-2008:08 time period, but adversely positive at the 2008:11-2011:08 time period. This results represent that relation between crude oil and food price chanced.

  19. A Sectoral Efficiency Analysis of Malaysian Stock Exchange Under Structural Break

    Directory of Open Access Journals (Sweden)

    Chin W. Cheong

    2008-01-01

    Full Text Available We investigated the weak-form market efficiency of nine daily sectoral indices of Malaysian stock market between 1996 and 2006. The structural break unit root tests evidenced most of the price indices characterized by mean-reverting process that violated the random walk process. These empirical results were in sharp contrast with the traditional unit-root test which ignored the economic crisis and currency control. Our findings concluded that the Malaysian sectoral stock markets were weak-form inefficient (except the property index under the structural change.

  20. 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.

  1. Theory and Application of Loss of Life Risk Analysis for Dam Break

    Institute of Scientific and Technical Information of China (English)

    孙月峰; 钟登华; 李明超; 李颖

    2010-01-01

    The loss of life risk evaluation model for dam break is built in this paper.By using an improved Monte Carlo method,the overtopping probability induced by concurrent flood and wind is calculated,and the Latin Hypercube Sampling is used to generate random numbers.The Graham method is used to calculate the loss of life resulting from dam failure.With Dongwushi reservoir located at Hebei Province taken as an example,the overtopping probability induced by concurrent flood and wind is calculated as 4.77×10-6.Los...

  2. Transient deformation properties of Zircaloy for LOCA simulation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hann, C. R.; Mohr, C. L.; Busness, K. M.; Olson, N. J.; Reich, F. R.; Stewart, K. B.

    1980-05-01

    This experimental data report is Volume 4 of a series of 5 volumes describing the oxidation and deformation rate behavior of Zircaloy cladding under simulated LOCA conditions. It contains listings of strain versus stress, time, and temperature evaluated from the numerical constitutive relationships and the original data used to develop them. This volume also contains listings of the ramp load, pressure, and temperature test data from both current and previous phases of the series, as well as material describing applications of the data.

  3. Life assessment and ageing management of LOCA qualified electrical equipment

    International Nuclear Information System (INIS)

    The LOCA (Loss of Coolant Accident) qualification of electrical and control equipment has to be done considering the operational loads during the service life at the locations in the plant in which a piece of equipment is expected to be operated. If not explicitly measured, the operational loads have to be assumed and applied prior to the LOCA load test. Measuring campaigns showed that values of the operational loads were different compared to the ones originally assumed. The isolated qualification of every single piece of equipment of a functional chain may not have been done considering the effects of the contribution of the other members of the functional chain simultaneously affected by the accidental load during a LOCA. And, last but not least, life time extension of NPP will have to be addressed in the near future. Facing this it was decided together with the German utilities to launch a program in the first instance with the aim of on-going qualification. In the course of this program all relevant qualification data for all German NPP were collected; several measuring campaigns in different NPP were performed in order to determine the operational loads at the locations of the relevant pieces of equipment; all relevant activation energy data of the materials constituting the qualified equipment were collected in order to evaluate the thermal ageing; procedures to handle equipment made out of different sensitive materials were developed, and at the very end a software based tool was provided which allows to calculate and record the remaining qualified life times - once the customer is in possession of compatible data - for each member of the functional chains, and the total sum of the leakage current comparing it with threshold values not allowed to be exceeded. The major results and the conclusions drawn, how to handle life assessment and ageing management of LOCA qualified electrical equipment in the future, will be presented

  4. Transient deformation properties of Zircaloy for LOCA simulation. Final report

    International Nuclear Information System (INIS)

    This experimental data report is Volume 4 of a series of 5 volumes describing the oxidation and deformation rate behavior of Zircaloy cladding under simulated LOCA conditions. It contains listings of strain versus stress, time, and temperature evaluated from the numerical constitutive relationships and the original data used to develop them. This volume also contains listings of the ramp load, pressure, and temperature test data from both current and previous phases of the series, as well as material describing applications of the data

  5. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Flow clogging characteristics were investigated based on data for the relation of pressure loss and flow velocity during flow clogging due to debris accumulation. Deposition of chemical precipitates on the fuel cladding using an electrically heated rod was investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis with a thermal-hydraulic code on the downstream effect has shown that the core could be cooled because the core inlet flow compensates a evaporation of coolant due to the decay-heat even if core inlet was 99% clogged just after the ECCS recirculation operation started during the cold-leg break LOCA in PWR plants. (author)

  6. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  7. Best estimate and uncertainty analysis of a critical large break loss of coolant accident at Darlington NGS

    International Nuclear Information System (INIS)

    This paper briefly describes the development and application of a Best Estimate Analysis with Uncertainty (BEAU) methodology to a critical large break loss of coolant accident at Darlington NGS. Best estimate and uncertainty predictions are obtained for maximum fuel centreline temperature, maximum fuel sheath temperature, maximum fuel string relative axial expansion, and peak pressure tube strain. The results are compared against similar results obtained using the Limit of Operating Envelope (LOE) approach reported in the most recent licensing submission. The comparison shows significant improvements in predicted safety margins can be achieved. (author)

  8. Past break-monsoon conditions detectable by high resolution intra-annual δ18O analysis of teak rings

    Science.gov (United States)

    Managave, S. R.; Sheshshayee, M. S.; Borgaonkar, H. P.; Ramesh, R.

    2010-03-01

    Intra-annual variations in the cellulose oxygen isotopic composition (δ18O) of several annual growth rings of three teak (Tectona grandis L.F.) trees from central India show a clear seasonal cycle with higher values in the early and late growing seasons and lower values in the middle. This cycle is useful to identify growth occurring during different phases of the growing season. Relative humidity (RH) appears to control the intra-annual δ18O variations rather than rainfall, and therefore past break-monsoon conditions associated with lower RH, could be detected by high resolution sub-sampling of annual rings for δ18O analysis.

  9. String breaking

    CERN Document Server

    Bali, G S; Lippert, T; Neff, H; Prkacin, Z; Schilling, K; Bali, Gunnar S; Dussel, Thomas; Lippert, Thomas; Neff, Hartmut; Prkacin, Zdravko; Schilling, Klaus

    2006-01-01

    We numerically investigate the transition of the static quark-antiquark string into a static-light meson-antimeson system. Improving noise reduction techniques, we are able to resolve the signature of string breaking dynamics for Nf=2 lattice QCD at zero temperature. We discuss the lattice techniques used and present results on energy levels and mixing angle of the static two-state system. We visualize the action density distribution in the region of string breaking as a function of the static colour source-antisource separation. The results can be related to properties of quarkonium systems.

