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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. 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)

  3. 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

  4. 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

  5. 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.

  6. Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.

    1984-08-01

    A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.

  7. 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)

  8. 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)

  9. 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.

  10. 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.)

  11. Containment Performance Analysis with Large Break LOCA for EU-APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Do Hyun; Lee, Keun Sung; Kim, Yong Soo [KHNP CRI, Daejeon (Korea, Republic of)

    2013-10-15

    In this paper for containment performance analysis, the containment pressurization analysis is performed and thermo-hydraulic response analysis of containment structure is carried out to provide basic understanding of containment transient states under a severe accident sequence. Especially, in EU-APR1400 design, to reduce containment pressure and temperature, Severe Accident Containment Spray System (SACSS) is designed to be actuated automatically when Core Exit Temperature (CET) reaches 922 K (649 .deg. C). The containment performance analysis was carried on LBLOCA sequence for EU-APR1400 with SACSS through MAAP code. If SACSS is actuated when CET reaches 922 K (649 .deg. C) , the containment pressure and temperature decrease to a sufficient low level. The predicted atmospheric pressure of containment will not exceed the ultimate pressure capacity (UPC) and have a sufficient margin to it even though the UPC of the reference plant (Shin-Kori Units 3 and 4) is used instead because the UPC calculation for EU-APR1400 has not been completed. The largest load on the containment by LBLOCA is estimated at 306.1 kPa. Thus the margin to UPC is estimated to be 330 % in comparison with 1.329 MPa as UPC for the reference plant.

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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)

  3. 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)

  4. Revisiting large break LOCA with the CATHARE-3 three-field model

    Energy Technology Data Exchange (ETDEWEB)

    Valette, Michel, E-mail: michel.valette@cea.fr [CEA Grenoble, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France); Pouvreau, Jerome, E-mail: jerome.pouvreau@cea.fr [CEA Grenoble, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France); Bestion, Dominique, E-mail: dominique.bestion@cea.fr [CEA Grenoble, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France); Emonot, Philippe, E-mail: philippe.emonot@cea.fr [CEA Grenoble, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France)

    2011-11-15

    Highlights: Black-Right-Pointing-Pointer CATHARE 3 enables a three-field analysis of a LB LOCA. Black-Right-Pointing-Pointer Reflooding experiments in isolated rod bundles are satisfactory predicted. Black-Right-Pointing-Pointer A BETHSY integral test simulation supports the CATHARE 3 3-field assessment. - Abstract: Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in a rod bundle is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit: core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic six-equation model is used in the other parts of the loop. An analysis of these first results is presented and future work is defined for improving the droplet behavior simulation in both the upper plenum and the

  5. 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)

  6. 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

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    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.

  9. 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

  10. 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

  11. 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

  12. 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.

  13. Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments

    Energy Technology Data Exchange (ETDEWEB)

    Berta, V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-05-01

    Data from the Loss-of-Fluid Test (LOFT) Program help quantify the margin of safety inherent in pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). As early as 1979, questions arose concerning the accuracy of LOFT fuel rod cladding temperature data during several large-break LOCA experiments. This report analyzes how well externally-mounted fuel rod cladding thermocouples in LOFT accurately reflected actual cladding surface temperature during large-break LOCA experiments. In particular, the validity of the apparent core-wide fuel rod cladding quench exhibited during blowdown in LOFT Experiments L2-2 and L2-3 is studied. Also addressed is the question of whether the externally-mounted thermocouples might have influenced cladding temperature. The analysis makes use of data and information from several sources, including later, similar LOFT Experiments in which fuel centerline temperature measurements were made, experiments in other facilities, and results from a detailed FRAP-T6 model of the LOFT fuel rod. The analysis shows that there can be a significant difference (referred to as bias) between the surface-mounted thermocouple reading and the actual cladding temperature, and that the magnitude of this bias depends on the rate of heat transfer between the fuel rod cladding and coolant. The results of the analysis demonstrate clearly that a core-wide cladding quench did occur in Experiments L2-2 and L2-3. Further, it is shown that, in terms of peak cladding temperature recording during LOFT large-break LOCA experiments, the mean bias is 11.4 {plus_minus} 16.2K (20.5 {plus_minus} 29.2{degrees} F). The best-estimate value of peak cladding temperature for LOFT LP-02-6 is 1,104.8 K. The best-estimate peak cladding temperature for LOFT LP-LB-1 is 1284.0 K.

  14. 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

  15. 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

  16. Multi-dimensional LOCA Analysis, Status and Perspective

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    The most extensive research project for LOCA was the 2D/3D program. It involved large or full size experiments such as CCTF (Cylindrical Core Test Facility), SCTF (Slab Core Test Facility) and UPTF (Upper Plenum Test Facility) as well as the most up-to-date two-fluid analysis code, TRAC. With the background of 2D/3D study, the topic of this paper is to investigate facts concerning LOCA application of the present day up-to-date code systems such as RELAP5, TRAC, CATHARE and COBRA-TF. Especially, focus has been put on the multi-dimensional phenomena. The rigorous conservative form of COBRA-TF is strongly recommended to handle the multi-dimensional multi-phase flow phenomena in the future. It is inevitable to develop the reasonable correlation for the covariance coefficients.

  17. 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.

  18. Parametric study on effect of break size during LOCA on thermal hydraulic conditions in an indian pressurized heavy water reactor (220 MWe)

    Energy Technology Data Exchange (ETDEWEB)

    Rao, G.S.; Gupta, S.K.; Raj, V.V. [Bhabha Atomic Research Centre, Mumbai (India)

    1999-07-01

    Loss Of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures. Coolant expulsion rates during LOCA are dictated by critical flow conditions governed by initial plant conditions prior to the accident, break geometry, location of break, etc. In addition the PHWRs have positive void-coefficient of reactivity for coolant resulting in reactor power rise in earlier part of LOCA, when the stored heat of the fuel has yet not been removed. If, in addition, heat transfer to the coolant drops sharply very high fuel surface temperatures are expected. The paper describes analyses carried out for three different break sizes. (author)

  19. 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.

  20. 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)

  1. 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

  2. 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)

  3. 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)

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    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)

  20. 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)

  1. 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

  2. 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)

  3. 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.

  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. 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)

  6. 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).

  7. Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage.

  8. 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)

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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。

  16. 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

  17. 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

  18. 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)

  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. 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)

  1. Automatic Prosodic Break Detection and Feature Analysis

    Institute of Scientific and Technical Information of China (English)

    Chong-Jia Ni; Ai-Ying Zhang; Wen-Ju Liu; Bo Xu

    2012-01-01

    Automatic prosodic break detection and annotation are important for both speech understanding and natural speech synthesis.In this paper,we discuss automatic prosodic break detection and feature analysis.The contributions of the paper are two aspects.One is that we use classifier combination method to detect Mandarin and English prosodic break using acoustic,lexical and syntactic evidence.Our proposed method achieves better performance on both the Mandarin prosodic annotation corpus — Annotated Speech Corpus of Chinese Discourse and the English prosodic annotation corpus —Boston University Radio News Corpus when compared with the baseline system and other researches' experimental results.The other is the feature analysis for prosodic break detection.The functions of different features,such as duration,pitch,energy,and intensity,are analyzed and compared in Mandarin and English prosodic break detection.Based on the feature analysis,we also verify some linguistic conclusions.

  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. 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

  5. Design basis neutronics calculations for NRU-LOCA experiments

    Energy Technology Data Exchange (ETDEWEB)

    Heaberlin, S.W.; Jenquin, U.P.; McNair, G.W.; Perry, R.T.; Trapp, T.J.; Zimmerman, M.G.

    1979-08-01

    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.

  6. 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

  7. 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

  8. 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)

  9. 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.

  10. 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)

  11. 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.

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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)

  17. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 4: External Pressurizer Surge Line Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  18. 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

  19. 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)

  20. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  1. 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.

  2. 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.

  3. 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)

  4. 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

  5. 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.

  6. 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.

  7. 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

    OpenAIRE

    鈴木 光弘; 竹田 武司; 中村 秀夫

    2010-01-01

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility of ROSA-V Program to have an insight into effects of accident management action on core cooling during a simulated vessel top break loss-of-coolant accident with a total failure assumption on the high pressure injection (HPI) system at a pressurized water reactor (PWR). Typical phenomena of vessel top break with break sizes between 1.0 and 0.1% cold leg break equivalent were clarifie...

  8. 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.)

  9. Break WEP Faster with Statistical Analysis

    OpenAIRE

    Chaabouni, Rafik

    2006-01-01

    The Wired Equivalent Protocol is nowadays considered as unsafe. However the only academic research that tries to break WEP has been done by Fluhrer, Mantin and Shamir, who have published a report on a specific attack. Nevertheless, an unknown person under the pseudonym Korek has published 17 attacks, which are now used by both AirCrack and WepLab. For a network with average load traffic, the FMS attack would need roughfly 40 days in order to find the key (4 millions packets needed), whereas K...

  10. 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.

  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. 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

  13. 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。

  14. 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)

  15. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Potapov, S. [Electricite de France (EDF-SEP / AMV), 92 - Clamart (France); Tephany, F. [Electricite de France (EDF SEPTEN), 69 - Villeurbanne (France)

    2001-07-01

    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)

  17. 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)

  18. 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)

  19. 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

  20. 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)

  1. 中小破口失水事故现实估算分析%Realistic Analysis of Intermediate and Small LOCA

    Institute of Scientific and Technical Information of China (English)

    余红星; 廖业宏

    2002-01-01

    介绍了以CATHARE和SAHASB计算机程序为基础的中小破口失水事故现实估算方法.在大亚湾1 8个月换料项目中,为了定义失水事故(LOCA)包络线和检查安全裕量,运用此方法进行了计算分析.结果表明,大亚湾核电站采用18个月换料之后,在中小破口失水事故时仍有较大安全裕量.

  2. 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)

  3. Moderator Circulation Simulation for 35% Reactor Inlet Header Break in the Wolsong Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    The objective of this study is present the results of moderator circulation simulations by the CFD code MODTURC{sub C}LAS V2.9-IST for the refurbished Wolsong unit 1. The present simulations were performed for a loss of Class IV power during a Large Break Loss Of Coolant Accident (35% inlet header break) without ECC injection and steam generator crash cool-down (LOCA/LOECC/LOCC). The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available for the inlet header break scenario

  4. 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

  5. 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.

  6. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Bolt, S.E.

    1977-11-04

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions.

  7. 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

  8. 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.

  9. 现实LOCA分析中不确定性量化分析方法研究%Investigation on Uncertainty Quantification Method in Realistic LOCA Analysis

    Institute of Scientific and Technical Information of China (English)

    林支康; 王婷; 林建树; 梁任; 卢向晖

    2016-01-01

    In this paper,we introduce many kinds of uncertainty quantification methods used in realistic LBLOCA analysis,and quantify the uncertainties of the output of large break loss of coolant accident based on the best estimate thermal hydraulic system code CATHARE GB LBLOCA model of CPR1000 nuclear power plant.Furthermore,we compare and identify the differences in many areas such as the process of input and output parameters and analysis procedure,and recognize that the sensitivity study method is most conservative because of the existing of the artificial conservative assumption,and the normal distribution test is required in the conventional parameter statistic method,Owen factor method and Bootstrap method,but not in Wilks method,and Wilks method can achieve more realistic results.%基于中国改进型三环路压水堆(CPR1000)核电厂最佳估算热工水力系统分析程序(CATHAREGB)大破口失水事故(LBLOCA)分析模型,采用多种不确定性量化分析方法量化LBLOCA分析结果的不确定性,并针对不同方法对输入参数的处理、分析过程和输出结果的处理3个方面进行比较.结果表明,敏感性分析方法由于存在人为的保守性,所以其计算结果最保守,传统的参数统计方法、欧文因子法和Bootstrap方法需进行输出结果的正态性检验,而Wilks方法不需要进行正态性检验且分析结果更加现实.

  10. 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

  11. 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) ...

  12. 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)

  13. ANALYSIS OF MEN'S AND WOMEN'S BASKETBALL FAST-BREAKS

    Directory of Open Access Journals (Sweden)

    Ignacio Refoyo

    2009-01-01

    opposition. For men, some dependence relationships were found between the fast break result and the following variables: duration, completion area, and opposition to its completion. For women, the results revealed a weak association between the fast break result and the opposition to its completion.

  14. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  15. 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.

  16. 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...

  17. 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)

  18. 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

  19. An analysis of dam-break flow on slope

    Institute of Scientific and Technical Information of China (English)

    潘存鸿; 王立辉

    2014-01-01

    The one-dimensional steep slope shallow water equations are used to model the dam-break flow down a uniform slope with arbitrary inclination, and analytical solutions are derived by the hodograph transformation and the Riemann’s method in terms of evaluated integrals. An implicit analytical solution is obtained to evaluate the spatio-temporal distributions of dam-break flood hydrographs along the slope. For convenience, the solution for representative wave profiles and velocity distributions is shown in charts. Comparing with the Dressler’s solution and WES experimental data, the analytical solution is seen reasonable.

  20. 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...

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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

  7. An analysis of gamma-ray burst spectral break models

    CERN Document Server

    Zhang, B; Zhang, Bing; Meszaros, Peter

    2002-01-01

    Typical gamma-ray burst spectra are characterized by a spectral break, Ep, which for bright BATSE bursts is found to be narrowly clustered around 300 keV. Recently identified X-ray flashes, which may account for a significant portion of the whole GRB population, seem to extend the Ep distribution to a broader range below 40 keV. On the other hand, within the cosmological fireball model, the issues concerning the dominant energy ingredient of the fireball as well as the location of the GRB emission site are not unambiguously settled, leading to several variants of the fireball model. Here we analyze these models within a unified framework, and critically reexamine the Ep predictions in the various model variants, focusing on their predicted properties. Attention is focused on the ability of the models to match a narrowness of the Ep distribution, and the correlations among Ep and some other measurable observables, as well as the effect of extending these properties to X-ray flash sources. These model propertie...

  8. 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...

  9. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

    Pestana Barros, Carlos, E-mail: cbarros@iseg.utl.p [Instituto Superior de Economia e Gestao and Research Unit on Complexity and Economics, Technical University of Lisbon, Lisbon (Portugal); Gil-Alana, Luis A., E-mail: alana@unav.e [University of Navarra, Pamplona (Spain); Payne, James E., E-mail: jepayne@ilstu.ed [Department of Economics, Illinois State University, Normal, IL 61790-4200 (United States)

    2011-01-15

    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: {yields}Mean reverting persistence in oil production with breaks identified in 10 out of the 13 countries examined. {yields} Standard analysis based on cointegration techniques and involving oil production should be examined in the more general context of fractional cointegraton. {yields} Analysis of outliers did not alter the main conclusions of the study.

  11. 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.

  12. 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.)

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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...

  18. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Barros, Carlos Pestana [Instituto Superior de Economia e Gestao and Research Unit on Complexity and Economics, Technical University of Lisbon, Lisbon (Portugal); Gil-Alana, Luis A. [University of Navarra, Pamplona (Spain); Payne, James E. [Department of Economics, Illinois State University, Normal, IL 61790-4200 (United States)

    2011-01-15

    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. (author)

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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)

  7. Radiation signature folowing the hypothesized LOCA. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.

    1977-09-01

    The study establishes the radiation source profile following the hypothesized Loss of Coolant Accident (LOCA) as suggested by the applicable Regulatory Guides. The source is specified as time-dependent gamma and beta energy release rates and energy spectra with dose and dose rate values presented for a generic containment structure. The results of the study will provide a basis for a comparison of radiation simulators used in (radiation) qualification testing of Class I components and an evaluation of simulator ''adequacy'' in duplicating the LOCA radiation environments and resultant component damage.

  8. 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...

  9. 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.

  10. 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

  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. Analysis of Steam Line Break for the Development of KNGR EOG

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, H.Y.; Lee, S.R. [Korea Electric Power Research Institute, Taejon (Korea)

    2000-02-01

    The steam line break accidents are analyzed for the development of KNGR emergency operation guidelines (EOG). Realistic plant initial conditions and assumptions are applied because it is important to evaluate the best-estimate plant behavior for the development of EOG. The steam line break analysis is required to prepare the Excess Steam Demand Event (ESDE) of the Optimal Recovery Guidelines (ORG). Five cases of SLBFP, SLBFPLOP, SLBZP, SLPFPNSI, and ZP2RCP are analysed considering the power level, plant conditions, and possible operator actions. The results show that the accident can be mitigated safely and the requirements described in the guidelines are satisfied when the plant is controlled following the provided procedures . (author). 3 refs., 70 figs.

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  2. 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

  3. 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

  4. 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

  5. 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...

  6. 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.

  7. System-level validation of CATHENA MOD-3.5D for early blowdown phase of large LOCA - RD-14M tests B0405-B0413

    Energy Technology Data Exchange (ETDEWEB)

    Gu, J.W.; McGee, G.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: guj@aecl.ca

    2008-07-01

    To investigate the integrated effect of multiple phenomena on CATHENA MOD-3.5d code uncertainty, for the early blowdown phase of large loss of coolant accident (LOCA), one RD-14M test series (B0405-B0413) is used to perform a system-level validation. The peak sheath temperature in the Fuel-Element-Simulator (FES) is selected as the key output parameter used to quantify the code bias and uncertainty in the validation. In the nine tests, the test conditions (break size, pump and power trip time, fluid sub-cooling and pressurizer isolation) are systematically varied and simulated, so that their effects on the magnitude and timing of the peak FES-sheath temperatures are demonstrated. The base test, B0405 is selected to perform sensitivity and uncertainty analyses. The sensitivity analyses show that the choice of film-boiling heat-transfer correlation has a significant effect on the prediction of the FES-sheath temperatures during the FES quenching period. Uncertainty analysis demonstrates a mean bias of about +20{sup o}C, with a range of about {+-}30{sup o}C to the upper and lower bounds. These results compare very well with the estimated code accuracy based on all nine tests of B0405-B0413. (author)

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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…

  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. Mechanical Analysis of Taekwondo Power Break%跆拳道威力击破的力学分析

    Institute of Scientific and Technical Information of China (English)

    张智

    2013-01-01

      文章运用文献资料法、逻辑分析法等研究方法,阐明了跆拳道威力击破的相关力学原理,并运用相关的力学原理对威力击破的全过程进行科学的力学分析,为跆拳道的威力击破提供理论性依据,更好的把握威力击破的本质,达到最佳的击破效果。%This paper uses the methods of literature review, logical analysis to clarify the mechanics principles of Taekwondo power break, the mechanical principle of power break, the whole process of the scientific analysis of the mechanics, to provide a theoretical basis for Taekwondo power break, more good grasp of the essence of power break, to achieve the best break.

  1. 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)

  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. 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

  4. 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.

  5. 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.

  6. 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

  7. 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)

  8. 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)

  9. 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.

  10. 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.

  11. Two-phase flow across a partially damaged core during the reflood phase of a loca

    Energy Technology Data Exchange (ETDEWEB)

    Ruyer, P., E-mail: pierre.ruyer@irsn.fr [IRSN PSN/SEMIA/LIMAR, B.P. 3, 13 115 St-Paul-Lez-Durance Cedex (France); Seiler, N.; Biton, B.; Lelong, F.; Secondi, F.; Baalbaki, D. [IRSN PSN/SEMIA/LIMAR, B.P. 3, 13 115 St-Paul-Lez-Durance Cedex (France); Gradeck, M. [LEMTA Nancy University CNRS Vandoeuvre les Nancy (France)

    2013-11-15

    This study focuses on thermal-hydraulic simulations, at sub-channel scale, of a damaged PWR reactor core during a Loss Of Coolant Accident (LOCA). The aim of this study is to simulate the thermal-hydraulics to provide the thermal-mechanical code DRACCAR with an accurate wall heat transfer law. This latter code is developed by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to evaluate the thermics and deformations of fuel assemblies within the core. The present paper first describes the use of CFD to perform analysis at sub-channel scale. Then we study the capabilities of existing codes CATHARE-3 and CESAR to simulate dispersed droplet flows.

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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)

  17. 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.

  18. Effect of thermal aging on the leak-before-break analysis of nuclear primary pipes

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Xuming; Li, Shilei [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China); Wang, Xitao [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China); Collaborative Innovation Center of Steel Technology, University of Science and Technology Beijing, Beijing 100083 (China); Wang, Yanli [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China); Wang, Zhaoxi [CPI Nuclear Power Institute, 18 Xizhimen St., Beijing 100044 (China); Xue, Fei [Suzhou Nuclear Power Research Institute, Suzhou 215004 (China); Zhang, Hailong, E-mail: hlzhang@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China)

    2014-12-15

    Highlights: • Thermal aging embrittlement is considered in LBB assessment of nuclear pipe. • Effect of thermal aging on growth behavior of partial-through crack is not obvious. • Detectable leakage crack length of thermally aged material is slightly increased. • Critical crack length of thermally aged material is significantly reduced. • Ignorance of thermal aging produces less conservative LBB results. - Abstract: Three-dimensional finite element analysis (FEA) models were built for pipes with circumferential cracks and the effect of thermal aging embrittlement on the leak-before-break (LBB) behavior was analyzed according to the Level 2 and Level 3 safety assessments. The detectable leakage crack length obtained using the two-phase critical flow model and the critical crack length calculated by the J-integral stability assessment diagram method were carried out to assess the LBB behavior. The propagation behavior of partial-through circumferential cracks for both unaged and thermally aged materials was estimated by testing fatigue crack growth rate. The results show that the effect of thermal aging on detectable leakage crack length is not obvious, whereas the critical crack length after thermal aging significantly decreases due to degradation of fracture toughness. The increments of partial-through cracks are insignificant after 40 years of service. In the Level 2 and Level 3 safety assessments for nuclear piping, LBB is shown to have sufficient safety margins, while it is suggested to decrease in the case of thermal aging. This work demonstrates that less conservative LBB assessment results will be produced if thermal aging embrittlement in piping steels is not taken into consideration.

  19. Radiative transfer during the reflooding step of a LOCA

    Science.gov (United States)

    Gérardin, J.; Seiler, N.; Ruyer, P.; Boulet, P.

    2013-10-01

    Within the evaluation of the heat transfer downstream a quench front during the reflood phase of a Loss of Coolant Accident (LOCA) in a nuclear power plant, a numerical study has been conducted on radiative transfer through a vapor-droplet medium. The non-grey behavior of the medium is obvious since it can be optically thin or thick depending on the wavelength. A six wide bands model has been tested, providing a satisfactory accuracy for the description of the radiative properties. Once the radiative properties of the medium computed, they have been introduced in a model solving the radiative heat transfer based on the Improved Differential Approximation. The fluxes and the flux divergence have been computed on a geometry characteristic of the reactor core showing that radiative transfer plays a relevant role, quite as important as convective heat transfer.

  20. THE RELATIONSHIP BETWEEN ENERGY CONSUMPTION AND ECONOMIC GROWTH: EVIDENCE FROM A STRUCTURAL BREAK ANALYSIS FOR TURKEY

    Directory of Open Access Journals (Sweden)

    Yasemin Dumrul

    2013-01-01

    Full Text Available In this study the aim was to investigate empirically the role of energy consumption in economic growth for the Turkish economy. The data used include annual energy consumption and economic growth series from 1960 to 2008. We used aggregate as well as various disaggregate data on energy consumption, including, oil, electricity, coal and renewable energy. Our contribution is that we take a structural breaks modeling approach in this paper. In the literature, the Kejriwal cointegration test has not been applied to date. The main conclusion of the study was that Turkey’s energy consumption and economic growth has a positive relationship varying quantity with structural breaks.

  1. Break It

    Institute of Scientific and Technical Information of China (English)

    MATTHEW PLOWRIGHT; GWYNN GUILFORD

    2008-01-01

    @@ Resolutions are not natural - otherwise you wouldn't have to "resolve" to execute them. This year, instead of planning how to commit to a slew of unattainable goals, why not prepare for breaking your resolutions the right way?

  2. Modeling the cool down of the primary heat transport system using shut down cooling system in normal operation and after events such as LOCA

    International Nuclear Information System (INIS)

    This paper aims at modeling the cooling of the primary heat transport system using shutdown cooling system (SDCS), for a CANDU 6 NPP in all operating modes, normal and abnormal (particularly in case of LOCA accident), using the Flowmaster calculation code. The modelling of heavy water flow through the shutdown cooling system and primary heat transport system was performed to determine the distribution of flows, pressure in various areas of the hydraulic circuit and the pressure loss corresponding to the components but also for the heat calculation of the heat exchangers related to the system. The results of the thermo-hydraulic analysis show that in all cases analyzed, normal operation and for LOCA accident regime, the performance requirements are confirmed by analysis

  3. 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…

  4. 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.

  5. Sump Pool Flow Simulation during Fill-up Phase of LOCA Using on CFD for OPR1000 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Kyung Sik; Park, Jong Pil; Joeng, Ji Hwan [Pusan National University, Busan (Korea, Republic of); Kim, Man Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-10-15

    During LOCA (Loss of Coolant Accident) in design bases accident (DBA), emergency core coolant supplements form a recirculation sump and cooled core and containment. When the double ended guillotine Break (DEGB) at the hot leg near steam generator, due to the jet impingement discharge flow, the debris could be potentially generated at pipe or wall nearby steam generator and be transported to the recirculation sump. Therefore, the debris, such as insulations and paint chips, could be accumulated and be clogged in the recirculation sump screen. If debris is blocked the sump strainer, the pressure drop is increased at the screen so as to increase the pressure loss of ECCS (Emergency Core Cooling System) pump NPSH (Net positive suction head). It is potentially influenced to decrease the long-term cooling capability of the recirculation sump. The recirculation sump screen clogging accident has happened in BWR of USA and Sweden. Considering the important to safety, US NRC (Nuclear Safety Commission) has issued the recirculation sump blockage as GSI-191(Generic Safety Issue-191). Moreover, US NRC published Regulatory Guide 1.82 Rev.3 incorporated the R and D findings and experiences in 2003. NEI (Nuclear Energy Institute) introduced the methodology procedure to solve this safety issue in the NEI 04-07 report. In the meanwhile, US NRC also published individually the regulatory guidelines as a SER (Safety Evaluation Report) report for PWR plant. However, the current available technical information including the reports is applicable to the generic PWR plants not the plant specific plant. Therefore, the additional research reflecting characteristics of plant specific plant is necessary to develop the methodology and technical guides on the recirculation sump clogging issue. The objective of this study is addressed to explore the characteristics of sump pool flow during LOCA by using CFD for the OPR1000 plant.

  6. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Bang, Je Geon; Kim, Dae Ho; Yang, Yong Sik; Song, Keun Woo

    2007-12-15

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced.

  7. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    International Nuclear Information System (INIS)

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced

  8. Fuel Behaviour and Modelling under Severe Transient and Loss of Coolant Accident (LOCA) Conditions. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    the art of the performance of nuclear fuel for water cooled reactors under severe transients and LOCA conditions. The meeting was attended by 83 specialists representing fuel vendors, nuclear utilities, research and development institutions, and regulatory authorities from 19 Member States. The papers submitted to the meeting were organized into seven sessions covering analytical and experimental RIA and LOCA studies and international programmes, power ramp, and severe accident analysis. These proceedings contain all the papers that were presented and discussed during the meeting, and highlight key findings and recommendations based on the summaries of the session chairpersons. While the Fukushima Daiichi accident influenced the discussions, it was not directly considered because of the lack of fuel behaviour data available at the time of the technical meeting

  9. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Directory of Open Access Journals (Sweden)

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  10. The Break

    DEFF Research Database (Denmark)

    Strand, Anete Mikkala Camille

    2016-01-01

    storytelling to enact fruitful breakings of patterns unbecoming. The claim being, that the hamster wheel of Work-life anno 2016 needs reconfiguration and the simple yet fruitful manner by which this is done is through acknowledging the benefits of bodies, spaces and artifacts – and the benefits of actually...... taking a break, discontinuing for a moment in order to continue better, wiser and more at ease. Both within and as part of the daily routines, and – now and then – outside these routines in the majesty of nature with time to explore and redirect the course of life in companionships with fellow man...

  11. 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.

  12. Short term hydrogen generation following LOCA and loss of ECCS

    International Nuclear Information System (INIS)

    The purpose of the present study is to estimate the amount of hydrogen that can be generated due to metal water reaction following LOCA and loss of ECCS in a 500 MWe PHWR. A computer code HYGEN (Hydrogen Generation) written in FORTRAN calculates time-dependent fuel temperature during the post blowdown period and the amount of hydrogen generated as a result of metal water reaction. It is seen from the analyses that metal water reaction depends on fuel bundle power, its initial temperature and steam flow conditions. At present, four groups of channels have been analysed for different steam flow conditions, and it is found that, for an about 5 gm/sec steam flow condition, the maximum of amount of hydrogen is generated (5.76 x 104 gm-mole) due to the zircaloy - steam reaction. This amount of hydrogen, when considered mixed in volume V1 (drywell) of the reactor building, means that the global concentration reaches about 2.76% by volume. So, it is seen that in the short term, the global hydrogen concentration in the reactor building is well below the flammability limit of 4% by volume. (author) 4 refs., 1 tab., 10 figs

  13. Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

    Directory of Open Access Journals (Sweden)

    Lázaro Lima

    2011-12-01

    Full Text Available “Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

  14. Pre-test analysis of an integral effect test facility for thermal-hydraulic similarities of 6 inches coldleg break and DVI injection line break using MARS-1D

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae Soon; Choi, Ki Yong; Park, Hyun Sik; Euh, Dong Jin; Baek, Won Pil [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    A pre-test analysis of a small-break loss-of-coolant accident (SBLOCA, DVI Line break) has been performed for the integral effect test loop of Korea Atomic Energy Research Institute (Korea Atomic Energy Research Institute-ITL), the construction of which will be started soon. The Korea Atomic Energy Research Institute-ITL is a full-height and 1/310 volume-scaled test facility based on the design features of the APR1400 (Korean Next Generation Reactor). This paper briefly introduces the basic design features of the Korea Atomic Energy Research Institute-ITL and presents the results of pre-test analysis for a postulated cold leg SBLOCA and DVI line break. Based on the same control logics and accident scenarios, the similarity between the Korea Atomic Energy Research Institute-ITL and the prototype plant, APR1400, is evaluated by using the MARS code, which is a multi-dimensional best-estimate thermal hydraulic code being developed by Korea Atomic Energy Research Institute. It is found that the Korea Atomic Energy Research Institute-ITL and APR 1400 have similar thermal hydraulic responses against the analyzed SBLOCA and DVI Line break scenario. It is also verified that the volume scaling law, applied to the design of the Korea Atomic Energy Research Institute-ITL, gives a reasonable results to keep a similarity with APR1400. 11 refs., 19 figs., 3 tabs. (Author)

  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. 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.

  17. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  20. 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)

  1. 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...

  2. 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

  3. 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。

  4. 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

  5. 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

  6. Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong Kwan, E-mail: jksuh@khnp.co.kr [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kim, Jae Won; Kwon, Sun Guk; Lee, Jae Yong [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Cho, Hyoung Kyu; Park, Goon Cherl [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

    2015-08-15

    Highlights: • In-vessel downstream effect tests were performed in the presence of LOCA-generated debris. • Available driving heads under each LOCA scenario were verified using experimental data. • Fibrous debris was prepared to satisfy the length distribution obtained from the bypass test. • Limiting test conditions were identified through sensitivity studies. - Abstract: Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

  7. Analysis of the electron component of EAS at observational level 700 g x cm(-2) with a scale breaking interaction model and gammaisation hypothesis

    Science.gov (United States)

    Procureur, J.; Stamenov, J. N.; Stavrev, P. V.; Ushev, S. Z.

    1985-01-01

    Scale breaking model and gammaisation processes for high energies give a correct description of the longitudinal development of extensive air showers (E.A.S.). From the analysis of phenomenological characteristics of E.A.S. at Tien-Shan experiment, it follows that for energies near 10 to the 6 GeV the secondary particle multiplicity increases with energy faster than is predicted by the accepted scale breaking model.

  8. Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs; Comparacion de la fenomenologia de las secuencias de SBO con y sin LOCA de sellos en reactores PWR Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

    2013-07-01

    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. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    International Nuclear Information System (INIS)

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m2, fluence 3 MW-y/m2, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass

  10. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    Energy Technology Data Exchange (ETDEWEB)

    Cambi, G. [Bologna Univ. (Italy). Physics Dept.; Cepraga, D.G. [ENEA, Bologna (Italy). Innovation Dept.; Di Pace, L. [ENEA, Frascati (Italy). Fusion Sector CR di Frascati; Porfiri, M.T. [ENEA, Frascati (Italy). Fusion Dept.

    1995-12-31

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m{sup 2}, fluence 3 MW-y/m{sup 2}, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass.

  11. The three-dimension model for the rock-breaking mechanism of disc cutter and analysis of rock-breaking forces

    Institute of Scientific and Technical Information of China (English)

    Zhao-Huang Zhang; Fei Sun

    2012-01-01

    To study the rock deformation with three-dimensional model under rolling forces of disc cutter,by carrying out the circular-grooving test with disc cutter rolling around on the rock,the rock mechanical behavior under rolling disc cutter is studied,the mechanical model of disc cutter rolling around the groove is established,and the theory of single-point and double-angle variables is proposed.Based on this theory,the physics equations and geometric equations of rock mechanical behavior under disc cutters of tunnel boring machine (TBM) are studied,and then the balance equations of interactive forces between disc cutter and rock are established.Accordingly,formulas about normal force,rolling force and side force of a disc cutter are derived,and their validity is studied by tests.Therefore,a new method and theory is proposed to study rock- breaking mechanism of disc cutters.

  12. 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

  13. Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10. Final report

    International Nuclear Information System (INIS)

    This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases and conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff approved acceptance criteria for LOCA-related hydrodynamic loads are provided in an appendix

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

    International Nuclear Information System (INIS)

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

  15. Breaking Routines

    DEFF Research Database (Denmark)

    Kesting, Peter; Jørgensen, Frances

    On some level, innovation begins when the current way of doing things is questioned and alternatives are sought. In cognitive terms, this can be conceptualized as the point at which an agent breaks with existing routine and returns to planning and decision-making. Thus far, however, very little...... is known about this cognitive structure or the factors that trigger the search for alternatives. In cooperation with the Danish Research Centre for Magnetic Resonance, University of Copenhagen, Denmark, we are in the process of designing an experimental study designed to gain insights into the triggers...

  16. Theory and Application of Loss of Life Risk Analysis for Dam Break

    Institute of Scientific and Technical Information of China (English)

    SUN Yuefeng; ZHONG Denghua; LI Mingchao; LI Ying

    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.Loss of life is 24 220 when the warning time is0.25-1 h and flood severity understanding is vague,which indicates that the risk is intolerable.The losses of life under three other conditions are tolerable: warning time 0.25-1 h,and precise flood severity understanding; warning time more than 1 h,and vague flood severity understanding; warning time more than 1 h,and precise flood severity understanding.