  10. Analysis of the main steam line break benchmark (Phase II) using ANCK/MIDAC code

    International Nuclear Information System (INIS)

    The three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANCK/MIDAC has been developed. ANCK/MIDAC consists of the 3D nodal kinetic code ANCK and the 3D drift flux T/H code MIDAC. In order to verify the adequacy of ANCK/MIDAC, the Phase II problem in the 'OECD main steam line break benchmark (MSLB benchmark)' was analyzed. This MSLB benchmark has been defined in order to simulate the core response and the reactor coolant system response to a relatively severe steam line break accident condition. The Phase II problem has a conservative condition that the control rod with the maximum worth is stuck in a fully withdrawn position throughout the transient. The simulation was performed using the core inlet temperatures and flow rates for 18 different regions, which were provided by the PSU best-estimate TRAC-PF1/NEM calculations. The comparison of the ANCK/MIDAC results with other participants' results shows the excellent agreement on main core parameters. ANCK/MIDAC has good capability and reliability for the best-estimation code and a reasonable calculation time. (author)

  11. The complex Langevin analysis of spontaneous symmetry breaking induced by complex fermion determinant

    CERN Document Server

    Ito, Yuta

    2016-01-01

    In many interesting physical systems, the determinant which appears from integrating out fermions becomes complex, and its phase plays a crucial role in the determination of the vacuum. An example of this is QCD at low temperature and high density, where various exotic fermion condensates are conjectured to form. Another example is the Euclidean version of the type IIB matrix model for 10d superstring theory, where spontaneous breaking of the SO(10) rotational symmetry down to SO(4) is expected to occur. When one applies the complex Langevin method to these systems, one encounters the singular-drift problem associated with the appearance of nearly zero eigenvalues of the Dirac operator. Here we propose to avoid this problem by deforming the action with a fermion bilinear term. The results for the original system are obtained by extrapolations with respect to the deformation parameter. We demonstrate the power of this approach by applying it to a simple matrix model, in which spontaneous symmetry breaking from...

  12. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [EDF/SEPTEN Villeurbanne (France); GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  13. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    International Nuclear Information System (INIS)

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the α/β transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates

  14. Fault tree analysis of tailings pond dam break%尾矿库溃坝的事故树分析

    Institute of Scientific and Technical Information of China (English)

    梁强; 司悦彤; 侯克鹏; 付学会

    2013-01-01

      This paper uses the fault tree method to analyze tailings dam break ,and constructs a fault tree of tailings pond.It calculates minimum cut sets and minimum path sets and analyses the structure importance ,so the key elementary events can be found which affect the stability of tailings dam .The analysis result shows that the focus of preventing dam break should be put on avoiding seepage failure without timely measures and heavy rainfalls .According to the analysis above,many steps should be taken in advance,and it has important guiding significance to prevent tail-ings dam break.%  利用事故树方法对尾矿库溃坝灾害进行分析,建立适合尾矿库的事故树,并进行最小割集、最小径集的计算和基本事件结构重要度的分析,找出影响尾矿坝稳定性的关键基本事件。分析结果表明,要重点预防发生渗流破坏但未及时采取措施进行处理,以及预防汛期雨量过大对尾矿库的影响,提前作出应对措施。这为防止尾矿库溃坝灾害的发生,具有重大指导意义。

  15. Supersymmetry breaking

    Indian Academy of Sciences (India)

    Emilian Dudas

    2009-01-01

    We review the various mechanisms of supersymmetry breaking and its trans-mission to the observable sector. We argue that hybrid models where gauge dominates over gravity mediation, but gravity provides the main contributions to the Higgs sector masses and the neutralino mass, are able to combine the advantages and reduce the disadvantages of the two transmission mechanisms.

  16. Locações operacionais e financeiras: estudo empírico das empresas cotadas no Euronext 100

    OpenAIRE

    Costa, Marta Mendes Ferreira Gomes Pereira

    2010-01-01

    As locações representam, actualmente, uma importante fonte de financiamento para muitas empresas. Contudo, a existência de dois tipos de locações (locação financeira e locação operacional), permite contabilizar operações substancialmente semelhantes de forma distinta. Baseada na literatura sobre o tema das locações, a presente investigação tem como objectivo determinar que factores distinguem as sociedades que recorrem apenas a um dos tipos de locação das que recorrem a ambos o...

  17. Calculation of Departure from Nucleate Boiling Ratio (DNBR) minimum for accident analysis of main steam line break at Angra-1

    International Nuclear Information System (INIS)

    The maintenance costs, the operational problems and the failures possibilities of the boron injection system, composed by pumps, valves, heated lines and the boron injection tank, make this tank removal or the boron concentration reduction advisable for Angra 1 Power Plant. The main accident from chapter XV of the final safety analysis report affected by this modification is the main steam line break. It is necessary the interaction of the areas of Accidents and Transients Analysis (RETRAN 02/Mod 5.1 code), Neutronics (APA System) and Thermohydraulics (COBRA IIIC/MIT) to analyse this accident. The present Angra 1 boron concentration is 20000 ppm and it could be reduced to 2000 ppm as a result of the present study. The Departure from Nucleate Boiling Ratio (DNBR) is the restrictive parameter of this accident, which is calculated from the initials and boundary conditions obtained from the Transients and Accidents Analysis and Neutronics areas. (author)

  18. Wave packet analysis and break-up length calculations for an accelerating planar liquid jet

    International Nuclear Information System (INIS)

    This paper examines the process of transition to turbulence within an accelerating planar liquid jet. By calculating the propagation and spatial evolution of disturbance wave packets generated at a nozzle where the jet emerges, we are able to estimate break-up lengths and break-up times for different magnitudes of acceleration and different liquid to air density ratios. This study uses a basic jet velocity profile that has shear layers in both air and the liquid either side of the fluid interface. The shear layers are constructed as functions of velocity which behave in line with our CFD simulations of injecting diesel jets. The non-dimensional velocity of the jet along the jet centre-line axis is assumed to take the form V (t) = tanh(at), where the parameter a determines the magnitude of the acceleration. We compare the fully unsteady results obtained by solving the unsteady Rayleigh equation to those of a quasi-steady jet to determine when the unsteady effects are significant and whether the jet can be regarded as quasi-steady in typical operating conditions for diesel engines. For a heavy fluid injecting into a lighter fluid (density ratio ρair/ρjet = q < 1), it is found that unsteady effects are mainly significant at early injection times where the jet velocity profile is changing fastest. When the shear layers in the jet thin with time, the unsteady effects cause the growth rate of the wave packet to be smaller than the corresponding quasi-steady jet, whereas for thickening shear layers the unsteady growth rate is larger than that of the quasi-steady jet. For large accelerations (large a), the unsteady effect remains at later times but its effect on the growth rate of the wave packet decreases as the time after injection increases. As the rate of acceleration is reduced, the range of velocity values for which the jet can be considered as quasi-steady increases until eventually the whole jet can be considered quasi-steady. For a homogeneous jet (q = 1), the