  17. 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

  18. Kinetic analysis of Yersinia pestis DNA adenine methyltransferase activity using a hemimethylated molecular break light oligonucleotide.

    Directory of Open Access Journals (Sweden)

    Robert J Wood

    Full Text Available BACKGROUND: DNA adenine methylation plays an important role in several critical bacterial processes including mismatch repair, the timing of DNA replication and the transcriptional control of gene expression. The dependence of bacterial virulence on DNA adenine methyltransferase (Dam has led to the proposal that selective Dam inhibitors might function as broad spectrum antibiotics. METHODOLOGY/PRINCIPAL FINDINGS: Herein we report the expression and purification of Yersinia pestis Dam and the development of a continuous fluorescence based assay for DNA adenine methyltransferase activity that is suitable for determining the kinetic parameters of the enzyme and for high throughput screening against potential Dam inhibitors. The assay utilised a hemimethylated break light oligonucleotide substrate containing a GATC methylation site. When this substrate was fully methylated by Dam, it became a substrate for the restriction enzyme DpnI, resulting in separation of fluorophore (fluorescein and quencher (dabcyl and therefore an increase in fluorescence. The assays were monitored in real time using a fluorescence microplate reader in 96 well format and were used for the kinetic characterisation of Yersinia pestis Dam, its substrates and the known Dam inhibitor, S-adenosylhomocysteine. The assay has been validated for high throughput screening, giving a Z-factor of 0.71+/-0.07 indicating that it is a sensitive assay for the identification of inhibitors. CONCLUSIONS/SIGNIFICANCE: The assay is therefore suitable for high throughput screening for inhibitors of DNA adenine methyltransferases and the kinetic characterisation of the inhibition.

  19. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  20. 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)

  1. Verification of human actions in SBO sequences with LOCA stamps in Westinghouse PWRs; Verificacion de las actuaciones humanas en secuencias de SBO con LOCA de sellos en reactores PWR Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Mena Rosell, L.; Jimenez Varas, G.

    2013-07-01

    The Fukushima accident has shown the need for tools and methodologies able to analyze human activities and / or capabilities of portable systems that has given the Spanish plants as a result of the stress tests . In this work we have applied the methodology of integrated safety analysis developed by the CSN , to SBO sequences with LOCA stamp. The aim is to show a methodology for testing the performances of the Emergency Operating Procedures and Guides Severe Accident Management. The simulations were performed with the tool SCAIS coupled to MAAP . The results show that there are human activities that may be beneficial in certain sequences but harmful in others. This type of problem is already known and referred to in the GGAS . However, FSR shows a practical way to check human actions cannot be obtained with other methods.

  2. 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...

  3. Assessing the impact of the dispersion of fuel in case of LOCA; Evaluacion del impacto de la dispersion de combustible en caso de LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Concejal, A.; Garcia Sedano, P. J.; Crespo, A.

    2013-07-01

    Recent studies conducted in Halden and Studsvik have indicated the possibility of obtaining highly fragmented fuel with relatively low temperatures (700 degree centigrade) and high burned (70 MWd / kgU). In case of accident loss of coolant (LOCA), the expulsion may occur outside the pod fuel fragments, which can affect the coolability, cause channel blockade and therefore an increase in the maximum temperature of sheath.

  4. Thermal-Hydraulic Assessment of W7-X Plasma Vessel Venting System in Case of 40 mm In-Vessel LOCA

    OpenAIRE

    E. Urbonavičius; T. Kaliatka

    2015-01-01

    This paper presents assessment of the capacity of W7-X venting system in response to in-vessel LOCA, rupture of 40 mm diameter pipe during operation mode “baking.” The integral analysis of the coolant release from the cooling system, pressurisation of PV, and response of the venting system is performed using RELAP5 code. The same coolant release rate was introduced to the COCOSYS code, which is a lumped-parameter code developed for analysis of processes in containment of the light water react...

  5. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm2, simulated with RELAP5 code

    International Nuclear Information System (INIS)

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm2-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

  6. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

  7. 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

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

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

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

  9. 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.

  10. Virginia Power's generic main steam-line-break DNBR (departure from nucleate boiling ratio) analysis

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.C.; Harrell, J.R.; Erb, J.O.

    1990-06-01

    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.

  11. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  12. Fitness for service after a LOCA: A process applied to Pickering NGS Unit 2

    International Nuclear Information System (INIS)

    The fitness for service process provides a unique proven methodology for assessing and correcting post-LOCA damage, essential to plant restart. The process uses the as-built plant configuration for modelling input and features self correcting feedback from inspection to validate assessment models. This paper focuses on the process steps and the infrastructure necessary to execute the process

  13. Downscaling humidity with Localized Constructed Analogs (LOCA) over the conterminous United States

    Science.gov (United States)

    Pierce, D. W.; Cayan, D. R.

    2016-07-01

    Humidity is important to climate impacts in hydrology, agriculture, ecology, energy demand, and human health and comfort. Nonetheless humidity is not available in some widely-used archives of statistically downscaled climate projections for the western U.S. In this work the Localized Constructed Analogs (LOCA) statistical downscaling method is used to downscale specific humidity to a 1°/16° grid over the conterminous U.S. and the results compared to observations. LOCA reproduces observed monthly climatological values with a mean error of ~0.5 % and RMS error of ~2 %. Extreme (1-day in 1- and 20-years) maximum values (relevant to human health and energy demand) are within ~5 % of observed, while extreme minimum values (relevant to agriculture and wildfire) are within ~15 %. The asymmetry between extreme maximum and minimum errors is largely due to residual errors in the bias correction of extreme minimum values. The temporal standard deviations of downscaled daily specific humidity values have a mean error of ~1 % and RMS error of ~3 %. LOCA increases spatial coherence in the final downscaled field by ~13 %, but the downscaled coherence depends on the spatial coherence in the data being downscaled, which is not addressed by bias correction. Temporal correlations between daily, monthly, and annual time series of the original and downscaled data typically yield values >0.98. LOCA captures the observed correlations between temperature and specific humidity even when the two are downscaled independently.

  14. Loss-of-coolant accident (LOCA) simulation tests on polymers: the importance of including oxygen

    International Nuclear Information System (INIS)

    Experiments were performed to survey the effects on material degradation of both aging conditions and the oxygen concentration during a LOCA simulation. Changes for a number of commercial materials commonly used as electric cable jackets and insulations in nuclear power plant applications were monitored in terms of weight, mechanical properties, solubility measurements and infrared spectroscopy. For a number of these materials (an EPR insulation, a chloroprene jacket and a PVC jacket), the concentration of oxygen during LOCA simulation was found to be an important parameter. For the first two materials, more degradation occurred when oxygen was present; for PVC, substantially increased swelling occurred as the oxygen concentration was lowered. Aging conditions were also found to have a very substantial influence. In particular, for a number of the materials, lowering the radiation dose rate used for aging led to enhanced degradation after both the aging and the LOCA simulation. The different materials examined showed very different behaviors in terms of the degradation resulting from aging and from LOCA simulation

  15. Molecular analysis of the distribution of chromosomal breakpoints: characterization of a 'hot' region for breaks in human chromosome 11

    International Nuclear Information System (INIS)

    Full text: Ionizing radiation randomly damages DNA and chromosomes whereas subsequent chromosome breaks are non-random. Assuming, as an ideal and naive but useful proposition, that breaks are equally likely anywhere in the chromosome and that a deletion always occurs between two breaks, the frequency of fragments would decrease linearly with increasing fragment size. This simple distribution is not, however, observed. To shed light on the 'real' situation of break formation we mapped breakpoints in the human chromosome no. 11 of 353 independent CD59- mutants isolated from human/hamster hybrid AL cells exposed to radiations (high and low dose-rate gamma rays, high LET carbon or nitrogen ions, protons) or chemicals (arsenic or irradiated, mutagenic histidine) or unexposed. The number of breaks per unit length of DNA differed significantly in different regions of chromosome 11.The highest level of breaks (140/mbp) were in the 0.8 mbp segment between CD59 and Catalase (CAT). Finer mapping of break points was carried out using 26 PCR primer pairs spread across this interval in 15 independent mutants. In two mutants, the break point was in a 107 bp fragment; in the other 13 the breaks were in a single 35 mbp fragment, but not all were at exactly the same site; 4 of 13 occurred in 3 different 3 mbp sub-segments. We are sequencing these fragments to look for such features as repeats: 'colder' regions like that between CD59 and WT will also be analyzed. But, since at least some breaks occurred at different sites and the frequency and distribution of breaks was about the same for all treatments, our we postulate that hot (and cold spots) may be due more to structural features or specific repair than to sequence or type of damage

  16. 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.

  17. Pressurized thermal shock. CNA-I behavior when a hot leg breaks of 50 cm2 is produced

    International Nuclear Information System (INIS)

    Pressurized thermal shock (PTS) phenomena in the CNA-I pressurize heavy water reactor is analyzed in this paper. The initiating event is a hypothetical 50 cm2 break of the line connecting the pressurizer and the primary system. The calculation procedure for obtaining the local thermal-hydraulic parameters in the reactor pressure vessel downcomer is described firstly. Results obtained lead to conclusions in different subjects. The first conclusion is that a simple tool of easy application is available to analyze PTS phenomena in cases of breaks in the primary system in cold and hot legs. This methodology is fully independent of the methodology utilized by the Utility. Another important conclusion comes from the analysis of the temperature evolution of the fluid below the cold leg level in the RPV downcomer, as a function of the THPI temperature of the TJ system injected water from. It is also concluded that the results obtained with the methodology adopted agree with the ones obtained with the methodologies validated against experiments in the UPTF facility. It is possible to observe that when THPI increase, the conditions suitable for PTS occurrence in a LOCA accident tend to diminish. The maximum value to the THPI may be fixed from the maximum temperature allowed to preserve the structural integrity of the fuel cladding. (author)

  18. [Clarification of a break-in theft crime by multiplex PCR analysis of cigarette butts].

    Science.gov (United States)

    Hochmeister, M; Haberl, J; Borer, V; Rudin, O; Dirnhofer, R

    1995-01-01

    This paper describes the first use of multiplex PCR amplification kits for the analysis of DNA extracted from cigarette butts in a criminal case. Two suspects could be excluded as potential contributors to the samples, whereas the multi locus PCR-based DNa profile derived from the cigarette butts was consistent with a DNA profile derived from a third suspect. For identity testing in criminal cases where cigarette butts are involved, commercially available PCR amplification kits provide currently the most powerful tool. Furthermore this PCR-based analysis can be implemented into most application orientated laboratories.

  19. Overview of the M5R Alloy behavior under RIA and LOCA Conditions

    International Nuclear Information System (INIS)

    Experience from irradiation in PWRs has confirmed the M5R possesses all the properties required for upgraded operation including new fuel management approaches and high duty reactor operation. In this paper accident behavior is demonstrated through a comparison of M5R and Zircaloy-4 cladding behavior under RIA (Reactivity Insertion Accident) and LOCA (Loss Of Coolant Accident) conditions. AREVA NP supports a significant experimental program of analytical and full -scale tests along with comprehensive analyses on both M5R and SRA low-tin Zircaloy-4. A key presumption in the conduct of such tests is that, for all Zirconium alloys, the primary effects of high burn-up on cladding thermal-mechanical properties arise from the accumulation of hydrogen within the cladding during operation. This hypothesis is supported through a summarisation of the results of the main RIA and LOCA tests performed on virgin, pre-hydrided, and irradiated M5R and SRA low-tin Zircaloy-4 cladding. The first part of the paper presents the results of recent Room Temperature (RT) and High Temperature High Pressure (HTHP) integral RIA tests, mainly from the NSRR and CABRI programs, and separate effects mechanical properties tests on high burn-up M5R and Zircaloy- 4 irradiated claddings. In the second part of this paper, studies of cladding performance under LOCA conditions are presented.. The discussion includes high temperature oxidation kinetics, quench behaviour and post quenched mechanical behaviour of virgin, pre-hydrided and irradiated M5R and Zircaloy-4 cladding tubes after oxidation at LOCA temperatures and various quenching scenarios. The hydrogen concentrations studied are alloy dependent. Included are mechanical tests and in-depth metallurgical investigations developed to understand the failure mechanisms with the differing alloys and hydrogen concentrations. The result is a confirmation that the effect of hydrogen uptake dominates on the RIA and LOCA response of Zirconium based cladding

  20. APT Blanket System Loss-of-Coolant Analysis Based on Initial Conceptual Design - Case 2: External HR Break HR Break at Pump Outlet with Pump Trip

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    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 system. These simulations were performed for the Preliminary Safety Analysis Report.

  1. a Simplified Methodology for the Prediction of the Small Break Loss-Of Accident.

    Science.gov (United States)

    Ward, Leonard William

    1988-12-01

    This thesis describes a complete methodology which has allowed for the development of a faster than real time computer program designed to simulate a small break loss -of-coolant accident in the primary system of a pressurized water reactor. By developing an understanding of the major phenomenon governing the small break LOCA fluid response, the system model representation can be greatly simplified leading to a very fast executing transient system blowdown code. Because of the fast execution times, the CULSETS code, or Columbia University Loss-of-Coolant Accident and System Excursion Transient Simulator code, is ideal for performing parametric studies of Emergency Core Cooling system or assessing the consequences of the many operator actions performed to place the system in a long term cooling mode following a small break LOCA. While the methodology was designed with specific application to the small break loss-of-coolant accident, it can also be used to simulate loss-of-feedwater, steam line breaks, and steam generator tube rupture events. The code is easily adaptable to a personal computer and could also be modified to provide the primary and secondary system responses to supply the required inputs to a simulator for a pressurized water reactor.

  2. Proceedings of a specialist meeting on leak before break in reactor piping and vessels

    International Nuclear Information System (INIS)

    This Specialist Meeting was organised by EDF, Framatome and CEA with the participation of SFEN, and it was sponsored jointly by the CEC DG XI, Nuclear Electric, IAEA, US NRC, and by the Principal Working Group 3 (PWG-3) on Reactor Component Integrity of the NEA CSNI. The activities of PWG-3 fall into three main areas: Non-Destructive Examination (NDE), fracture analysis and aging/materials degradation. In fracture analysis, the activities are organised by the Fracture Analysis Group, and include the round robins on Fracture Analysis of Large Scale International Reference Experiments (FALSIRE). The topic of the workshop falls mainly into the second area of fracture analysis. The objective of the meeting was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. The formal proceedings of the meeting were published by US NRC as a NUREG report (NUREG/CP--0155). This includes the final versions of papers

  3. Breaks and Convergence in U.S. Regional Crime Rates: Analysis of Their Presence and Implications

    OpenAIRE

    Steve Cook; Duncan Watson

    2013-01-01

    The literature examining the relative properties of U.S. regional crime rates is extended. Using a novel method, convergence in alternative classifications of crime is detected over the period 1965 to 2009. Subsequent statistical analysis identifies distinct epochs in the evolution of crime which match those noted anecdotally in the literature. The findings concerning convergence within these epochs prove interesting, with results found to vary both between the alternative crime classificatio...

  4. Analysis of seismic disaster failure mechanism and dam-break simulation of high arch dam

    Science.gov (United States)

    Zhang, Jingkui; Zhang, Liaojun

    2014-06-01

    Based on a Chinese national high arch dam located in a meizoseismal region, a nonlinear numerical analysis model of the damage and failure process of a dam-foundation system is established by employing a 3-D deformable distinct element code (3DEC) and its re-development functions. The proposed analysis model considers the dam-foundation-reservoir coupling effect, influence of nonlinear contact in the opening and closing of the dam seam surface and abutment rock joints during strong earthquakes, and radiation damping of far field energy dissipation according to the actual workability state of an arch dam. A safety assessment method and safety evaluation criteria is developed to better understand the arch dam system disaster process from local damage to ultimate failure. The dynamic characteristics, disaster mechanism, limit bearing capacity and the entire failure process of a high arch dam under a strong earthquake are then analyzed. Further, the seismic safety of the arch dam is evaluated according to the proposed evaluation criteria and safety assessment method. As a result, some useful conclusions are obtained for some aspects of the disaster mechanism and failure process of an arch dam. The analysis method and conclusions may be useful in engineering practice.

  5. NDT techniques for strain characterization on zircaloy clad during ballooning under simulated LOCA conditions

    International Nuclear Information System (INIS)

    Zirconium alloys have been widely used as fuel cladding for most of the water reactors. Considerable attention has been given to study of its behaviour under abnormal reactor conditions such as Loss of Coolant Accident (LOCA). In order to obtain benchmark data on ballooning that could occur in such conditions in PHWR, simulated experiments are planned by heating the clad and pressurizing it internally. A system has been designed and fabricated at Atomic Fuels Division for applying different heating rates and pressures upto 100 bar on zircaloy clad. The objective of the experiments is to collect, amongst other parameters, data on ballooning behaviour and strain rate till burst. On line, in-situ measurement of strain rate accurately to assist fuel design and safety analysis has been a challenging task. Different systems have been employed for this purpose by different laboratories engaged in similar work. An extensive literature survey had revealed that various workers have adopted different techniques to suit their specific requirement which are based on the design codes followed by them. An in-depth study of the problem was carried out and several possible methods were considered. Some of them were evaluated for their possible application in our set up. Wherever required experiments were carried out for this purpose. The paper discusses the in-situ strain rate measurement techniques for Zircaloy clad and problems of each of them. Feasible techniques are further elaborated and results of experiments are described. It also compares the advantages and limitations of X-ray radiography, fluoroscopy, optical films, electronic recordings, laser scanning and Acoustic Emission (AE) methods for strain rate measurement of zircaloy. Data on experiments conducted with AE techniques are analysed. Based on the data collected, possibility of its use for in-service inspection (ISI) of zircaloy components is explored. (author). 3 refs., 4 figs

  6. Analysis of boundary point (break point) in Linear Delay Model for nanoscale VLSI standard cell library characterization at PVT corners

    CERN Document Server

    Agarwal, Gaurav Kumar

    2014-01-01

    In VLSI chip design flow, Static Timing Analysis (STA) is used for fast and accurate analysis of data-path delay. This process is fast because delay is picked from Look Up Tables (LUT) rather than conventional SPICE simulations. But accuracy of this method depends upon the underlying delay model with which LUT was characterized. Non Linear Delay Model (NLDM) based LUTs are quite common in industries. These LUT requires huge amount to time during characterization because of huge number of SPICE simulations done at arbitrary points. To improve this people proposed various other delay models like alpha-power and piecewise linear delay models. Bulusu et al proposed Linear Delay Model(LDM) which reduces LUT generation time to 50 percent. LDM divides delay curve w.r.t input rise time(trin) into two different region one is linear and other is non-linear. This boundary point between linear and non- linear region was called break point (trb). Linear region will be done if we simulate at only two points. This advantage...

  7. The technical analysis of drilling wire rope breaking%钻井钢丝绳断裂原因的技术分析

    Institute of Scientific and Technical Information of China (English)

    汪践; 秦高建

    2001-01-01

    钻机提升系统的钻井钢丝绳在油田钻井过程中突然断裂,造成设备顿钻事故。通过对钢丝绳断裂部位、断口形貌、制造质量以及游动系统滑轮参数等调查情况进行分析,剖析了钢丝绳断裂产生的原因。%The drilling wire rope of a rig's hoisting system suddenly brokeduring drilling process,cause serious damage of equipment. Through analysis of the breaking location,shape of breaking section,manufacturing quality and sheave factor of hoisting system ,the breaking reason of the drilling wire rope is discussed.

  8. Thermal-Hydraulic Assessment of W7-X Plasma Vessel Venting System in Case of 40 mm In-Vessel LOCA

    Directory of Open Access Journals (Sweden)

    E. Urbonavičius

    2015-01-01

    Full Text Available This paper presents assessment of the capacity of W7-X venting system in response to in-vessel LOCA, rupture of 40 mm diameter pipe during operation mode “baking.” The integral analysis of the coolant release from the cooling system, pressurisation of PV, and response of the venting system is performed using RELAP5 code. The same coolant release rate was introduced to the COCOSYS code, which is a lumped-parameter code developed for analysis of processes in containment of the light water reactors and the detailed analysis of the plasma vessel and the venting system is performed. Different options of coolant release modeling available in COCOSYS are compared to define the base case model, which is further used for assessment of the other parameters, that is, the failure of one burst disk, the temperature in the environment, and the pressure losses in the piping of venting system. The performed analysis identified the best option for coolant release modeling and showed that the capacity of the W7-X venting system is enough to prevent overpressure of the plasma vessel in the case of in-vessel LOCA.

  9. BEMUSE phase II report - Re-Analysis of the ISP-13 Exercise, Post Test Analysis of the LOFT L2-5 Test Calculation

    International Nuclear Information System (INIS)

    The BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme is focused on applications of the uncertainty methodologies to Large Break LOCA scenarios. The main goals of the Programme are: - To evaluate the practicability, quality and reliability of best-estimate methods including uncertainty evaluations in applications relevant to nuclear reactor safety; - To develop common understanding; - To promote / facilitate their use by the regulator bodies and the industry. The scope of the Phase II of BEMUSE is to perform Large Break LOCA analysis making reference to the experimental data of LOFT L2-5 in order to address the issue of 'the capabilities of computational tools', including the scaling / uncertainty analysis. The operational objective of the activity is the quality demonstration of the system code calculations in performing LBLOCA analysis through the fulfilment of a comprehensive set of common criteria established in correspondence of different steps of the code assessment process. In particular criteria and threshold values for selected parameters have been adopted for: a) The developing of the nodalization; b) The evaluation of the steady state results; c) The qualitative and quantitative comparison between measured and calculated time trends. Main achievements of the Phase II, to be considered in the following phases of BEMUSE, are summarized as follows: - Almost all performed calculations appear qualified against the fixed criteria; - Dispersion bands of reference results appear substantially less than in ISP-13; - The sensitivity study shall be used as guidance for deriving the uncertainty bands in the following Phase III of the Programme

  10. Does Government Debt Promote Economic Growth? An Empirical Analysis with Structural Breaks for the Economy of China

    OpenAIRE

    Stylianou Tasos

    2012-01-01

    This paper investigates the relationship between economic growth and government debt for one of the biggest economies in the world, the economy of China. The data were fitted into a regression equation using econometric techniques such as unit root tests and Granger causality. Regarding unit root tests we are using three kind of tests: i) The conventional unit tests, which do not take into account structural breaks, ii) the unit root tests that take into account one structural break and iii) ...

  11. Theoretical and numerical analysis of the hydrodynamic waves induced by dam breaks and their interaction with structures located downstream

    OpenAIRE

    Dewals, Benjamin; Erpicum, Sébastien; Archambeau, Pierre; Detrembleur, Sylvain; Pirotton, Michel

    2006-01-01

    The present paper first describes briefly the hydrodynamic model WOLF 2D, developed at the University of Liege and suitable for conducting dam break risk assessment. Secondly, the simulation of the flood induced by the Malpasset dam break enables to highlight the effectiveness of WOLF 2D. Lastly, the model is applied to demonstrate the key advantages of the most contemporary approach for reproducing the effect of buildings on the flow behaviour. The recommended modelling technique is based on...

  12. An analysis of MB-2 100% steam line break test T-2013 using RELAP5/MOD2

    International Nuclear Information System (INIS)

    This report presents RELAP5/MOD2 calculations of the 100% steam line break test T-2013 performed on the Westinghouse Model Boiler-2 facility (MB-2). The input deck uses a noding structure typical of what would be used for an integral rig or full plant study using the RELAP5/MOD2 code. Sensitivity calculations were performed for the break junction discharge coefficient and the separator drain line loss coefficient. (author)

  13. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J. [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs.

  14. Assessment of LOCA with loss of class IV power for CANDU-6 reactors using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Recently, there is an effort to improve the accuracy and reality in the transient simulation of nuclear power plants. In the prediction of the system transient, the system code simulates the system transient using the power transient curve predicted from the reactor core physics code. However, the pre-calculated power curve could not adequately predict the behavior of power distribution during transient since the coolant density change has influence on the power shape due to the change of the void reactivity. Therefore, the consolidation between the reactor core physics code and the system thermal-hydraulic code takes into consideration to predict more accurate and realistic for the transient simulation. In this regard, there are two codes are developed to assess the safety of CANDU reactor. RELAP-CANDU is a thermal-hydraulic system code for CANDU reactors developed on the basis of RELAP5/MOD3 in such a way to modify inside model for simulating the thermal-hydraulic characteristics of horizontal type reactors. SCAN (SNU CANDU-PHWR Neutronics) is a three dimensional neutronics nodal code to simulate the core physics characteristics for CANDU reactors. To couple SCAN code with RELAP-CANDU code, SCAN code was improved as a spatial kinetics calculation module in such a way to generate a SCAN DLL (dynamic linked library version of SCAN). The coupled code system, RELAP-CANDU/SCAN, enables real-time feedback calculations between thermal-hydraulic variables of RELAP-CANDU and reactor powers of SCAN. To verify the reliability of RELAP-CANDU/SCAN coupled code system, an assessment of 40% reactor inlet header (RIH) break loss of coolant accident (LOCA) with loss of Class IV power (LOP) for Wolsong Unit 2 conducted using RELAP/CANDU-SCAN coupled system. The LOCA with LOP is one of GAI (Generic Action Items) for CANDU reactors issued by CNSC (Canadian Nuclear Safety Commission) and IAEA (International Atomic Energy Agency)

  15. Reactor inlet header critical break identification and analysis for KAPP-3 and 4 using computer code RELAP-5/Mod.3.2

    International Nuclear Information System (INIS)

    Kakrapar Atomic Power Project units-3 and 4 (KAPP-3 and 4) are 700 MWe Pressurized Heavy Water Reactors (PHWR) are presently under construction. This paper presents the identification of critical break in Reactor Inlet Header and its analysis performed, for KAPP-3 and 4 as a part of safety studies to investigate the plant behavior. The limiting/critical break size at Reactor Inlet Header is identified by considering the peak sheath temperature during the Loss of coolant accident. System thermal hydraulics code RELAP-5/MOD3.2 has been used for the analysis. Here the overall thermal hydraulics of the plant along with various control systems, trip and actuation logics have been simulated. High pressure accumulators and low pressure recirculation system of emergency core cooling system are modeled. The modeling of secondary system includes modeling of Atmospheric Steam Discharge Valves (ASDVs), Safety Relief Valves (SRVs), Condensate Steam Discharge Valves (CSDVs), and Governor Valves, the U-tubes of the steam generator, the riser, the separator and the steam drum. Using this model, critical break size in the Reactor Outlet Header was identified and consequence of the event on maximum peak clad temperature and core parameters were evaluated. Following postulated accidents, the event progression and the variations of different parameters like different Header pressures, mass flow rate in the core, fuel clad temperature and rate of discharge from break etc have been studied. (author)

  16. The Effect of Protective Coating on the LOCA Simulation of Zircaloy-4 Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    In this study, a transient fuel performance code has been used to study the impact of coating the Zircaloy-4 cladding by Silicon Carbide (SiC) on the fuel performance under design basis accident conditions, particularly a loss of coolant accident (LOCA). To evaluate the effectiveness of protective coating on normal and transient fuel performance, the material properties of protective coating under irradiation has to be considered. In addition to the oxidation behavior, further studies should cover the effects of the mechanical properties, corrosion, irradiation behavior, thermal expansion, fatigue and creep of candidate protective coating materials. The preliminary transient analyses show that the protective coating on Zircaloy-4 cladding can lead to the minimization of LOCA consequences, because the steam oxidation rate of coated surface is reduced compared with that of bare Zircaloy-4 surface.

  17. Experimental and MARS code simulation of DVI line 25% and 50% break LOCAs using SNUF for APR1400

    International Nuclear Information System (INIS)

    APR1400, is an evolutionary PWR (Pressurized Water Reactor) based on the well-proven OPR1000 design. The APR1400 has adopted DVI (Direct Vessel Injection) system instead of CLI (Cold Leg Injection) system as advanced safety features of ECCS (Emergency Core Cooling System). The configuration of the improved ECCS in the APR1400 is completely different from that in the OPR1000 in which pipes for the safety water injection are connected to the cold leg. In APR1400, the safety water injection pipes are directly connected to the RPV (Reactor Pressure Vessel). Thus, the safety water injection system in the APR1400 is called the DVI (Direct Vessel Injection) system. Moreover, safety water injections by the HPSI (High Pressure Safety Injection) pumps are mechanically separately in the APR1400.

  18. KSTAR Severe Accident Analysis using MELCOR : Ex-vessel Coolant Pipe Break with Failure of Fusion Power Termination System

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    To investigate the consequence of severe accidents in fusion reactor, a number of thermal hydraulics simulation codes were used (ECART, INTRA, ATHENA/RELAP and so on). MELCOR is chosen as the thermal hydraulics code to simulate the consequence of radioactive material release from accident in preliminary safety report. Capability of the simulation code for fusion reactor severe accident analysis is ability to simulate the hydraulic system in ITER and the transport phenomenon of radionuclides. MELCOR is a fully integrated code that models the accidents in Light Water Reactor (LWR). There are three kinds of radioactive materials in fusion reactor; tritium (or Tiritiated water: HTO), activation products (AP) of divertor or first-wall and activated corrosion products(ACP). In generic Site Safety Report (GSSR), the release guidelines for tritium and activation products are listed for normal operation, incidents, and accidents. And this guidelines presented in Table 1. Not only ITER, the KSTAR (Korea Superconducting Tokamak Advanced Research) is also developing fusion research reactor. The scale of facility is smaller than ITER but this small scale of facility offers the experimental flexibility to develop fusion technology. The major differences between KSTAR and ITER systems are presented in Table 2. Fusion source difference between KSTAR and ITER is D-D fusion reaction (Deuterium-Deuterium fusion reaction) and D-T fusion reaction (Deuterium-Tritium fusion reaction). This D-D fusion makes one tritium by 50 percent chance. The radioactivity of tritium is small to consider compared to radioactive materials in nuclear fission reactor. This reaction is presented in equation (1) In the present work, conservatively estimated tritium inventory amount in KSTAR is used with one of the most severe accident in ITER; Ex-vessel pipe break with Fusion Power Termination System (FPTS). The MELCOR KSTAR input is made by scaling down the ITER input deck. So, the detail system is not same

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

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F.; Gauthier, G.; Carlin, F. [and others

    1996-10-01

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

  20. Experimental investigation on the causes for pellet fragmentation under LOCA conditions

    Science.gov (United States)

    Bianco, A.; Vitanza, C.; Seidl, M.; Wensauer, A.; Faber, W.; Macián-Juan, R.

    2015-10-01

    This paper addresses a separate effect experiment performed with irradiated fuel to study fuel fragmentation and fission gas release during a loss of coolant accident (LOCA). The paper presents a qualitative and quantitative investigation of the effects of the removal of the geometrical constraint provided by the cladding and the removal of the constraint given by the rod internal pressure in determining the extent of fuel fragmentation and fission gas release during a LOCA for fuel segments with a burnup of approximately 52 MWd/kgU. A review of previous LOCA tests was the starting point for the identification of these constraints and for the selection of the fuel rod burnup, the experiment's procedure and the boundary conditions. An out-of-pile test was considered representative for the scope, and the experiment was performed at the Halden Reactor Project hot cell in Kjeller (Norway) with heat provided by an electric oven. Three fuel rod segments were studied: 1) a fuel segment that experienced only ballooning without burst, 2) a fuel segment that experienced ballooning and burst and 3) a fuel segment that experienced neither ballooning nor burst. The neutron radiography and fuel fragment sifting showed that both cladding constraint and internal pressure play a role in the formation of fuel cracks and fragmentation, and the study of the fission gas release during the transient showed that removing the cladding constraint or the internal pressure increased the amount of fission gas release.

  1. Experimental investigation of sedimentation of LOCA - generated fibrous debris and sludge in BWR suppression pools

    International Nuclear Information System (INIS)

    Several tests were conducted in a 1:2.4 scale model of a Mark I suppression pool to investigate the behavior of fibrous insulation and sludge debris under LOCA conditions. NUKON trademark shreds, manually cut and tore up in a leaf shredder, and iron oxide particles were used to simulate fibrous and sludge debris, respectively. The suppression pool model included four downcomers fitted with pistons to simulate the steam-water oscillations during chugging expected during a LOCA. The study was conducted to provide debris settling velocity data for the models used in the BLOCKAGE computer code, developed to estimate the ECCS pump head loss due to clogging of the strainers with LOCA generated debris. The tests showed that the debris, both fibrous and particulate, remains fully mixed during chugging; they also showed that, during chugging, the fibrous debris underwent fragmentation into smaller sizes, including individual fibers. Measured concentrations showed that fibrous debris settled slower than the sludge, and that the settling behavior of each material is independent of the presence of the other material. Finally, these tests showed that the assumption of considering uniform debris concentration during strainer calculations is reasonable. The tests did not consider the effects of the operation of the ECCS on the transport of debris in the suppression pool

  2. Development of Novel Visual-Plus Quantitative Analysis Systems for Studying DNA Double-Strand Break Repairs in Zebrafish

    Institute of Scientific and Technical Information of China (English)

    Jingang Liu; Lu Gong; Changqing Chang; Cong Liu; Jinrong Peng; Jun Chen

    2012-01-01

    The use of reporter systems to analyze DNA double-strand break (DSB) repairs,based on the enhanced green fluorescent protein (EGFP) and meganuclease such as I-Sce Ⅰ,is usually carried out with cell lines.In this study,we developed three visual-plus quantitative assay systems for homologous recombination (HR),non-homologous end joining (NHEJ) and single-strand annealing (SSA) DSB repair pathways at the organismal level in zebrafish embryos.To initiate DNA DSB repair,we used two I-Sce Ⅰ recognition sites in opposite orientation rather than the usual single site.The NHEJ,HR and SSA repair pathways were separately triggered by the injection of three corresponding I-Sce I-cut constructions,and the repair of DNA lesion caused by I-Sce Ⅰ could be tracked by EGFP expression in the embryos.Apart from monitoring the intensity of green fluorescence,the repair frequencies could also be precisely measured by quantitative real-time polymerase chain reaction (qPCR).Analysis of DNA sequences at the DSB sites showed that NHEJ was predominant among these three repair pathways in zebrafish embryos.Furthermore,while HR and SSA reporter systems could be effectively decreased by the knockdown of rad51 and rad52,respectively,NHEJ could only be impaired by the knockdown of ligaseⅣ (lig4) when the NHEJ construct was cut by I-Sce Ⅰ in vivo.More interestingly,blocking NHEJ with lig4-MO increased the frequency of HR,but decreased the frequency of SSA.Our studies demonstrate that the major mechanisms used to repair DNA DSBs are conserved from zebrafish to mammal,and zebrafish provides an excellent model for studying and manipulating DNA DSB repair at the organismal level.