  19. Pretest analysis of containment studies facility model for simulated loss of coolant accident conditions

    International Nuclear Information System (INIS)

    An experimental facility called Containment Studies Facility (CSF) has been constructed at Bhabha Atomic Research Centre (BARC), Trombay for the purpose of research and development in the area of nuclear reactor containment thermal hydraulics. The facility consists of reinforced concrete containment structural model and a Primary Heat Transport Model (PHTM) vessel. The containment model is approximately 1:250 volumetrically scaled down model of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) containment system and the PHTM represents the primary heat transport system of the prototype reactor. The PHTM with a pressure vessel and associated pump and piping system is designed for simulating the Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) conditions within the containment model. As part of CSF project thermal hydraulic analysis, a pretest analysis was carried out for simulated LOCA conditions. Blow down mass and energy discharge data were obtained using Relap/MOD3.2 code for different blow down conditions and were used as inputs to CONTRAN code for simulating LOCA or main steam line break (MSLB) conditions in the containment model. Pressure and temperature transients in the CSF model for different blow down conditions and a number of parametric studies were conducted to assess the influence of a large number of thermodynamic and geometrical parameters which are known to affect the transients and alter the peak pressure and temperature values. (author)

  20. Analysis of the OECD main steam line break benchmark using ANC-K/MIDAC code

    International Nuclear Information System (INIS)

    A three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANC-K/MIDAC has been developed. It is the combination of the 3D nodal kinetic code ANC-K and the 3D drift flux thermal hydraulic code MIDAC. In order to verify the adequacy of this code, we have performed several international benchmark problems. In this paper, we show the calculation results of ''OECD Main Steam Line Break Benchmark (MSLB benchmark)'', which gives the typical local power peaking problem. And we calculated the return-to-power scenario of the Phase II problem. The comparison of the results shows the very good agreement of important core parameters between the ANC-K/MIDAC and other participant codes. (author)

  1. Give me a better break: Choosing workday break activities to maximize resource recovery.

    Science.gov (United States)

    Hunter, Emily M; Wu, Cindy

    2016-02-01

    Surprisingly little research investigates employee breaks at work, and even less research provides prescriptive suggestions for better workday breaks in terms of when, where, and how break activities are most beneficial. Based on the effort-recovery model and using experience sampling methodology, we examined the characteristics of employee workday breaks with 95 employees across 5 workdays. In addition, we examined resources as a mediator between break characteristics and well-being. Multilevel analysis results indicated that activities that were preferred and earlier in the work shift related to more resource recovery following the break. We also found that resources mediated the influence of preferred break activities and time of break on health symptoms and that resource recovery benefited person-level outcomes of emotional exhaustion, job satisfaction, and organizational citizenship behavior. Finally, break length interacted with the number of breaks per day such that longer breaks and frequent short breaks were associated with more resources than infrequent short breaks. PMID:26375961

  2. Follow-up Study of ITER Safety Analysis : Large In-vessel First Wall Pipe Break with Wet Confinement Bypass

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    Previous researches have been analyzed risk assessments of fusion reactors that are dangerous in the severe accidents where the radioactive material released from confinement building to the environment. To simulate the severe accidents in ITER, a number of thermal hydraulics simulation codes were used. Before construction of the fusion reactor, to obtain ITER license about safety issue, MELCOR is chosen as one of the several codes to be used to perform ITER safety analyses. Qualification of the simulation code is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. In the nuclear fusion system, the amount of released radioactive material is criteria for safety permission. Tritium (or tritiated water: HTO) and radioactive dust aerosol are the source of radioactive leakage. In the Generic Site Safety Report (GSSR) for the ITER plant, Table I lists the release guidelines for tritium and activation products for normal operation, incidents and accidents. Several accident analyses have been studied to know how much radioactive material could be released from the severe accidents. In the present work, The MELCOR input deck of large First Wall (FW) coolant leak (pipe break) is used to study and radioactive material leakage thorough bypass accident are studied to follow up the ITER safety analysis. In this research, follow-up study of the in-vessel inboard/inboard-outboard FW pipe break was analyzed to investigate the amount of leakage of radioactive aerosol. All of the accident cases released the lower amount of radioactive aerosol compared to the IAEA guide lines. In addition, the OBB pipe break made lower HTO aerosol leakage because of condensation of HTO and adsorption between coolant and aerosol.

  3. Follow-up Study of ITER Safety Analysis : Large In-vessel First Wall Pipe Break with Wet Confinement Bypass

    International Nuclear Information System (INIS)

    Previous researches have been analyzed risk assessments of fusion reactors that are dangerous in the severe accidents where the radioactive material released from confinement building to the environment. To simulate the severe accidents in ITER, a number of thermal hydraulics simulation codes were used. Before construction of the fusion reactor, to obtain ITER license about safety issue, MELCOR is chosen as one of the several codes to be used to perform ITER safety analyses. Qualification of the simulation code is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. In the nuclear fusion system, the amount of released radioactive material is criteria for safety permission. Tritium (or tritiated water: HTO) and radioactive dust aerosol are the source of radioactive leakage. In the Generic Site Safety Report (GSSR) for the ITER plant, Table I lists the release guidelines for tritium and activation products for normal operation, incidents and accidents. Several accident analyses have been studied to know how much radioactive material could be released from the severe accidents. In the present work, The MELCOR input deck of large First Wall (FW) coolant leak (pipe break) is used to study and radioactive material leakage thorough bypass accident are studied to follow up the ITER safety analysis. In this research, follow-up study of the in-vessel inboard/inboard-outboard FW pipe break was analyzed to investigate the amount of leakage of radioactive aerosol. All of the accident cases released the lower amount of radioactive aerosol compared to the IAEA guide lines. In addition, the OBB pipe break made lower HTO aerosol leakage because of condensation of HTO and adsorption between coolant and aerosol

  4. MELCOR Analyses of Divertor Ex-vessel LOCA During Normal Operation. Contract EFDA 01/599, Deliverable 3 - Final Report

    International Nuclear Information System (INIS)

    A MELCOR model of ITER-FEAT divertor cooling system has been developed for the analyses of thermal-hydraulic accidents as specified in the Accident Analysis Specifications (AAS-3) for the ITER-FEAT Generic Site Safety Report (GSSR). The model is based on data from the Safety Analysis Data List (SADL-3). The report presents the results of DV ex-vessel LOCA with plasma shutdown from MELCOR calculations. The intention is to verify previous analyses with ATHENA and INTRA to update parts of GSSR documenting the analysis of representative accident sequences for ITER

  5. The Break

    DEFF Research Database (Denmark)

    Strand, Anete Mikkala Camille

    2016-01-01

    The chapter elaborates on how to deal with one of the major challenges facing organizations worldwide; Stress. The Break enacts a quantum approach to meet the challenges by proposing a combination of three different quantum storytelling technologies; protreptic mentoring, walking and material sto...... provider and witness to your elaborations. It’s really that simple! The chapter concludes towards a set of Dogmas for future reference in addressing these challenges in this manner....

  6. Leak-before-break analysis of a dissimilar metal welded joint for connecting pipe-nozzle in nuclear power plants

    International Nuclear Information System (INIS)

    Highlights: ► Leak-before-break (LBB) analysis for a dissimilar metal weld joint (DMWJ) is made. ► Pipe-nozzle geometry and inhomogeneous material property of DMWJ are incorporated. ► LBB behavior of a defect can be assessed by LBB assessment diagram and LBB curve. ► Feasibility region of LBB is enlarged with decreasing load and increasing JR. -- Abstract: This paper presents a leak-before-break (LBB) analysis for a dissimilar metal welded joint (DMWJ) connected the safe end to pipe-nozzle of a reactor pressure vessel of which is relevant to safety of nuclear power plant. Three-dimensional finite element analysis models were built for the DMWJ structure, and the initial inner circumferential surface cracks were postulated at the interface between A508 steel and buttering Alloy82. Based on the elastic–plastic fracture mechanics theory of J-integral, the crack growth stability was analyzed, and the pipe-nozzle geometry effect and inhomogeneous material properties of the DMWJ have been incorporated. Base on the analysis results, the LBB curves and LBB assessment diagrams were constructed for the DMWJ, and effects of applied bending moment loads and J-resistance curves of materials on LBB behavior were analyzed. The results show that the LBB behavior of a defect in the DMWJ under an upmost severe load can be assessed and predicted by plotting the defect size and its propagation path in the LBB assessment diagrams. With decreasing the maximum bending moment load and increasing the crack growth resistance of materials, the ligament instability lines shift upward and the critical crack length lines move to the right in the LBB assessment diagrams, which leads to enlargement of the feasibility region in the LBB behavior

  7. Investigation of main coolant pump trip problem in case of SB LOCA for Kozloduy Nuclear Power Plant, WWER-440/V230

    International Nuclear Information System (INIS)

    Highlights: • In this study we investigated scenarios with trip of MCP in case of SB LOCA. • The reference power plant for the analyses is Unit 4 at Kozloduy NPP. • The RELAP/MOD 3.2 computer code is used in performing the analyses. • The results are done in support of development of SB EOPs. - Abstract: This paper presents the results of thermal-hydraulic calculation of accident scenarios that involve the trip of main coolant pump (MCP) in case of Small break loss of coolant accident (SB LOCA) for WWER-440/V230 units at Kozloduy Nuclear Power Plant (KNPP), done in support of the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The main purpose of these analyses is to show how the different time of MCP switching off results in primary inventory depletion in case of SB LOCA and it is reflect on peak cladding temperature. According to this, the SB LOCA scenario is regarded from the point of view of an inadequate core cooling. Therefore, the primary concern is Critical Safety Function (CSF) “Core cooling” and “Primary inventory”. High core residual heat, minimal safety injection flow and other initial conditions challenging the mentioned CSFs are the main particularities of the accepted scenarios. The RELAP5/MOD3.2 computer code has been used to perform the analyses in a WWER-440 Nuclear Power Plant (NPP) model. A model of WWER-440 based on Unit 4 of Kozloduy NPP has been developed for the system’s thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia

  8. Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs

    International Nuclear Information System (INIS)

    SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

  9. Development of main steam line break mass and energy release analysis methodology with RETRAN-3D code

    International Nuclear Information System (INIS)

    The estimation methodology of the mass and energy (M/E) release in the main steam line break (MSLB) has been developed with the RETRAN-3D code. In the case of equipment qualification (EQ), the over-estimated temperature would exceed the design limits of some cables or valves. In order to have a more flexible EQ profiles from the MSLB M/E release, the methodology with the best-estimated code was used. The major conditions affecting the MSLB M/E were found to be the initial SG level, heat transfer between primary and secondary sides, power level, operable protection system, main or auxiliary feedwater availability, and break conditions. The RETRAN-3D models were developed for the Kori unit 1(KRN-1) which is typical two loop Westinghouse (WH) designed plant. Particularly, a detailed model of the steam generators was developed to estimate a more realistic two-phase heat transfer effect of the steam flow. After the modeling, the methodology has been developed through the sensitivity analyses. The M/E release data generated from the analyses have been used as the input to the inside containment pressure and temperature (P/T) analysis. According to the results at the point of view containment P/T, the Kori unit 1 can have more margin of 5 - 15 kPa in pressure and 8 - 15degC in temperature. (author)

  10. Level-Swell Prediction With RETRAN-3D And Its Application To A BWR Steam-Line-Break Analysis

    International Nuclear Information System (INIS)

    Level-swell experiments have often been simulated using system codes, such as TRAC and RELAP, but only cursory assessments have been performed with the operational-transient code RETRAN-3D, the main system code used within the STARS project. The present study, initiated in the framework of a BWR Steam-Line-Break (SLB) accident scenario, addresses this lacuna by performing RETRAN simulations of the General Electric Level-Swell experiments, and by investigating their implications on power plant accident analyses. Parameters to which the predicted level swell is sensitive have been identified, and recommendations on code options are made. The SLB analysis objective was to determine the amount of steam and liquid discharged through the break under specified boundary conditions, and to gauge the results against reference values. The impact of the nodalization of the upper part of the reactor pressure vessel was investigated and found to play an important role, whereas the level swell induced from flashing was found not to be the predominant factor for these simulations. (author)

  11. Review of the licensing basis for RIA and LOCA transients in light of new evidence on high burnup fuel behavior

    International Nuclear Information System (INIS)