  3. Timing analysis of PWR fuel pin failures

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Straka, M. (Halliburton NUS, Idaho Falls, ID (United States))

    1992-09-01

    Research has been conducted 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 W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report.

  4. Timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    Research has been conducted 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) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report

  5. RIA and LOCA simulating tests on experimental fuel elements in TRIGA MT reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Full text: One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident condition. A total of 39 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 100 test fuel elements have been irradiated in TRIGA SS MTR in different power history conditions. LOCA simulating tests are planned to be performed in C2 LOCA tests capsule and in Loop A of TRIGA SS MTR of INR Pitesti. The LOCA tests in capsule C2 are instrumented to measure fuel, sheath and coolant temperature, internal element and coolant pressure during the entire irradiation period. In the second phase of the experiment the C2 capsule will be connected to the sweep gas system with the on-line gamma ray spectrometer included. RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. This paper

  6. Geoparsing and geosemantics for social media: spatio-temporal grounding of content propagating rumours to support trust and veracity analysis during breaking news

    OpenAIRE

    Middleton, Stuart E.; Krivcovs, Vadims

    2016-01-01

    In recent years there has been a growing trend to use publically available social media sources within the field of journalism. Breaking news has tight reporting deadlines, measured in minutes not days, but content must still be checked and rumours verified. As such journalists are looking at automated content analysis to pre-filter large volumes of social media content prior to manual verification. This paper describes a real-time social media analytics framework for journalists. We extend o...

  7. RELAP5/MOD2 analysis of a postulated ''cold leg SBLOCA'' simultaneous to a ''total black-out'' event in the Jose Cabrera Nuclear Station

    International Nuclear Information System (INIS)

    Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ''lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic ampersand Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumption of a ''total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ''time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested

  8. RELAP5/MOD2 analysis of a postulated ``cold leg SBLOCA`` simultaneous to a ``total black-out`` event in the Jose Cabrera Nuclear Station

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. [Union Electrica, SA, Madrid (Spain)

    1992-04-01

    Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ``lessons learned`` in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic & Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumption of a ``total black-out condition`` coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ``time to core overheating startup`` as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.

  9. Possibility of air ingress into a BWR containment during a LOCA in case of activation of containment venting system

    International Nuclear Information System (INIS)

    The pressure relief systems installed in BWRs protect the containment from overpressure in case of a Loss of Coolant Accident (LOCA). This paper analyzes the possibility of air ingress, which can cause hydrogen burn, through the rupture disks of the filtered and non-filtered venting systems. Two scenarios were considered: a LOCA without SBO (Station Blackout) and a LOCA with SBO. The thermal-hydraulic code GOTHIC® was used with 3D models of the drywell and wetwell of a Nordic-type BWR. In the LOCA event, we found no activation of the rupture disks within the considered transient simulation. Moreover, the containment spray ensured a low pressure in the drywell and induced a continuous mixing of the wetwell pool. In the LOCA with SBO event, the development of thermal stratification in the wetwell pool accelerated the pressure increase in the drywell, which led to activation of the rupture disk of the filtered venting system. However, no air ingress through the vent was found during the depressurization of the containment, and hence no risk of hydrogen burn under the given assumptions. (author)

  10. GOBLIN computer code. Comparison between calculations and TLTA small break test

    International Nuclear Information System (INIS)

    GOBLIN calcuations have been performed for two simulation tests of the boiling water reactor (BWR) small break loss-of-coolant accidents (LOCAs) which were conducted in the two loop test apparatus (TLTA). The first test investigated the small break with nondegraded emergency core coolant (ECC) systems and the second test studied the same small break but with degraded ECC systems in which the high pressure core spray (HPCS) was assumed unavailable. Very good agreement between test data and calculations is achieved. The second test is the most challenging from code comparison point of view and the code prediction of the complicated mass distribution pattern which changes with time is very satisfactory. In the first test and to some extent late in the second test multidimensional subchannel effects are evident in the core bundle region. These are not and cannot be reproduced by the code since the bundle model of GOBLIN is strictly one-dimensional. (Author)

  11. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station

    International Nuclear Information System (INIS)

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  12. Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

  13. RELAP5/MOD2 analysis of a postulated cold leg SBLOCA'' simultaneous to a total black-out'' event in the Jose Cabrera Nuclear Station

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Electrica, SA, Madrid (Spain))

    1992-04-01

    Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumption of a total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.

  14. RBMK-LOCA-Analyses with the ATHLET-Code

    Energy Technology Data Exchange (ETDEWEB)

    Petry, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Kurfuerstendamm, Berlin (Germany); Domoradov, A.; Finjakin, A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    1995-09-01

    The scientific technical cooperation between Germany and Russia includes the area of adaptation of several German codes for the Russian-designed RBMK-reactor. One point of this cooperation is the adaptation of the Thermal-Hydraulic code ATHLET (Analyses of the Thermal-Hydraulics of LEaks and Transients), for RBMK-specific safety problems. This paper contains a short description of a RBMK-1000 reactor circuit. Furthermore, the main features of the thermal-hydraulic code ATHLET are presented. The main assumptions for the ATHLET-RBMK model are discussed. As an example for the application, the results of test calculations concerning a guillotine type rupture of a distribution group header are presented and discussed, and the general analysis conditions are described. A comparison with corresponding RELAP-calculations is given. This paper gives an overview on some problems posed and experience by application of Western best-estimate codes for RBMK-calculations.

  15. Break-glass handling exceptional situations in access control

    CERN Document Server

    Petritsch, Helmut

    2014-01-01

    Helmut Petritsch describes the first holistic approach to Break-Glass which covers the whole life-cycle: from access control modeling (pre-access), to logging the security-relevant system state during Break-Glass accesses (at-access), and the automated analysis of Break-Glass accesses (post-access). Break-Glass allows users to override security restrictions in exceptional situations. While several Break-Glass models specific to given access control models have already been discussed in research (e.g., extending RBAC with Break-Glass), the author introduces a generic Break-Glass model. The pres

  16. Analysis and Verification of Direct Vessel Injection Line Break event tree for AP1000 reactor with TRACE code

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    The AP1000 PRA thermal hydraulic simulations were performed with MAAP code, which allows simulating sequences with low computational efforts. On the other hand, the use of best estimate codes allows verifying PRA results as well as obtaining a greater knowledge of the phenomenology of such sequences. The initiating event with the greatest contribution to core damage is Direct Vessel Injection Line Break (DVILB). This paper presents a review of DVILB sequences of AP1000 with TRACE code for verifying sequences previously analyzed by Westinghouse with MAAP code. The sequences which configure the DVILB event tree during short term have been simulated. The results obtained confirm the ones obtained in AP1000 PRA.

  17. Transient simulation of feedwater vaporization during a DBA LOP/LOCA using RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Harrell, J.R. [ENERCON Services, Inc., Atlanta, GA (United States); Fuller, R.W. [Entergy Operations, Inc., Port Gibson, MS (United States)

    1996-07-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station (GGNS) are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. The original design and testing requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. Given this condition, the appropriate testing criteria would be based on air with a relatively tight allowable limit. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leakage flow exists from the reactor vessel to the condenser through the feedwater piping during the reactor vessel blowdown phase. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  18. An outcome of nuclear safety research in JAERI. Case study for LOCA, FP, criticality and reprocessing

    International Nuclear Information System (INIS)

    An outcome of nuclear safety research done by JAERI was case studied by the bibliometric method. (1) For LOCA (loss-of-coolant accident) a domestic share of JAERI in monoclinic research paper was 63% at the past (20) 1978-1982 but was decreased to 40% at the present 1998-2002. For co-authored papers a domestic share between JAERI and PS (public sectors) is almost zero at past (20) but increased to 4% at the present. Research cooperation is active between Tokyo University and JAERI or between JAERI and Nagoya University. (2) Project-type research is to have a large monopolization in papers and that of basic-type research is to have a large development of research networking (DRN). (3) For FP, a share of co-authored paper is high due to an enhanced cooperation among JAERI-PO (Public Organization)-PS. For criticality, research activity was enhanced after JCO accident, especially at NUCEF. (4) For reprocessing, PS had a monopolistic position with a domestic share of 71% and a share of JAERI was about 20%. (5) LOCA and RIA outputs born by NSR-JAERI coincided partly to those of the Safety Licensing Guidelines but a share of contribution done by JAERI was ambiguous due to the lack of necessary information. (author)

  19. Study of two-phase flow under low velocity in PWR-LOCA

    International Nuclear Information System (INIS)

    In this research, when the hydrothermal behavior at the time of LOCA in PWRs is forecast with high accuracy by the optimal forecast code based on a two-fluid model, there are some problems in the accuracy of forecast of low velocity two-phase flow, therefore, the subjects related to it were studied. The subjects are the forecast using the multi-dimensional optimal forecast code of the heat transfer acceleration phenomena in high power output fuel assemblies originating in the radial power distribution in a core in the reflooding period at the time of LOCA, the evaluation of the applicability of the hydrothermal model in a core in reflooding period which was developed for 15 x 15 type fuel assemblies to 17 x 17 type fuel assemblies, the grasping of the mechanism of opposite flow limiting phenomena in hot leg, the elucidation of the scale effect of channels and the development of a model which is applicable up to actual reactor scale. The studies carried out on these subjects and the conclusions of respective studies are reported. (K.I.) 79 refs

  20. LOCA- and ATWS-calculations for homogeneous and heterogeneous advanced pressurized water reactors

    International Nuclear Information System (INIS)

    LOCA and ATWS calculations have been performed for the two KfK reference designs (homogeneous with p/d=1.2 and heterogeneous reactor) of APWR and for a homogeneous reactor with a tighter fuel rod lattice (p/d=1.123) as well as for a reference PWR. The calculations have been performed with the Ispra version of the code RELAP5/Mod.1. New correlations have been introduced in the code to account for the core geometry, which is different from that of a PWR. The results of the calculations show that during the LOCA the fuel rod cladding hot spot temperatures in the seed of the heterogeneous reactor reach values which are about 2500C higher than the corresponding temperatures for a PWR, and that during the ATWS the pressure inside the primary circuit exceeds the maximal allowable pressure in the case of the homogeneous reactor with p/d=1.123. On the basis of the present calculations only the homogeneous reactor with p/d=1.2 appears to be acceptable from a safety point of view. These results need of course experimental confirmation. (orig.)

  1. The LORELEI test device for LOCA experiments in the Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    For the operational starting of the Jules Horowitz material testing reactor (JHR) in France, CEA intends to promote different experimental devices aiming at testing materials and fuels under irradiation. More specifically, in the safety studies domain, the LORELEI 'Light water One Rod Equipment for LOCA Experimental Investigations' test device is dedicated to the study of the thermal-mechanical behavior of a fuel rod and quantification of Fission Product (FP) and fissile material releases under Loss Of Coolant Accident (LOCA) type conditions. Lorelei will allow investigating ballooning and burst of the fuel cladding, clad corrosion phenomena (oxidation and hydriding), post quench behavior and FP release (for that aim, the device will be connected to water and gas sampling lines to the Fission Product Laboratory of the JHR). This paper presents the scientific objectives of the Lorelei device, its main technical characteristics, its experimental protocol and the main options of the preliminary design, based on thermal-hydraulic and mechanical studies. Examinations planned on non-destructive underwater benches and characterization laboratories of the JHR will be detailed, in order to highlight their essential support for gaining as soon as possible exclusive and valuable scientific data on the 'as tested' sample. (author)

  2. Analysis of gene repair tracts from Cas9/gRNA double-stranded breaks in the human CFTR gene.

    Science.gov (United States)

    Hollywood, Jennifer A; Lee, Ciaran M; Scallan, Martina F; Harrison, Patrick T

    2016-01-01

    To maximise the efficiency of template-dependent gene editing, most studies describe programmable and/or RNA-guided endonucleases that make a double-stranded break at, or close to, the target sequence to be modified. The rationale for this design strategy is that most gene repair tracts will be very short. Here, we describe a CRISPR Cas9/gRNA selection-free strategy which uses deep sequencing to characterise repair tracts from a donor plasmid containing seven nucleotide differences across a 216 bp target region in the human CFTR gene. We found that 90% of the template-dependent repair tracts were >100 bp in length with equal numbers of uni-directional and bi-directional repair tracts. The occurrence of long repair tracts suggests that a single gRNA could be used with variants of the same template to create or correct specific mutations within a 200 bp range, the size of ~80% of human exons. The selection-free strategy used here also allowed detection of non-homologous end joining events in many of the homology-directed repair tracts. This indicates a need to modify the donor, possibly by silent changes in the PAM sequence, to prevent creation of a second double-stranded break in an allele that has already been correctly edited by homology-directed repair. PMID:27557525

  3. Evaluation of the simultaneous action of earthquake, LOCA and SRV on Mark-III containment and drywell structures. [Safety Relief Valve

    Energy Technology Data Exchange (ETDEWEB)

    Philippacopoulos, A.J.; Reich, M.

    1981-01-01

    The containment and drywell structures of a generic Mark-III nuclear power facility are investigated with respect to their structural response when subjected to various load combinations that may occur during their operational lifetime. Both structures are idealized with three-dimensional finite element models. In addition a lumped-mass cantilever beam representation of the containment and drywell structures is developed and used for the soil-structure interaction analysis. The types of dynamic events considered include earthquakes, loss-of-coolant-accidents (LOCA) and those generated by safety/relief valve actuations. The evaluation of the combined actions of these dynamic events is first performed by using combination methods such as Absolute Sum (ABS) and Square-Root-of-the-Sum-of-the-Squares (SRSS). A Monte Carlo simulation procedure is subsequently used for probabilistic evaluation of the combinations.

  4. Near-infrared spectroscopic analysis of the breaking force of extended-release matrix tablets prepared by roller-compaction: influence of plasticizer levels and sintering temperature.

    Science.gov (United States)

    Dave, Vivek S; Fahmy, Raafat M; Hoag, Stephen W

    2015-06-01

    The aim of this study was to investigate the feasibility of near-infrared (NIR) spectroscopy for the determination of the influence of sintering temperature and plasticizer levels on the breaking force of extended-release matrix tablets prepared via roller-compaction. Six formulations using theophylline as a model drug, Eudragit® RL PO or Eudragit® RS PO as a matrix former and three levels of TEC (triethyl citrate) as a plasticizer were prepared. The powder blend was roller compacted using a fixed roll-gap of 1.5 mm, feed screw speed to roller speed ratio of 5:1 and roll pressure of 4 MPa. The granules, after removing fines, were compacted into tablets on a Stokes B2 rotary tablet press at a compression force of 7 kN. The tablets were thermally treated at different temperatures (Room Temperature, 50, 75 and 100 °C) for 5 h. These tablets were scanned in reflectance mode in the wavelength range of 400-2500 nm and were evaluated for breaking force. Tablet breaking force significantly increased with increasing plasticizer levels and with increases in the sintering temperature. An increase in tablet hardness produced an upward shift (increase in absorbance) in the NIR spectra. The principle component analysis (PCA) of the spectra was able to distinguish samples with different plasticizer levels and sintering temperatures. In addition, a 9-factor partial least squares (PLS) regression model for tablets containing Eudragit® RL PO had an r(2) of 0.9797, a standard error of calibration of 0.6255 and a standard error of cross validation (SECV) of 0.7594. Similar analysis of tablets containing Eudragit® RS PO showed an r(2) of 0.9831, a standard error of calibration of 0.9711 and an SECV of 1.192.

  5. Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

    Energy Technology Data Exchange (ETDEWEB)

    Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

    1997-04-01

    The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

  6. Transient recovery voltage analysis for various current breaking mathematical models: shunt reactor and capacitor bank de-energization study

    Directory of Open Access Journals (Sweden)

    Oramus Piotr

    2015-09-01

    Full Text Available Electric arc is a complex phenomenon occurring during the current interruption process in the power system. Therefore performing digital simulations is often necessary to analyse transient conditions in power system during switching operations. This paper deals with the electric arc modelling and its implementation in simulation software for transient analyses during switching conditions in power system. Cassie, Cassie-Mayr as well as Schwarz-Avdonin equations describing the behaviour of the electric arc during the current interruption process have been implemented in EMTP-ATP simulation software and presented in this paper. The models developed have been used for transient simulations to analyse impact of the particular model and its parameters on Transient Recovery Voltage in different switching scenarios: during shunt reactor switching-off as well as during capacitor bank current switching-off. The selected simulation cases represent typical practical scenarios for inductive and capacitive currents breaking, respectively.