    The fuel of power reactors can be damaged by both rapid reactivity insertion accidents (RIA) and by Loss of Coolant Accidents (LOCA) due to the break of a coolant pipe. The consequences of such postulated accidents with a very low probability can be mitigated by dedicated safety systems. Yet, the possible consequences need to be assessed as part of the safety evaluation in the framework of the reactor licensing process as these transients form part of the so-called design basis. Most importantly, coolability of the reactor core must be demonstrated. The current paper reviews the important fuel related phenomena that may occur during such transients and their relation to the current licensing basis. Finally, possible changes in light of new experimental evidence from high burnup fuel behavior tests are discussed. (author)

  12. Calculations of partial LOCA in a swimming-pool-reactor with MTR-elements and planned mock-up experiment

    International Nuclear Information System (INIS)

    A partial uncovering of the MTR fuel plates of the swimming pool reactor SAPHIR located at the Swiss Federal Reactor Research Institute (E.I.R.) could be caused by a loss of coolant accident due to a beam tube break. The transient temperature excursions of the fuel plates during the LOCA have been predicted with computer simulations. Because a reliable prediction of the flow regime and hence the heat transfer in the uncovered part of the plate is not possible with current knowledge, a parametric study employing different heat transfer models is presented in this paper. The results show, that the heat transfer model in the uncovered part of the fuel plate has an important influence on the predicted temperatures. A mock-up experimental facility, which will supply data for the heat transfer occuring in the uncovered part, will also be described at the end of the paper. (author)

  13. Leak rate assessment in the leak before break analysis of a PWR piping system

    International Nuclear Information System (INIS)

    In order to validate predictive models used in the L.B.B. (Leak Before Break) procedure for PWR, a cooperation program between CEA, EDF and FRAMATOME is carried out. One item concerns water leak rate assessment through a real fatigue crack under representative thermal hydraulic conditions of a PWR primary loop. In the framework of this cooperation program, experimental tests are carried out at CEA Cadarache. These tests are dedicated to the validation of two-phase flow rate models. For this, a pipe mock up, with a scale factor of two compared to the primary loop, is tested under representative pressure and temperature of 155 bars and 320 Celsius degrees. The real through wall crack is obtained from an initial surface notch pre-cycled under a fatigue loading. The steam flowing out of the pipe from the crack is condensed and weighed to have an experimental measurement of the leak rate. Crack opening area is measured with an optical device coupled to a post processing software. Acoustic gages are also located on the test pipe in order to qualify a detection system based on an acoustic measurement. The aim of the paper is to present the experimental program and results obtained in the first campaign. The validation of a two-phase leak rate model with experimental results is also discussed in this paper, as well as the comparison of crack opening area assessment, as proposed in the L.B.B. procedure, with experimental results. (authors)

  14. Analysis of DNA double-strand break repair pathways in mice

    International Nuclear Information System (INIS)

    During the last years significant new insights have been gained into the mechanism and biological relevance of DNA double-strand break (DSB) repair in relation to genome stability. DSBs are a highly toxic DNA lesion, because they can lead to chromosome fragmentation, loss and translocations, eventually resulting in cancer. DSBs can be induced by cellular processes such as V(D)J recombination or DNA replication. They can also be introduced by exogenous agents DNA damaging agents such as ionizing radiation or mitomycin C. During evolution several pathways have evolved for the repair of these DSBs. The most important DSB repair mechanisms in mammalian cells are nonhomologous end-joining and homologous recombination. By using an undamaged repair template, homologous recombination ensures accurate DSB repair, whereas the untemplated nonhomologous end-joining pathway does not. Although both pathways are active in mammals, the relative contribution of the two repair pathways to genome stability differs in the different cell types. Given the potential differences in repair fidelity, it is of interest to determine the relative contribution of homologous recombination and nonhomologous end-joining to DSB repair. In this review, we focus on the biological relevance of DSB repair in mammalian cells and the potential overlap between nonhomologous end-joining and homologous recombination in different tissues

  15. Analysis of the ATLAS main steam line break experiment using the MARS code

    International Nuclear Information System (INIS)

    In this study, a main steam line break (MSLB) test at the ATLAS facility was simulated using the best-estimate thermal-hydraulic system code, MARS. This has been performed as an activity at the third domestic standard problem for code benchmark (DSP-03) that has been organized by Korea Atomic Energy Research Institute (KAERI). The results of the MSLB experiment and the MARS input data prepared for the previous DSP-02 using the ATLAS facility were provided to participants. The preliminary MSLB simulation using the base input data, however, showed unphysical calculation results in the primary-to-secondary heat transfer. To resolve the problems, two improvements were implemented in the MARS input modelling, which includes the use of fine meshes for the bottom region of the steam generator secondary side and proper thermal-hydraulics calculation options, finally leading to realistic results. In addition, other input model improvements, such as considering heat loss or flow restrictor modelling, were made and the results are analyzed in detail. (author)

  16. Uncertainty analysis for containment response of U.S. EPR TM reactor to large break loss-of-coolant accidents

    International Nuclear Information System (INIS)

    This paper presents an uncertainty analysis applying the GOTHIC containment analysis code to simulate the first 24-hours following a large-break loss-of-coolant accident (LBLOCA) in AREVA's U.S. EPR TM plant. The uncertainty method is modeled after a study performed by the Gesellschaft fur Anlagen und Reaktorsicherheit (GRS) using data from the Heifidampf-Reaktor (HDR) Test T31.5. The analysis method incorporates an assessment of phenomenological importance, identifying the dominant contributors that influence the principle analysis metric, containment pressure. As with the GRS approach, this study employs non-parametric statistics. This analysis illustrates U.S. EPR containment response sensitivity to realistic variation in a set of important model parameters influencing containment conditions during LBLOCA. In considering a set of model uncertainty parameters, a number of GOTHIC variation calculations were performed (59 calculations) to effect a best estimate plus uncertainty result at 95/95 coverage/confidence level for the key metric, containment pressure. The results of the importance analysis showed condensation phenomena on the surface of the containment structures to be important during the passive cooling period, which occurred prior to the start of HL (hot leg) injection of SI (safety injection). In this study, hot leg injection was assumed to initiate at 1.5 hours. Condensation phenomena faded in importance after 1.5 hours due to the hot leg injection of SI suppressing steaming. Structure conduction, especially, the physical properties of concrete, retained importance throughout the transient. (authors)

  17. Considerations on burn-up dependent RIA and LOCA criteria

    International Nuclear Information System (INIS)