  7. SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

    Energy Technology Data Exchange (ETDEWEB)

    Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

    1995-09-01

    The 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 (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) 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 1994, 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 passive and active safety systems and a main steam line break transient to demonstrate the boration 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. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

  8. Characterization of fracture behavior of zirconium alloys for fuel rod cladding of nuclear power plant in the post-quench stage of a LOCA (Loss of Coolant Accident)

    International Nuclear Information System (INIS)

    In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to characterize the ductility of cladding after a Loss of Coolant Accident (LOCA). The thesis is about the characterization of the fracture behavior of cold-worked stress-relieved Zircaloy-4 claddings which have undergone LOCA conditions simulated in laboratory by a high temperature oxidation followed by a cooling. The high temperature oxidation is carried out at 1100 C and 1200 C with different times, which leads to different oxidation levels varying from 3% to 30% ECR (Equivalent Cladding Reacted). The high temperature oxidation is followed by two types of cooling: water quench and air cooling. The oxidized claddings contain two fragile layers - the outer zirconium oxide ZrO2 layer and the middle a(O) layer, and a layer which can have residual ductility - the inner ex-β layer. Characterizations by means of optical microscopy, electron probe micro analysis and nano-indentation have been carried out on the oxidized claddings. A correlation between the oxygen concentration and the nano-hardness and the Young's modulus has been proposed.The Expansion Due to Compression (EDC) test has been developed with an instrumentation of stereo digital image correlation, and then used to characterize the mechanical behavior of the oxidized claddings. The behavior of the oxidized claddings has been studied via macroscopic EDC test curves and observations of fractured or pre-deformed test samples. A fracture scenario of the oxidized claddings has been proposed. The fracture scenario has then been validated via EDC tests on oxidized claddings whose ZrO2 and a(O) layers have been removed, and via finite element modeling of EDC tests. Moreover, a fracture criterion has been established. The mechanical behavior modeling and the proposed fracture criterion have been validated by modeling of ring compression test. (author)

  9. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  10. Topological amplitudes in $D$ decays to two pseudoscalars: a global analysis with linear $SU(3)_F$ breaking

    CERN Document Server

    Müller, Sarah; Schacht, Stefan

    2015-01-01

    We study decays of $D^0$, $D^+$, and $D_s^+$ mesons into two pseudoscalar mesons by expressing the decay amplitudes in terms of topological amplitudes. Including consistently SU(3)$_F$ breaking to linear order, we show how the topological-amplitude decomposition can be mapped onto the standard expansion using reduced amplitudes characterized by SU(3) representations. The tree and annihilation amplitudes can be calculated in factorization up to corrections which are quadratic in the color-counting parameter $1/N_c$. We find new sum rules connecting $D^+\\rightarrow K_SK^+$, $D_s^+\\rightarrow K_S\\pi^+$ and $D^+\\rightarrow K^+\\pi^0$, which test the quality of the $1/N_c$ expansion. Subsequently, we determine the topological amplitudes in a global fit to the data, taking the statistical correlations among the various measurements into account. We carry out likelihood ratio tests in order to quantify the role of specific topological contributions. While the SU(3)$_F$ limit is excluded with a significance of more th...

  11. CFD simulations of ballooned regions in a damaged core during the LOCA reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, N.; Ruyer, P.; Lelong, F.; Secondi, F. [IRSN/DPAM/SEMCA/LEMAR, CE Cadarache, Saint Paul lez Durance (France); Gradeck, M. [LEMTA Nancy Univ. CNRS, Vandoeuvre les Nancy (France)

    2011-07-01

    This study focuses on the cooling capacity of a damaged PWR reactor core during the reflooding phase of a Loss Of Coolant Accident (LOCA). Downstream the quench front, the core cooling is provided by an over-heated vapour flow carrying water droplets and may impact the ballooned fuel cladding and provide an additional cooling. The present paper will deal with the development of a CFD code to simulate such droplet dispersed flows with the final aim of carrying out sensitivity studies of blockage ratio and length on wall cooling. Adequate closure laws for the momentum and energy balances as well as for the heat transfer at droplet impact and interfacial area transport are given and some simulation results are presented. (author)

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

    International Nuclear Information System (INIS)

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

  13. Comparison of code calculations with experiments on containment response during LOCA conditions

    International Nuclear Information System (INIS)

    A series of experiments were performed on a one-tenth scale model of PHWR containment, incorporating pressure suppression system. The pressure-temperature transients in the model containment observed during simulated LOCA (Loss of Coolant) blowdown conditions were compared against calculated results form computer code PACSR, for purposes of verification of the code. Comparison of results indicated that calculated values of peak pressure in various compartment were significantly higher than observed ones. This disagreement was attributed mainly to modelling for energy absorption from containment atmosphere to structural surfaces, this effect being particularly important in a scaled down model. Good agreement between calculation and experiment was obtained after heat transfer correlation for energy absorption on surfaces were modified in the code. The study demonstrates the conservatism of the results from the code. (author). 6 refs., 1 tab., 9 figs

  14. Simulation of CANDU Fuel Behaviour into In-Reactor LOCA Tests

    International Nuclear Information System (INIS)

    The purpose of this work is to simulate the behaviour of an instrumented, unirradiated, zircaloy sheathed UO2 fuel element assembly of CANDU type, subjected to a coolant depressurization transient in the X-2 pressurized water loop of the NRX reactor at the Chalk River Nuclear Laboratories in 1983. The high-temperature transient conditions are such as those associated with the onset of a loss of coolant accident (LOCA). The data and the information related to the experiment are those included in the OECD/NEA-IFPE Database (IFPE/CANDU-FIO-131 NEA-1783/01). As tool for this simulation is used the TRANSURANUS fuel performance code, developed at ITU, Germany, along with the corresponding fabrication and in-reactor operating conditions specific of the CANDU PHWR fuel. The results, analyzed versus the experimental ones, are encouraging and perfectible. (author)

  15. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station; Analisis de eventos internos para la Unidad 1 de la Central Nucleolelectrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1993-07-01

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  16. PIE of the second fuel rod from the LOCA experiment (IFA-650.2)

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Jenssen, H.K.; Espeland, M.; Solum, N.O.

    2005-07-01

    The LOCA experiment on the second rod (IFA-650.2) a fresh, low-tin Zr-4, pressurised PWR rod was carried out in May 2004. The main objective was to produce ballooning, to determine the time to burst and to assess the material oxidation and hydriding kinetics. The rod pressure at hot conditions and peak PCT were 70 bar and 1050 C, respectively. To document the effect of the LOCA testing on the cladding, rod 2 was subjected in PIE to visual inspection, profilometry and metallography. On visual inspection the clad shows a typical balloon. In the region of max ballooning the clad shows a 35 mm long, < 20 mm burst opening. In the balloon region the outer clad diameter increased by <35% and locally the wall thickness reduction is >55%. The entire rod is covered with a black oxide layer. Below and above the split opening the continuous oxide layer was 40 to 50mum both on water and fuel side of the clad. At the locations of the upper and lower cladding thermocouples the oxide thickness was in the range 24-27 mum. Widmanstaetten structure is seen in the bulk of the clad and confirms the high temperature experienced. A some 40mum wide, hard and brittle zone with oxygen rich globular alpha-grains is found both at the outer and the inner edge of the clad in the balloon region. The zone is widest near the axial split (crack). In the clad few, arbitrary oriented hydride platelets are observed in the balloon area. (Author)

  17. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  18. Liquid-Bridge Breaking Limits

    Science.gov (United States)

    Macner, Ashley; Steen, Paul

    2011-11-01

    Wet adhesion by liquid bridges in large arrays shows promise for use in lightweight, controllable on-demand devices. Applications include grab/release of wafer substrates, transport of micron-sized tiles for use in 3D printing and micro-dosing of personalized pharmaceutical drugs. By wetting and spreading, a drop can form a bridge and thereby ``grab'' a nearby solid substrate. By volume decrease or extension, the bridge can break. The breaking limit corresponds to bridge instability which can be predicted, knowing the static mechanical response of the bridge. Mechanical behaviors include force-volume (FV), pressure-volume (pV) and force-length (FL) responses. Instability crucially depends on the mode of failure - failure under constant-force or constant length are typical cases. We study single bridge equilibria for their breaking limits. FV diagrams for the pin-pin equal and pin-pin unequal radii boundary conditions for different bridge heights are measured in the laboratory. The FL response in the case of pin-pin equal radii is also measured. Results are compared to predictions of static theory. Static results are then used to compare to dynamical sequences where volume is driven quasistatically by syringe or an electro-osmotic pump. As the breaking limit is approached, the shape deformation accelerates leading to non-equilibrium shapes not captured by the static analysis.

  19. Ballooning analysis for the Sizewell B PWR using symmetric MABEL calculations

    International Nuclear Information System (INIS)

    An analysis of the fuel clad ballooning potential associated with the Sizewell B PWR following a design basis large break cold leg LOCA is described. Calculations employ MABEL-2C code. No allowance has been made for asymmetries in power or geometry, thus precluding any amelioration offered by early clad rupture. Thermal hydraulic data were derived from a TRAC-PD2 best estimate analysis of the LOCA and the work includes a detailed sensitivity study which leads to a correlation between peak clad temperature and clad strain. For the best estimate start of cycle 1 peak rod rating, no loss of coolability is expected within 95 percent confidence limits on peak clad temperature. No loss of coolability is expected either for rods at the design basis peak rod rating. The temperature does not have to be much higher than the 95 percent confidence limit on the best estimate rating or much beyond that of the design basis rating for rod contact and severe blockage to follow. This indicates that to establish a complete safety case the added complexity of pellet eccentricity and rod to rod power variations must be considered. (U.K.)

  20. ALARM-P1: a computer program for pressurized water reactor blowdown analysis

    International Nuclear Information System (INIS)

    The computer program ALARM-P1 written in FORTRAN-IV for FACOM 230-75 is a part of the code series for evaluation of performance of the emergency core cooling system (ECCS) in pressurized water reactors according to the safety evaluation guidelines provided by the Atomic Energy Commission of Japan. ALARM-P1 is for analyzing the thermo-hydraulic phenomena during blowdown following a large break in the primary coolant system. ALARM-P1 models the PWR system fluid conditions including flow, pressure, mass inventory, fluid quality and heat transfer. It solves integral forms of fluid conservation and state equations for user-defined volumes treated as one-dimensional homogeneous, thermal-equilibrium elements with interconnecting flow paths and also finite difference forms of the one-dimensional heat conduction equations describing temperature profiles within solid material and the fluid-solid interface conditions. In addition, the ALARM-P1 provides the initial conditions for analysis of the last portion of the LOCA transient, a reflood phase, and the information for core heat-up analysis during the whole LOCA. This report describes the state-of-art methods and models of ALARM-P1 in June 1978 and gives information for users. (author)

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

    International Nuclear Information System (INIS)

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

  2. Break the ice

    Institute of Scientific and Technical Information of China (English)

    2015-01-01

    Jacky:My sister is mad at me.She refuses(拒绝)to talk to me.What can I do to break the ice?Ella:You can buy her a little gift.Break的意思是"打破",ice是指"冰块"。冰是又冷又硬的东西,作为俗语break the ice是指"打破沉默(僵局)"。Jacky惹妹妹生气,妹妹不理他了,他能通过送小礼物break the ice吗?

  3. Ice-Induced Fatigue Analysis by Spectral Approach for Offshore Jacket Platforms with Ice-Breaking Cones

    Institute of Scientific and Technical Information of China (English)

    YUE Qian-jin; LIU Yuan; ZHANG Ri-xiang; QU Yan; WANG Rui-xue

    2007-01-01

    The spectral methods and ice-induced fatigue analysis are discussed based on Miner's linear cumulative fatigue hypothesis and S-N curve data.According to the long-term data of full-scale tests on the platforms in the Bohai Sea,the ice force spectrum of conical structures and the fatigue environmental model are established.Moreover,the finite element model of JZ20-2MSW platform,an example of ice-induced fatigue analysis,is built with ANSYS software.The mode analysis and dynamic analysis in frequency domain under all kinds of ice fatigue work conditions are carried on,and the fatigue life of the structure is estimated in detail.The methods in this paper can be helpful in ice-induced fatigue analysis of ice-resistant platforms.

  4. Testing cosmological supersymmetry breaking

    CERN Document Server

    Kabat, D; Kabat, Daniel; Rajaraman, Arvind

    2001-01-01

    Banks has proposed a relation between the scale of supersymmetry breaking and the cosmological constant in de Sitter space. His proposal has a natural extension to a general FRW cosmology, in which the supersymmetry breaking scale is related to the Hubble parameter. We study one consequence of such a relation, namely that coupling constants change as the universe evolves. We find that the most straightforward extension of Banks' proposal is disfavored by experimental bounds on variation of the fine structure constant.

  5. Self-Breaking Technicolor

    CERN Document Server

    Martin, S P

    1993-01-01

    We propose a scenario in which the electroweak symmetry is spontaneously broken by an $SU(4)$ technicolor gauge interaction which also manages to break itself completely. The technicolor gauge bosons and technifermions are not confined by the technicolor force, but get large masses. Starting with a single technidoublet, one emerges with a complete standard model family of technifermions after the symmetry breaking is complete. This suggests a broad new avenue for model building. A few variations on the theme are mentioned.

  6. Development of regulatory technology on a coupled 3-dimensional core neutronics and system thermal-hydraulics/analysis of steam line breaks for Kori Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Lee, J. I.; Yang, C. Y.; Jeong, H. Y.; Jang, C. S.; Jung, J. W.; Na, W. J

    2001-02-15

    From different mechanical properties in highly irradiated state of high-burnup fuels, it was finalized that the regulatory fuel failure criteria for reactivity-induced accident and loss-of-coolant accident should be re-evaluated. Since it is supposed in the near future that these criteria will be set far below the current ones, one of the ways to examine the issues of the high-burnup fuels is to apply best-estimate methodology for licensing design. However, there are still many problems in the application of the methodology for licensing. This study finds out the general regulatory issues and problems in the application of the multi-dimensional approach of the core, by using RELAP5/PARCS, which is a coupled 3-dimensional core neutronics and system thermal-hydraulics code. The RELAP5/PARCS input deck was developed for a core configuration of cycle 19 with EOC of Kori unit 1. The most important process for a PARCS input deck was the way to generate macroscopic cross sections. CASMO-3 code was used for their generation. The process that macroscopic cross sections of a PARCS input are produced from a CASMO-3 output was computerized. The RELAP5/PARCS input deck developed was updated and improved by comparing the results of RELAP5 stand-alone calculation for steam line break accidents. Various sensitivity studies were carried out for break areas, safety injection setpoints, fuel classifications, and etc. It is important to understand the uncertainty of the fuel storage energy calculated from best-estimate methods. The uncertainty is clarified through the sensitivity studies. The systematic procedure to produce RELAP5/PARCS input deck would be help performing the sensitivity analysis.

  7. Research on fuel cladding failure temperature criteria and gap release radioactive aftereffect for LOCA of ship reactor

    International Nuclear Information System (INIS)

    Aiming at the particular safety requirement of ship reactor, the research on fuel cladding failure temperature criteria for LOCA was carried out. The conservative assumption was abandoned, the reasonable fuel cladding failure temperature criteria was gained with the best estimate model, and the fraction of failure and gap release radioactive aftereffect were calculated by MELCOR code. The calculation results can give a reference for evaluating cabin dose and ensuring the safety for operators. (authors)

  8. Development of LOCA Response Strategy during Shutdown Operation in Kori 3 and 4 and Ygn 1 and 2 Units

    International Nuclear Information System (INIS)

    The abnormal transients and accidents during shutdown operation modes of pressurized water reactor, in contrast with those during full power operation modes, have been recently issued due to their potential risk for leading to a severe abnormal condition because the various safety-related systems and equipments may be unavailable. The NRC issued Loss of Reactor Coolant Inventory and Potential Loss of Emergency Mitigation Functions While in a Shutdown Condition, on January 12, 1995, to alert addressees to an incident at the Wolf Creek plant involving the loss of reactor coolant inventory while the reactor was in a hot shutdown condition. Considering those situations, abnormal response mitigation strategies in shutdown operation modes have been developed by evaluating the plant specific thermal hydraulic behavior following LOCA. The objectives of this study are: (1) to verify the effectiveness of abnormal operating procedure (AOP) and emergency core cooling system (ECCS) on reactor safety at shutdown operation, developing the operation strategy at the condition and abnormal operation guideline in the light water reactors; (2) to enhance understanding of the thermal-hydraulic phenomena during abnormal and accident conditions. In this paper, some accident scenarios are analyzed to develop the abnormal operating strategy following LOCA in shutdown operation modes for Kori 3 and 4 and Ygn 1and 2 units. This study shows the adequacy of operator action time to mitigate the LOCA, the thermal hydraulic behavior, reactor inventory distribution and possibility of cold overpressurization after SBLOCA during shutdown operation modes using computer code RELAP5 /MOD3.2

  9. Electroweak symmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Chanowitz, M.S.

    1990-09-01

    The Higgs mechanism is reviewed in its most general form, requiring the existence of a new symmetry-breaking force and associated particles, which need not however be Higgs bosons. The first lecture reviews the essential elements of the Higgs mechanism, which suffice to establish low energy theorems for the scattering of longitudinally polarized W and Z gauge bosons. An upper bound on the scale of the symmetry-breaking physics then follows from the low energy theorems and partial wave unitarity. The second lecture reviews particular models, with and without Higgs bosons, paying special attention to how the general features discussed in lecture 1 are realized in each model. The third lecture focuses on the experimental signals of strong WW scattering that can be observed at the SSC above 1 TeV in the WW subenergy, which will allow direct measurement of the strength of the symmetry-breaking force. 52 refs., 10 figs.

  10. Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

    Directory of Open Access Journals (Sweden)

    Lázaro Lima

    2011-12-01

    Full Text Available

    Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

  11. Prediction of Reactor Vessel Water Level Using Fuzzy Neural Networks in Severe Accidents due to LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soonho; Kim, Jaehawn; Na, Mangyun [Chosun Univ., Gwangju (Korea, Republic of)

    2013-05-15

    When the initial events that may lead to the severe accident such as Loss Of Coolant Accident (LOCA) and Steam Generator Tube Rupture (SGTR) occurs at a nuclear power plant, it is most important to check the status of the plant conditions by observing the safety-related parameters such as neutron flux, pressurizer pressure, steam generator pressure and water level. In this paper, we propose a method of predicting the water level of coolant in the reactor vessel that directly affect the important events such as the exposure of the reactor core and the damage of reactor vessel by using a Fuzzy Neural Network (FNN) method. In addition, the data for verifying a proposed model was obtained by simulating the severe accident scenarios for the OPR1000 nuclear power plant using the MAAP4 code. In this paper, a prediction model was developed for predicting the reactor vessel water level using the FNN method. The proposed FNN model was verified based on the simulation data of OPR1000 by using MAAP4 code. As a result of simulation, we could see that the performance of the proposed FNN model is quite satisfactory but some large errors are observed occasionally. If the proposed FNN model is optimized by using a variety of data, it is possible to predict the reactor vessel water level exactly.

  12. Post-Test Examination of the Hydrogen Distribution in Zirconium Claddings After Loca Tests

    International Nuclear Information System (INIS)

    In the framework of the posttest examinations of the large scale LOCA simulation tests on fuel rod bundle scale, the hydrogen distributions in specimens prepared from the QUENCH-L0 and –L1 tests were determined by means of neutron imaging. The hydrogen distributions in samples prepared from the two tests differ significantly. Whereas clearly visible hydrogen bands were found in the inner rods of the QUENCH-L0 test; the hydrogen enrichments in no specimen prepared from the inner rods of the QUENCH-L1 test are more blurred. The reasons for these different behaviors can be the different times between reaching the temperature maxima and the quenching. In the QUENCH-L0 test the bundle was quenched immediately after reaching the maximal temperature. In QUENCH-L1 the hydrogen has about 130 s to diffuse and reach more homogeneous distributions without very pronounced contrasts between the hydrogen bands and the neighboring regions in the neutron images. In outer rods of both tests no hydrogen enrichments were found except two rods (#14 and #17) of test QUENCH-L0. The reason for it is the slightly lower temperature of the outer rods compared to the inner rods. (author)

  13. Proceedings of the seminar on leak before break in reactor piping and vessels

    International Nuclear Information System (INIS)

    The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  14. Proceedings of the seminar on leak before break in reactor piping and vessels

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [ed.] [Electricite de France, Villeurbanne (France); Gilles, P. [ed.] [Framatome, Paris (France)

    1997-04-01

    The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  15. Plant application uncertainty evaluation of LBLOCA analysis using RELAP5/MOD3/KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Chung, Bub Dong; Hwang, Tae Suk; Lee, Guy Hyung; Chang, Byung Hoon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the revised ECCS rule has been developed and this report describes the application of new REM to large break LOCA. A computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/ MOD3.1 was used as the best estimated code for the analysis and Kori unit 3 and 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by using Monte-Carlo sampler for each response surface. This study shows that plant application uncertainty can be quantified and demonstrates the applicability of the new realistic evaluation methodology. (Author) 29 refs., 40 figs., 8 tabs.

  16. Detecting Structural Breaks using Hidden Markov Models

    DEFF Research Database (Denmark)

    Ntantamis, Christos

    Testing for structural breaks and identifying their location is essential for econometric modeling. In this paper, a Hidden Markov Model (HMM) approach is used in order to perform these tasks. Breaks are defined as the data points where the underlying Markov Chain switches from one state to anoth...... in the monetary policy of United States, the dierent functional form being variants of the Taylor (1993) rule.......Testing for structural breaks and identifying their location is essential for econometric modeling. In this paper, a Hidden Markov Model (HMM) approach is used in order to perform these tasks. Breaks are defined as the data points where the underlying Markov Chain switches from one state to another....... The estimation of the HMM is conducted using a variant of the Iterative Conditional Expectation-Generalized Mixture (ICE-GEMI) algorithm proposed by Delignon et al. (1997), that permits analysis of the conditional distributions of economic data and allows for different functional forms across regimes...

  17. Breaking the Waves

    DEFF Research Database (Denmark)

    Christensen, Poul Rind; Kirketerp, Anne

    2006-01-01

    The paper shortly reveals the history of a small school - the KaosPilots - dedicated to educate young people to carriers as entrepreneurs. In this contribution we want to explore how the KaosPilots managed to break the waves of institutionalised concepts and practices of teaching entrepreneurship...

  18. Model Breaking Points Conceptualized

    Science.gov (United States)

    Vig, Rozy; Murray, Eileen; Star, Jon R.

    2014-01-01

    Current curriculum initiatives (e.g., National Governors Association Center for Best Practices and Council of Chief State School Officers 2010) advocate that models be used in the mathematics classroom. However, despite their apparent promise, there comes a point when models break, a point in the mathematical problem space where the model cannot,…

  19. Phylogenetic analysis of Tomato spotted wilt virus (TSWV) NSs protein demonstrates the isolated emergence of resistance-breaking strains in pepper.

    Science.gov (United States)

    Almási, Asztéria; Csilléry, Gábor; Csömör, Zsófia; Nemes, Katalin; Palkovics, László; Salánki, Katalin; Tóbiás, István

    2015-02-01

    Resurgence of Tomato spotted wilt virus (TSWV) worldwide as well as in Hungary causing heavy economic losses directed the attention to the factors contributing to the outbreak of this serious epidemics. The introgression of Tsw resistance gene into various pepper cultivars seemed to solve TSWV control, but widely used resistant pepper cultivars bearing the same, unique resistance locus evoked the rapid emergence of resistance-breaking (RB) TSWV strains. In Hungary, the sporadic appearance of RB strains in pepper-producing region was first observed in 2010-2011, but in 2012 it was detected frequently. Previously, the non-structural protein (NSs) encoded by small RNA (S RNA) of TSWV was verified as the avirulence factor for Tsw resistance, therefore we analyzed the S RNA of the Hungarian RB and wild type (WT) isolates and compared to previously analyzed TSWV strains with RB properties from different geographical origins. Phylogenetic analysis demonstrated that the different RB strains had the closest relationship with the local WT isolates and there is no conserved mutation present in all the NSs genes of RB isolates from different geographical origins. According to these results, we concluded that the RB isolates evolved separately in geographic point of view, and also according to the RB mechanism. PMID:25331341

  20. Loss of coolant accident analysis and evolution of emergency core cooling system for an inpile irradiation facility

    International Nuclear Information System (INIS)

    This paper deals with the Loss of Coolant Accident (LOCA) analysis of an inpile facility using RELAP4/MOD6 computer code. The present study is the culmination of a three part LOCA analysis done earlier by the authors. Blowdown analysis had been extended to include reflood part of the transient. Based on the analysis an Emergency Core Cooling System (ECCS) has been evolved. (author). 5 figs., 2 tabs

  1. Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

  2. Breaking Computational Barriers: Real-time Analysis and Optimization with Large-scale Nonlinear Models via Model Reduction.

    Energy Technology Data Exchange (ETDEWEB)

    Drohmann, M.; Tuminaro, Raymond S.; Boggs, Paul T.; Ray, Jaideep; van Bloemen Waanders, Bart Gustaaf; Carlberg, Kevin Thomas

    2014-11-01

    -model errors. This enables ROMs to be rigorously incorporated in uncertainty-quantification settings, as the error model can be treated as a source of epistemic uncertainty. This work was completed as part of a Truman Fellowship appointment. We note that much additional work was performed as part of the Fellowship. One salient project is the development of the Trilinos-based model-reduction software module Razor , which is currently bundled with the Albany PDE code and currently allows nonlinear reduced-order models to be constructed for any application supported in Albany. Other important projects include the following: 1. ROMES-equipped ROMs for Bayesian inference: K. Carlberg, M. Drohmann, F. Lu (Lawrence Berkeley National Laboratory), M. Morzfeld (Lawrence Berkeley National Laboratory). 2. ROM-enabled Krylov-subspace recycling: K. Carlberg, V. Forstall (University of Maryland), P. Tsuji, R. Tuminaro. 3. A pseudo balanced POD method using only dual snapshots: K. Carlberg, M. Sarovar. 4. An analysis of discrete v. continuous optimality in nonlinear model reduction: K. Carlberg, M. Barone, H. Antil (George Mason University). Journal articles for these projects are in progress at the time of this writing.

  3. Breaking Computational Barriers: Real-time Analysis and Optimization with Large-scale Nonlinear Models via Model Reduction

    Energy Technology Data Exchange (ETDEWEB)

    Carlberg, Kevin Thomas [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Quantitative Modeling and Analysis; Drohmann, Martin [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Quantitative Modeling and Analysis; Tuminaro, Raymond S. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Computational Mathematics; Boggs, Paul T. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Quantitative Modeling and Analysis; Ray, Jaideep [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Quantitative Modeling and Analysis; van Bloemen Waanders, Bart Gustaaf [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Optimization and Uncertainty Estimation

    2014-10-01

    -model errors. This enables ROMs to be rigorously incorporated in uncertainty-quantification settings, as the error model can be treated as a source of epistemic uncertainty. This work was completed as part of a Truman Fellowship appointment. We note that much additional work was performed as part of the Fellowship. One salient project is the development of the Trilinos-based model-reduction software module Razor , which is currently bundled with the Albany PDE code and currently allows nonlinear reduced-order models to be constructed for any application supported in Albany. Other important projects include the following: 1. ROMES-equipped ROMs for Bayesian inference: K. Carlberg, M. Drohmann, F. Lu (Lawrence Berkeley National Laboratory), M. Morzfeld (Lawrence Berkeley National Laboratory). 2. ROM-enabled Krylov-subspace recycling: K. Carlberg, V. Forstall (University of Maryland), P. Tsuji, R. Tuminaro. 3. A pseudo balanced POD method using only dual snapshots: K. Carlberg, M. Sarovar. 4. An analysis of discrete v. continuous optimality in nonlinear model reduction: K. Carlberg, M. Barone, H. Antil (George Mason University). Journal articles for these projects are in progress at the time of this writing.

  4. A dynamic event tree informed approach to probabilistic accident sequence modeling: Dynamics and variabilities in medium LOCA

    International Nuclear Information System (INIS)

    In Probability Safety Assessments, accident scenario dynamics are addressed in the accident sequence analysis task. In an analyst-driven, iterative process, assumptions are made about equipment responses and operator actions and simulations of the scenario evolution are performed. To calculate how scenario dynamics and stochastic variabilities may affect the results of this process in terms of estimated risk, this work applies Dynamic Event Trees (DETs) to more comprehensively examine the accident scenario space. Alternative event tree models are developed and the core damage frequency is quantified to reveal the effects of different delineations of the sequences and of the bounding assumptions underlying success criteria. The results from a case study on Medium-break Loss of Coolant Accident scenarios in a Pressurized Water Reactor are presented, considering the break size, available injection trains, and the timing of rapid cooldown and the switchover to recirculation. The results show not only that estimated risk can be very sensitive to the numerous assumptions made in current accident sequence analysis but also that bounding assumptions do not always result in conservative risk estimates, thereby confirming the benefits that DETs provide in terms of characterizing scenario dynamics. - Highlights: • The overall most challenging MLOCA break is at neither extreme of the size range. • Selecting the limiting break size influenced estimated risk strongly (6″ vs 7″). • Success criteria can be defined more realistically by splitting the MLOCA range. • A more demanding success criterion for one top event can reduce overall risk. • Non-limiting success branches may lead to more demanding subsequent success criteria

  5. Co-breaking: Recent Advances and a Synopsis of the Literature.

    OpenAIRE

    David F. Hendry; Massmann, Michael

    2007-01-01

    This article has two aims. First, we provide a synopsis of the literature on co-breaking that has developed in several, seemingly disconnected, strands. We establish a consistent terminology, collect theoretical results, delimit co-breaking to cointegration and common features, and review recent contributions to co-breaking regressions and the budding analysis of co-breaking rank. Second, we present new results in the field, particularly, on the importance of co-breaking for policy analysis, ...

  6. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  7. Realistic methods for calculating the releases and consequences of a large LOCA

    International Nuclear Information System (INIS)

    This report describes a calculational route to predict realistic radiological consequences for a successfully terminated large-loss-of-coolant accident (LOCA) at a pressurized-water reactor (PWR). All steps in the calculational route are considered. For each one, a brief comment is made on the significant differences between the methods of calculation that were identified in the benchmark studies and recommendations are made for the methods and data for carrying out realistic calculations. These are based on the best supportable methods and data and the technical basis for each recommendation is given. Where the lack of well-validated methods or data means that the most realistic method that can be justified is considered to be very conservative, the need for further research is identified. The behaviour of inorganic iodine and the removal of aerosols from the atmosphere of the reactor building are identified as areas of particular importance. Where the retention of radioactivity is sensitive to design features, these are identified and, for the most importance features, the impact of different designs on the release of activity is indicated. The predictions of the proposed model are calculated for each stage and compared with the releases of activity predicted by the licensing methods that were used in the earlier benchmark studies. The conservative nature of the latter is confirmed. Methods and data are also presented for calculating the resulting doses to members of the public of the National Radiological Protection Boards as a result of work carried out by several national bodies in the UK. Other, equally acceptable, models are used in other countries of the Community and some examples are given

  8. Breaking News as Radicalisation

    DEFF Research Database (Denmark)

    Hartley, Jannie Møller

    The aim of the paper is to make explicit how the different categories are applied in the online newsroom and thus how new categories can be seen as positioning strategies in the form of radicalisations of already existing categories. Thus field theory provides us with tools to analyse how online...... journalists are using the categorisations to create hierarchies within the journalistic field in order to position themselves as specialists in what Tuchman has called developing news, aiming and striving for what today is know as breaking news and the “exclusive scoop,” as the trademark of online journalism...... provides us with the following two research questions: How does the category of breaking news fit into Tuchmans typology related to time, planning and technology? What types of stories are providing journalistic capital and how are online news stories categorised relatively within the journalistic field?...

  9. Single sector supersymmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Luty, Markus A.; Terning, John

    1999-03-18

    We review recent work on realistic models that break supersymmetry dynamically and give rise to composite quarks and leptons, all in a single sector. These models have a completely natural suppression of flavor-changing neutral currents, and the hierarchy of Yukawa couplings is explained by the dimensionality of composite states. The generic signatures are unification of scalar masses with different quantum numbers at the compositeness scale, and lighter gaugino, Higgsino, and third-generation sfermion masses.

  10. Routinizing Breaking News

    DEFF Research Database (Denmark)

    Hartley, Jannie Møller

    2011-01-01

    This chapter revisits seminal theoretical categorizations of news proposed three decades earlier by US sociologist Gaye Tuchman. By exploring the definition of ”breaking news” in the contemporary online newsrooms of three Danish news organisations, the author offers us a long overdue re......-theorization of journalistic practice in the online context and helpfully explores well-evidenced limitations to online news production, such as the relationship between original reporting and the use of ”shovelware.”...

  11. Predicting appointment breaking.

    Science.gov (United States)

    Bean, A G; Talaga, J

    1995-01-01

    The goal of physician referral services is to schedule appointments, but if too many patients fail to show up, the value of the service will be compromised. The authors found that appointment breaking can be predicted by the number of days to the scheduled appointment, the doctor's specialty, and the patient's age and gender. They also offer specific suggestions for modifying the marketing mix to reduce the incidence of no-shows. PMID:10142384

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

    International Nuclear Information System (INIS)

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

  13. Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

    International Nuclear Information System (INIS)

    In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

  14. An analysis of ROSA-IV/LSTF 10% main steam line break test run SB-SL-01 using RELAP5/MOD3

    International Nuclear Information System (INIS)

    This paper presents RELAP5/MOD3 code calculations of a 10% main steam line break test, designated as RUN SB-SL-01, conducted using the ROSA-4 Large Scale Test Facility (LSTF). The RELAP5/MOD3 input deck of LSTF, which includes 189 volumes, 200 junctions, and 180 heat slabs, was modeled to obtain best-estimate predictions of several important features during the main steam line break accident in order to property evaluate the consequences of this accident. The main conclusions drawn were that the results of RELAP5/MOD3 code calculations were in reasonable agreements with test RUN SB-SL-01, especially for the trends of key parameters. Detailed investigations indicated minor discrepancies in RCS pressure during the period of time that voiding occurred in the upper head. This is possible due to emptying of the pressurizer and voiding in the upper head. Sensitivity studies were also performed for the break junction discharge coefficient and the separator drain line loss coefficient. These parameters had significant effects on the steam quality on the secondary side and on the break flow through the change of water inventory on the secondary side. This phase separation process was adequately predicted during all transients with break junction discharge coefficient of 0.85 and separator drain line loss coefficient of 10

  15. An application of the LOCA code Cupidon: an assessment of the cladding behaviour in the flash tests

    International Nuclear Information System (INIS)

    The Cupidon code has been developed to analyze the thermo-mechanical behaviour of a fuel rod during a Loss Of Coolant Experiment. Models included are drawn from out-of-pile results such Edgar and the first use is to predict and calculate the tests carried out in the Phebus facility. Although the Flash program initiated at Grenoble is devoted to the study of fission product release during a LOCA (Loss Of Coolant Accident), interesting informations have been obtained on in-pile cladding deformation during transients. Analyses of the PIE (Post Irradiated Examination) results in the two first experiments with Cupidon code have shown fairly good agreement regarding diametral strain

  16. Cuerpos en revuelta: la loca y el militante del deseo desde la mirada Néstor Perlongher

    OpenAIRE

    Cohendoz, Mónica

    2014-01-01

    La poética de Néstor Perlongher (1949-1992) y el cuerpo antipatriarcal que figura para poner en crisis el cuerpo social son los dos temas recortados para analizar las prácticas comunicacionales que producen su experiencia estética. Los problemas que se han de considerar son: en primera instancia, el discurso de los cuerpos en lucha que Néstor Perlongher produce tanto con sus textos como con su militancia de género; en segundo lugar, se abordarán las imágenes de la loca y del militante del des...