    For RIA transients, a fuel failure threshold has been derived and compared with recent experimental data relevant for BWR and PWR fuel. The threshold can be applied to HZP and CZP transients, account taken for the different initial enthalpy and for the lower ductility at cold conditions. It can also be used for non-zero power transients, provided that a term accounting for the initial power is incorporated. The proposed threshold predicts reasonably well the results obtained in the CABRI and NSRR tests when the different state of the cladding, i.e. ductile or brittle, is taken into account. Apart from some exceptions discussed in the paper, such as the effect of oxide spalling, one should consider ductile state for HZP conditions and brittle state for CZP conditions. The threshold applies equally well to UO2 and MOX fuel, but the database on MOX is limited. For LOCA transients, the cladding limit may decrease with burn-up due to cladding corrosion and hydrogen pick-up. A provisional criterion shows that the predicted burn-up effect is moderate or negligible if one uses the results obtained with actual high burn-up cladding. On the other hand, a large effect is predicted based on the results obtained with non-irradiated, pre-hydrided cladding specimens. There is a question however on as to whether these specimens can be representative for high burn-up material. The experimental evidence is still scarce and more data on high burn-up cladding is needed in order to arrive to firm conclusions. Most of the data currently available relates to Zr-4 cladding. The experiments made on ZIRLO and M5 cladding show that these alloys have a RIA and LOCA behaviour similar to or better than Zr-4. However, the data is limited, especially for LOCA conditions, where only un-irradiated specimens have been tested so far. (author)

  18. ROSA-II experimental program for PWR LOCA/ECCS integral tests

    International Nuclear Information System (INIS)

    This paper is the final report of the ROSA-II experimental program, in which summary of the integral test results on thermal hydraulic behavior in a loss-of-coolant accident (LOCA) of pressurized water reactor (PWR) and on the effect of emergency core cooling system (ECCS) is presented. The ROSA-II test facility has a volume scaling factor of approximately 1/400 and core heating power of 2.4 MW. Specific feature of the facility is the versatility of the break conditions, the ECCS injection conditions and the secondary system conditions. After numbers of integral tests under various test conditions, (1) condensation-depressurization effect due to ECC water, (2) stored heat release from the structural materials and (3) counter current flow limitation (CCFL) at the specific locations were found to be important phenomena for the core cooling. To supply cooling water as soon as possible to the core was indicated to be very important for successful core cooling. Based on these results, more effective ECCS was proposed and the effectiveness of the proposed ECCS was experimentally verified. On the other hand, part of the experimental data was utilized to evaluate the predictability of RELAP-3 and RELAP-4J computer codes. (author)

  19. Calculated and measured behaviour of zircaloy fuel sheaths during in-reactor LOCA tests

    International Nuclear Information System (INIS)

    Six CANDU type fuel elements, containing UO2 fuel and sheathed in Zircaloy, have been subjected to transient conditions simulating a hypothetical large break loss of coolant accident (LOCA) in a CANDU reactor. The maximum transient sheath temperature was about 1273 K. Two single-element experiments were conducted at Chalk River Nuclear Laboratories in the X-2 loop of the NRX reactor. The other four elements experienced a single transient in the Power Burst Facility at Idaho National Engineering Laboratory. Comparisons are presented between data from these experiments and calculations by ELOCA, a computer code simulating fuel performance during transient conditions. The parameters evaluated included fuel sheath strain, internal element gas pressure, the mechanisms and timing of fuel element failure, fuel centreline temperature, sheath microstructure, and the thicknesses of zirconia and oxygen stabilized alpha-Zr layers on the sheaths. ELOCA calculations agreed well with the data. Some of the work described here was jointly funded by Atomic Energy of Canada Limited and Ontario Hydro, a Canadian utility, through the co-operative research and development program, CANDEV

  20. Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

  1. Pipe rupture test results: 4-inch pipe whip tests under PWR LOCA conditions

    International Nuclear Information System (INIS)

    This report summarizes the results of 4-inch pipe whip tests (RUN No. 5506, 5507, 5508 and 5604) under the PWR LOCA conditions. The dynamic behaviors of the test pipe and restraints were studied in the tests. In the tests, the gap between the test pipe and the restraints was kept at the constant value of 8.85 mm and the overhang length was varied from 250 mm to 650 mm. The dynamic behaviors of the test pipe and the restraint were made clear by the outputs of strain gages and the measurements of residual deformations. The data of water hammer in subcooled water were also obtained by the pressure transducers mounted on the test pipe. The main conclusions obtained from the tests are as follows. (1) The whipping of pipe can be prevented more effectively as the overhang length becomes shorter. (2) The load acting on the restraint-support structure becomes larger as the overhang length becomes shorter. (3) The restraint farther from the break location does not limit the pipe movement except for the first impact when the overhang length is long. (4) The ultimate moment M sub(u) of the pipe at the restraint location can be used to predict the plastic collapse of the whipping pipe. (5) The restraints slide along the pipe axis and are subjected to bending moment, when the overhang length is long. (author)

  2. Best-estimate LOCA radiation signature for equipment qualification

    International Nuclear Information System (INIS)

    The radiation aspect of reactor equipment qualification depends on a knowledge of the appropriate source term. An attempt has been made to define a realistic radiation source corresponding to the loss-of-coolant accident. This best-estimate source is based on available fission product release data from damaged fuel during an unterminated LOCA as described in the Reactor Safety Study (WASH-1400). Energy release rates as a function of time have been calculated for both betas and gamma rays. The results are significantly different from the sources specified in Regulatory Guide 1.89. Spectra corresponding to the best-estimate source have also been computed at selected cooling times

  3. Inter-system LOCA risk assessment

    International Nuclear Information System (INIS)

    Inter-systems loss-of-coolant accidents (ISLOCAs) have been included in probabilistic risk assessments (PRAs) since WASH-1400. While estimated as being relatively low contributors to core damage frequency, ISLOCAs have been identified as major contributors to risk at nuclear power plants (NPPs). They have the potential to result in core melt and containment bypass, which may lead to the early release of large quantities of fission products. Recent events at several operating reactors have been identified as ISLOCA precursors. The occurrence of these events have raised concerns that the frequency of ISLOCA sequences might be underestimated in current state-of-the-art PRAs. In order to expand the current state-of-the-art, a Nuclear Regulatory Commission research program is being conducted by ED and G Idaho, Inc. at the Idaho National Engineering Laboratory. The objective of the ISLOCA research program is to generate qualitative and quantitative information on the hardware, human factors, and accident consequence issues that dominate nuclear power plant risks for ISLOCA. To meet this objective, the approach being taken includes analysis of all interfaces between the primary reactor coolant system and other, lower pressure systems. This historical experience (primarily, licensee event reports) has provided the basis for determining the scope of the analysis with respect to potential failure mechanisms of the pressure isolation boundary. It is important to note that in the vast majority of these events, the dominant failure was a human error. Because of their significance, human errors are given particular attention in the present analysis

  4. The Analysis of the Effectiveness of Simultaneous Inversion of Turning and Head Waves First Breaks - Model Study

    Science.gov (United States)

    Kasina, Zbigniew

    2012-09-01

    In the presented paper the model data were used to analyse the effectiveness of simultaneous inversion of the turning and head waves first breaks in comparison with the effectiveness of the inversion of only first breaks of turning waves or head waves. The analysis was undertaken for the gradient velocity models of the near surface layer with the low' velocity anomaly and for the CDP aquisition. The effect of the numerical ray tracing on the traveltime calculations and inversion results was estimated taking into account the wave equation modeling of seismic records. The effect of the errors of the starting velocity field m the process of inversion as well as the effect of spatial smothing of resulting velocity fields were considered too. The analysis confirmed some improvement in the imaging of the near surface velocity anomalies when we use simultaneous inversion of the turning and head waves first breaks. W przedstawionej pracy wykorzystano dane modelowe do analizy efektywności jednoczesnej inwersji pierwszych wstąpień fal czołowych i refragowanych w porównaniu do efektywności inwersj i tylko pierwszy ch wstąpień fali refragowanej lub czołowej. Analizę podjęto dla gradientowych modeli strefy przypowierzchniowej z niskoprędkościową anomalią dla akwizycji metody pokryć wielokrotnych. Oszacowano wpływ numerycznego trasowania promieni na wyniki obliczeń czasów przebiegu i inwersji uwzględniając wyniki modelowania rekordów sejsmicznych z równania falowego. Rozważano także wpływ błędów startowego pola prędkości w procesie inwersji, jak również wpływ przestrzennego wygładzania wynikowych pól prędkości. Analiza potwierdziła pewną poprawę w odwzorowaniu anomalii prędkościowych strefy przy- powierzchniowej, gdy wykorzystujemy jednoczesną inwersję pierwszych wstąpień fal czołowych i refragowanych.

  5. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    This report present the thermal-hydraulic analysis of 2 in-vessel Loss-of-Coolant Accidents (LOCAs) in the first wall cooling system of the Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. It has been calculated that about 53% (almost 12000 kg) of the initial coolant inventory will be discharged into the vacuum vessel during the first 490 s of the intermediate break in-vessel LOCA. About 5% (1130 kg) of the initial coolant inventory will be discharged into the vacuum vessel during the first 570 s of the small break in-vessel LOCA. The calculations of the vacuum vessel response will be carried out by the Gesellschaft fuer Reaktorsicherheit (GRS). The analyses of the vacuum vessel response will be based on the characteristics of the expelled cooling water provided by the Netherlands Energy Research Foundation (ECN), Petten, Netherlands, in the present study. (A.S.). 13 refs.; 79 figs.; 11 tabs

  7. A novel encoding water level monitor system during and after LOCAs in a nuclear heating reactor

    International Nuclear Information System (INIS)

    The water level in a nuclear reactor vessel is an important parameter during and after LOCAs. Nuclear safety specifications can not be carried out when the water level is measured using a pressurizer which does show the level in the vessel. It is difficult to monitor the water level in the vessel of a Daqing 200MW nuclear heating reactor (NHR-200) using the present differential pressure transducers. Based on the heat transfer differences between water (or liquid) and steam (or gas), a novel level detector, which includes encoding heating shell thermocouples, has been developed and verified for use experimentally under pressures of 0.15-3.0 MPa. A novel encoding water level monitoring system was designed, made up of an assembly that contains several detectors, a signal encoder and an intelligent processor. Analysis and experiments have shown that the new system is correct in principle, reliable and feasible in structure for monitoring the water level in the NHR-200 reactor. (orig.)

  8. A new water level monitor system for nuclear reactor vessel during and after LOCAs

    International Nuclear Information System (INIS)

    The water level in nuclear reactor vessel is an important parameter during and after LOCAs. It can not meet the nuclear safety specification to use the water level measured in the pressurizer to show the level in the vessel. It is difficult to monitor the water level in the vessel of NHR-200 nuclear heating reactor with present differential pressure transducers. A new level detector based on the heat transfer difference between water (or liquid) and steam (or gas) is developed and proven for use by experiments under the pressure 0.15-3.0 MPa. A new water level monitor system including a detector assembly that contains several detectors, a signal encoder and an intellectual processor, is designed. The analysis and experiments show that the new system is correct in principle, reliable in working properties and feasible in structure for monitoring the water level in the NHR-200 reactor

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

    International Nuclear Information System (INIS)

    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. Analysis of in vivo and in vitro DNA strand breaks from trihalomethane exposure

    Directory of Open Access Journals (Sweden)

    DeAngelo Anthony

    2004-01-01

    Full Text Available Abstract Background Epidemiological studies have linked the consumption of chlorinated surface waters to an increased risk of two major causes of human mortality, colorectal and bladder cancer. Trihalomethanes (THMs are by-products formed when chlorine is used to disinfect drinking water. The purpose of this study was to examine the ability of the THMs, trichloromethane (TCM, bromodichloromethane (BDCM, dibromochloromethane (DBCM, and tribromomethane (TBM, to induce DNA strand breaks (SB in (1 CCRF-CEM human lymphoblastic leukemia cells, (2 primary rat hepatocytes (PRH exposed in vitro, and (3 rats exposed by gavage or drinking water. Methods DNA SB were measured by the DNA alkaline unwinding assay (DAUA. CCRF-CEM cells were exposed to individual THMs for 2 hr. Half of the cells were immediately analyzed for DNA SB and half were transferred into fresh culture medium and incubated for an additional 22 hr before testing for DNA SB. PRH were exposed to individual THMs for 4 hr then assayed for DNA SB. F344/N rats were exposed to individual THMs for 4 hr, 2 weeks, and to BDCM for 5 wk then tested for DNA SB. Results CCRF-CEM cells exposed to 5- or 10-mM brominated THMs for 2 hr produced DNA SB. The order of activity was TBM>DBCM>BDCM; TCM was inactive. Following a 22-hr recovery period, all groups had fewer SB except 10-mM DBCM and 1-mM TBM. CCRF-CEM cells were found to be positive for the GSTT1-1 gene, however no activity was detected. No DNA SB, unassociated with cytotoxicity, were observed in PRH or F344/N rats exposed to individual THMs. Conclusion CCRF-CEM cells exposed to the brominated THMs at 5 or 10 mM for 2 hr showed a significant increase in DNA SB when compared to control cells. Additionally, CCRF-CEM cells exposed to DBCM and TBM appeared to have compromised DNA repair capacity as demonstrated by an increased amount of DNA SB at 22 hr following exposure. CCRF-CEM cells were found to be positive for the GSTT1-1 gene, however no activity

  11. Breaking Symmetries

    CERN Document Server

    Peters, Kirstin

    2010-01-01

    A well-known result by Palamidessi tells us that {\\pi}mix (the {\\pi}-calculus with mixed choice) is more expressive than {\\pi}sep (its subset with only separate choice). The proof of this result argues with their different expressive power concerning leader election in symmetric networks. Later on, Gorla of- fered an arguably simpler proof that, instead of leader election in symmetric networks, employed the reducibility of "incestual" processes (mixed choices that include both enabled senders and receivers for the same channel) when running two copies in parallel. In both proofs, the role of breaking (ini- tial) symmetries is more or less apparent. In this paper, we shed more light on this role by re-proving the above result-based on a proper formalization of what it means to break symmetries-without referring to another layer of the distinguishing problem domain of leader election. Both Palamidessi and Gorla rephrased their results by stating that there is no uniform and reason- able encoding from {\\pi}mix i...

  12. The Break

    DEFF Research Database (Denmark)

    Strand, Anete Mikkala Camille; Larsen, Jens

    2015-01-01

    to Explore your Leadership” . ”Time to reflect closer to heaven as we did in the Pyrenees, makes me humble and simplifies the thoughts on how to lead within my own set of values. It´s all about energy”, (Lars Lund Hansen, manager, Novo Nordisk) A few objects; a neckless, a candle, a dragon and five crystal...... terrain break elaborates the terrain of Organizations anno 2015 as a terrain of complexity, streamlining, language-orientation and dis-functionality. The latter in regard to a WHO acknowledged concern for health issues related to work-related stress (Prætorius, 2012) and an ongoing urge for learning...... that language and the social has been granted too much power on the dispense of the bodily, physical and biological – or in short, in dispense of the material. The break To be or not to be poses the theoretical notion of dis-/continuity (Barad, 2007, 2010) from the quantum approach to storytelling (Strand 2012...

  13. SPES-2, the full-height, full-pressure test facility simulating the AP600 plant comparison among 2' small break tests located on different lines

    International Nuclear Information System (INIS)

    SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL, ENEA, SIET and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November'94, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with both passive and active non-safety systems, and a main steam line break transient to demonstrate the capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. The author will describe the results obtained during experimental test facility having the same break size (2 inches) but located in different plant positions (cold leg, Direct Vessel Injection line, cold leg-CMT balance line) in order to determine the effect of break location on the plant behaviour

  14. Bayesian analysis of the predictive power of the yield curve using a vector autoregressive model with multiple structural breaks

    OpenAIRE

    Katsuhiro Sugita

    2015-01-01

    In this paper we analyze the predictive power of the yield curve on output growth using a vector autoregressive model with multiple structural breaks in the intercept term and the volatility. To estimate the model and to detect the number of breaks, we apply a Bayesian approach with Markov chain Monte Carlo algorithm. We find strong evidence of three structural breaks using the US data.

  15. Revolutionary change and structural breaks: A time series analysis of wages and commodity prices in Britain 1264-1913

    OpenAIRE

    Casson, Catherine; Fry, J. M.

    2011-01-01

    In this paper we empirically test the hypothesis that economic revolutions are associated with structural breaks in historical economic data. A simple test for structural breaks in economic time series is applied to British wage and price data from the medieval to the modern period. Evidence for structural change is found in nearly half of the series studied -- suggesting that structural breaks are an intrinsic feature of such historic data. Structural changes are most closely linked to th...

  16. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

    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 15000C; 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. Analysis of Angra-1 fuel rod during the large break loss-of-coolant accident

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  19. Finite Element Analysis of PVC window profile &aluminium window profile with and without thermal break

    Directory of Open Access Journals (Sweden)

    ENG. Mohammad Buhemdi

    2016-05-01

    Full Text Available Examine a thermal analysis .Numerous analogies exist between thermal and structuralanalysis for PVC window profile &aluminium window profile with and without thermalbreak ,Finite Element Analysis, commonly called FEA, is a method of numerical analysis. FEA isused for solving problems in many engineering disciplines such as machine design,acoustics, electromagnetism, soil mechanics, fluid dynamics, and many others. Inmathematical terms, FEA is a numerical technique used for solving field problemsdescribed by a set of partial differential equations. In mechanical engineering, FEA iswidely used for solving structural, vibration, and thermal problems. However, FEA is notthe only available tool of numerical analysis. Other numerical methods include the FiniteDifference Method, the Boundary Element Method, and the Finite Volumes Method tomention just a few. However, due to its versatility and numerical efficiency, FEA has cometo dominate the engineering analysis software market, while other methods have beenrelegated to niche applications. When implemented into modern commercial software,both FEA theory and numerical problem formulation become completely transparent tousers.

  20. Leak-Before-Break analysis for Pickering 'A' Unit 1 and Unit 4 large diameter main steam line pipes

    International Nuclear Information System (INIS)

    This paper presents a Leak-Before-Break (LBB) analysis of large diameter main steam line pipes (i.e. NPS 28'' and 30'') running from reactor building to main steam balance header in Pickering nuclear plant Unit 1 and Unit 4. Recent development in LBB technology summarized in U.S. Nuclear Regular Commission report NUREG/CR-6765 was adopted. Based on the tiered approach of LBB philosophy, this LBB analysis belongs to level 2 or level 3 LBB analysis. Detailed fracture tolerance analyses and leakage rate calculations were performed. EPFM (elastic plastic fracture mechanics) theory of J-integral, resistance curve versus ductile crack extension was adopted in carrying out all fracture tolerance analyses. Through-wall cracks in axial and circumferential directions on both straight pipes and elbows were postulated and analyzed. The loads applied on the postulated cracked pipes were obtained from detailed piping stress analysis under deadweight load, design pressure, thermal expansion, seismic design based earthquake (DBE) and thrust load due to the opening of relief valves. J-resistance data were derived from the lowest fracture toughness testing data obtained from Ontario Power Generation's PHT (primary heat transport) LBB material testing programs. A margin of 2 on crack size was chosen in establishing maximum allowable crack sizes. Leakage rates were calculated using SQUIRT Windows Version 1.1 program. The fluid inside the main steam line pipes was assumed single phase steam at 100% quality. One tenth of the calculated leakage rates was proposed as the requirement for minimum leakage detection capability. The paper concludes that the absence of through-wall crack larger than 91.16 mm in length should be maintained in order to ensure the structural integrity of large diameter main steam line pipes. In lieu of this crack size requirement, a reliable leakage detection capability which could quantify mass steam leakage rate of 0.01678 kg per second, or volume leakage rate