  17. Risk Analysis of an interfacing system LOCA in a generic Westinghouse PWR

    OpenAIRE

    Favre, Jean-Baptiste

    2014-01-01

    This project has been developed during my internship in the field of Probabilistic Safety Assessment (PSA) in the offices of Westinghouse. These are in the enclosure of Vandellòs’ Nuclear Plant in Hospitalet de l'Infant. The goal of my internship was the modelling and computation of the frequency that an Interfacing System Loss of Coolant Accident (ISLOCA) occurs in the nuclear power plants of Vandellòs and Ascó. In order to achieve this goal, I applied a method to calculate...

  18. Best-estimate analysis and decision making under uncertainty

    International Nuclear Information System (INIS)

    In many engineering analyses of system safety the traditional reliance on conservative evaluation model calculations is being replaced with so called best-estimate analysis. These best-estimate analyses differentiate themselves from the traditional conservative analyses through two ingredients, namely realistic models and an account of the residual uncertainty associated with the model calculations. Best-estimate analysis, in the context of this paper, refers to the numerical evaluation of system properties of interest in situations where direct confirmatory measurements are not feasible. A decision with regard to the safety of the system is then made based on the computed numerical values of the system properties of interest. These situations generally arise in the design of systems that require computed and generally nontrivial extrapolations from the available data. In the case of nuclear reactors, examples are criticality of spent fuel pools, neutronic parameters of new advanced designs where insufficient material is available for mockup critical experiments and, the large break loss of coolant accident (LOCA). In this paper the case of LOCA, is taken to discuss the best-estimate analysis and decision making. Central to decision making is information. Thus, of interest is the source, quantity and quality of the information obtained in a best-estimate analysis, and used to define the acceptance criteria and to formulate a decision rule. This in effect expands the problem from the calculation of a conservative margin to a predefined acceptance criterion, to the formulation of a consistent decision rule and the computation of a test statistic for application of the decision rule. The latter view is a necessary condition for developing risk informed decision rules, and, thus, the relation between design basis analysis criteria and probabilistic risk assessment criteria is key. The discussion is in the context of making a decision under uncertainty for a reactor

  19. Breaking the Silence

    Institute of Scientific and Technical Information of China (English)

    1998-01-01

    Many women who suffer from vaginitis have kept silent about their illness because they think it is shameful to have such a disease. The International Women’s Health Coalition (IWHC) has publicized the problem, referring to it as a "culture of silence"inherited from traditional thinking. The coalition has made attempts to improve women’s health conditions by changing people’s misconceptions about the disease. In 1997, under a grant from the American Ford Foundation, the Sichuan Provincial Women’s Federation carried out a study on women’s repro-ductive health, aimed at "breaking the silence."

  20. Symmetries and Symmetry Breaking

    CERN Document Server

    Van Oers, W T H

    2003-01-01

    In understanding the world of matter, the introduction of symmetry principles following experimentation or using the predictive power of symmetry principles to guide experimentation is most profound. The conservation of energy, linear momentum, angular momentum, charge, and CPT involve fundamental symmetries. All other conservation laws are valid within a restricted subspace of the four interactions: the strong, the electromagnetic, the weak, and the gravitational interaction. In this paper comments are made regarding parity violation in hadronic systems, charge symmetry breaking in two nucleon and few nucleon systems, and time-reversal-invariance in hadronic systems.

  1. Breaking the silence

    DEFF Research Database (Denmark)

    Konradsen, Hanne; Kirkevold, Marit; McCallin, Antoinette;

    2012-01-01

    and individual interviews were analyzed using the grounded theory method. The findings revealed that the main concern of the patients was feeling isolated, which was resolved using a process of interactional integration. Interactional integration begins by breaking the silence to enable the progression from......Little is known about the psychosocial effects of facial disfigurement. We present the results of a qualitative study following 15 patients who had been surgically treated for head, neck, or eye cancer over the course of their first postoperative year. Taped nurse-patient conversations...

  2. RELAP5/MOD1.5 analysis of steam line break transients for a 3-loop and a 4-loop Westinghouse nuclear steam supply system

    International Nuclear Information System (INIS)

    RELAP5/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients. In developing the 4-loop plant model, a three loop W plant model was modified at ANL into a four loop plant. This plant model retained the two loop nodalization of the original model, but split them such that one loop represented the faulted steam generator (broken at the outlet just past the integral flow restrictor) while the other loop represented all three of the intact loops. In the case of the 3-loop plant model, a RELAP5 model was modified to perform this replication calculation. One loop represented the faulted steam generator (broken between the steam generator and the non-integral flow restrictor) while the other loop represented the remaining two intact loops. In both cases (3-loop and 4-loop), the reactor vessel was modelled as two parallel channels to accommodate the Westinghouse steam line break methodology

  3. Bootstrap Dynamical Symmetry Breaking

    Directory of Open Access Journals (Sweden)

    Wei-Shu Hou

    2013-01-01

    Full Text Available Despite the emergence of a 125 GeV Higgs-like particle at the LHC, we explore the possibility of dynamical electroweak symmetry breaking by strong Yukawa coupling of very heavy new chiral quarks Q . Taking the 125 GeV object to be a dilaton with suppressed couplings, we note that the Goldstone bosons G exist as longitudinal modes V L of the weak bosons and would couple to Q with Yukawa coupling λ Q . With m Q ≳ 700  GeV from LHC, the strong λ Q ≳ 4 could lead to deeply bound Q Q ¯ states. We postulate that the leading “collapsed state,” the color-singlet (heavy isotriplet, pseudoscalar Q Q ¯ meson π 1 , is G itself, and a gap equation without Higgs is constructed. Dynamical symmetry breaking is affected via strong λ Q , generating m Q while self-consistently justifying treating G as massless in the loop, hence, “bootstrap,” Solving such a gap equation, we find that m Q should be several TeV, or λ Q ≳ 4 π , and would become much heavier if there is a light Higgs boson. For such heavy chiral quarks, we find analogy with the π − N system, by which we conjecture the possible annihilation phenomena of Q Q ¯ → n V L with high multiplicity, the search of which might be aided by Yukawa-bound Q Q ¯ resonances.

  4. Generating debate on issues surrounding the venting of containment in the event of LOCA to ensure an optimised safe outcome

    International Nuclear Information System (INIS)

    Following the incidents at the Japanese Fukushima Nuclear facility, when three units experienced LOCA, the consequences of those events have caused ripples across the world. Regulators around the world are examining the need to prevent excessive pressurisation of NPP Containment, and safely evacuate the gaseous consequences of LOCA, there is a need to return to fundamental principles and examine each putative set of data derived from the various models describing the consequence of LOCA and other events, a LOCA being a 'Loss Of Coolant Accident', and represents the outcome from a range of incidents ranging from fuel pellets being exposed to cooling (heat exchange) water, to a complete melt down of the fuel load with consequence evaporation of the concrete structures of the reactor containment buildings. Clearly it would be highly desirable in both economic and safety terms to minimise the external impact of these events inside the containment. Since one of the results of a LOCA is the pressurisation of the internal space of the containment building, regulators and the industry are looking at the mandatory installation of suitable vent systems to vent the pressure building up inside the containment, whilst ensuring the minimum impact on the surrounding environment. This is clearly a filtration/separation/recombination issue, and as an expert engineering company in the nuclear industry, Porvair has concerns that the issues of Containment Venting are not being addressed by expert filter companies, but by expert nuclear engineering companies with only a superficial knowledge of the complexities and nuances on filtration processes. The paper will describe in depth and detail the individual consequences of each particular aspect of the modelled data. Looking at flowing conditions (pressure, temperature, gas constituents including water vapour and Caesium and Iodine compounds), vent pressure philosophy, deposition of solids (size, type and quantity), decay heat

  5. Report of Break Out Group 1

    DEFF Research Database (Denmark)

    Alward, Randy; Carley, Kathleen M.; Madsen, Fredrik Huitfeldt;

    2006-01-01

    , action" (OODA) loop. The break out group discussed vulnerability presentation needs common across various application domains, particularly in support of network discovery and network analysis tasks in those domains. Finally, the break out group wished to determine whether there is a means...... of characterizing a vulnerability. This would take into account the potential for the vulnerability to be exploited as well as the potential impact on the operations supported by the network, and on the network structure itself, of a successful exploit of that vulnerability....

  6. Chiral symmetry and chiral-symmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Peskin, M.E.

    1982-12-01

    These lectures concern the dynamics of fermions in strong interaction with gauge fields. Systems of fermions coupled by gauge forces have a very rich structure of global symmetries, which are called chiral symmetries. These lectures will focus on the realization of chiral symmetries and the causes and consequences of thier spontaneous breaking. A brief introduction to the basic formalism and concepts of chiral symmetry breaking is given, then some explicit calculations of chiral symmetry breaking in gauge theories are given, treating first parity-invariant and then chiral models. These calculations are meant to be illustrative rather than accurate; they make use of unjustified mathematical approximations which serve to make the physics more clear. Some formal constraints on chiral symmetry breaking are discussed which illuminate and extend the results of our more explicit analysis. Finally, a brief review of the phenomenological theory of chiral symmetry breaking is presented, and some applications of this theory to problems in weak-interaction physics are discussed. (WHK)

  7. An Analysis of the Interactive Development between Lhasa and Shigatse from the Perspective of the Intensity of their Economic Contact and Regional Breaking Point

    Institute of Scientific and Technical Information of China (English)

    QIU Lilan

    2014-01-01

    In the Tibet Autonomous Region , Lhasa and Shigatse are the two core cities in the region’ s development , and they play a leading role for the development of other counties and cit-ies.From the perspective of research on urban ag-glomeration and urban geography , the two cities have a relationship of interactive contact and com-mon development .Within the context of rapid ur-banization in China , the interactive development between Lhasa and Shigatse has become an inevita-ble research focus for the socio-economic construc-tion of Tibet . The intensity of economic contact is used tomeasure the degree of regional economic ties.Onthe one hand, it can reflect the city’ s economiccenter ability to radiate out to the surrounding areas.On the other hand, it can also reflect the levelof acceptance of the surrounding areas to the city’s economic center ability to radiate out tothem.According to the results of economic calculation,the intensity of economic contact betweenLhasa and Shigatse from 2007 to 2011 steadily increasedwith an annual rate of about 20%.Thecontinuous increase of the intensity of economiccontact also meant that the influence of Lhasa andShigatse on the surrounding cities is expanding .Moreover, comparative data show that due to restrictionson the population and level of economicdevelopment, there is still a big gap in Lhasa andShigatse if compared with other urbanized areas inChina.However, regarding the increasing speedof the intensity of contact with other core cities,Lhasa and Shigatse have achieved remarkable a -chievements. During the 1930’ s, William J.Reilly proposedthe Law of Retail Gravitation which was usedby P.D.Converse in the field of urban managementto differentiate the scope of influence ofneighboring cities.In 1949, he put forward theconcept of “Breaking Point” together with a relevantcomputational formula.The Breaking Point isnormally determined by the scale of two cities andthe distance between them.According to the

  8. Experimental Study on Coal-Breaking Pressure for Compressed Air

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Based on lab model experiments and through the limit analysis, the theoretical formula of calculauoncoal-breaking pressure with compressed air was derived. The experimental result shows that blasting pressure mustexceed 84.0 MPa to break coal with compressive strength of 13.2 MPa. The research provides an important theoretical basis for the design of airshooting mining and industrial tests.

  9. Break the Pattern!

    DEFF Research Database (Denmark)

    Hasse, Cathrine; Trentemøller, Stine

    Break the Pattern! A critical enquiry into three scientific workplace cultures: Hercules, Caretakers and Worker Bees is the third publication of the international three year long project "Understanding Puzzles in the Gendered European Map" (UPGEM). By contrasting empirical findings from academic...... workplaces in the five UPGEM-countries (Denmark, Estonia, Finland, Italy and Poland) we identify three clusters of cultural patterns in physics as culture. We call these Hercules, Caretakers and Worker Bees. We also consider the influence of national cultural historical processes on the scientific culture...... (physics in culture) and discuss how physics as and in culture influence the perception of science, of work and family life, of the interplay between religion and science as well as how physics as culture can either hinder or promote the career of female scientists....

  10. Violent breaking wave impacts

    DEFF Research Database (Denmark)

    Bredmose, Henrik; Peregrine, D.H.; Bullock, G.N.

    2009-01-01

    When an ocean wave breaks against a steep-fronted breakwater, sea wall or a similar marine structure, its impact on the structure can be very violent. This paper describes the theoretical studies that, together with field and laboratory investigations, have been carried out in order to gain...... a better understanding of the processes involved. The wave's approach towards a structure is modelled with classical irrotational flow to obtain the different types of impact profiles that may or may not lead to air entrapment. The subsequent impact is modelled with a novel compressible-flow model...... for a homogeneous mixture of incompressible liquid and ideal gas. This enables a numerical description of both trapped air pockets and the propagation of pressure shock waves through the aerated water. An exact Riemann solver is developed to permit a finite-volume solution to the flow model with smallest possible...

  11. Analytical model for estimating drag forces on rigid submerged structures caused by LOCA and safety relief valve ramshead air discharges. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    Basic information is presented for estimating drag forces on rigid structural members submerged in a pressure suppression pool, caused by either the air discharge from a loss-of-coolant accident (LOCA), or the air bubble oscillation following safey relief valve ramshead discharge. Methods are described for estimating acceleration (unsteady) and standard (velocity-squared) drag force components for a variety of structural geometries.

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

    International Nuclear Information System (INIS)

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

  13. A two-dimensional analysis of the sensitivity of a pulse first break to wave speed contrast on a scale below the resolution length of ray tomography.

    Science.gov (United States)

    Willey, Carson L; Simonetti, Francesco

    2016-06-01

    Mapping the speed of mechanical waves traveling inside a medium is a topic of great interest across many fields from geoscience to medical diagnostics. Much work has been done to characterize the fidelity with which the geometrical features of the medium can be reconstructed and multiple resolution criteria have been proposed depending on the wave-matter interaction model used to decode the wave speed map from scattering measurements. However, these criteria do not define the accuracy with which the wave speed values can be reconstructed. Using two-dimensional simulations, it is shown that the first-arrival traveltime predicted by ray theory can be an accurate representation of the arrival of a pulse first break even in the presence of diffraction and other phenomena that are not accounted for by ray theory. As a result, ray-based tomographic inversions can yield accurate wave speed estimations also when the size of a sound speed anomaly is smaller than the resolution length of the inversion method provided that traveltimes are estimated from the signal first break. This increased sensitivity however renders the inversion more susceptible to noise since the amplitude of the signal around the first break is typically low especially when three-dimensional anomalies are considered. PMID:27369139

  14. Electroweak breaking in supersymmetric models

    CERN Document Server

    Ibáñez, L E

    1992-01-01

    We discuss the mechanism for electroweak symmetry breaking in supersymmetric versions of the standard model. After briefly reviewing the possible sources of supersymmetry breaking, we show how the required pattern of symmetry breaking can automatically result from the structure of quantum corrections in the theory. We demonstrate that this radiative breaking mechanism works well for a heavy top quark and can be combined in unified versions of the theory with excellent predictions for the running couplings of the model. (To be published in ``Perspectives in Higgs Physics'', G. Kane editor.)

  15. Comet assay analysis of repair of DNA strand breaks in normal and deficient human cells exposed to radiations and chemicals. Evidence for a repair pathway specificity of DNA ligation

    International Nuclear Information System (INIS)

    The induction and resealing of DNA strand breaks in a cell line with a proven defect in DNA ligase I, 46BR, and in two Bloom's syndrome cell lines. YBL6 and GM 1492, were compared to those observed in normal human 1BR/3 fibroblasts after treatment with a variety of genotoxic agents whose lesions are processed by different repair pathways. This analysis was performed using the single-cell gel electrophoresis assay. The three types of cells were found to have similar capabilities to recognize and incise ultraviolet photoproducts and also demonstrated similar amounts of DNA breaks immediately after γ irradiation. During post-treatment incubation, 46BR cells showed a marked DNA re-ligation defect after ultraviolet radiation damage, GM 1492 cells demonstrated a highly reduced DNA joining ability after relatively high doses of ultraviolet radiation, and YBL6 cells were particularly affected in DNA re-ligation after damage by 4-nitroquinoline-1-oxide. The two Bloom's syndrome cell lines and 46BR cells had a nearly normal ability to reseal breaks resulting from γ irradiation or treatment with xanthine plus xanthine oxidase. These findings suggest that different DNA ligases may be involved in different DNA repair pathways in human cells. 60 refs., 7 figs

  16. Modeling LOCA performance for the generation IV gas-cooled fast reactor design

    International Nuclear Information System (INIS)

    Full text of publication follows: Generation IV nuclear energy systems are next-generation technologies that will offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance. Expected to be available for worldwide deployment by 2030, these energy systems would provide electrical power for the subsequent decades. The Gas-Cooled Fast Reactor (GFR) is a Generation IV concept that features a fast-neutron spectrum, direct Brayton cycle gas turbine, and a closed fuel cycle. Through the combination of a fast neutron spectrum and the full recycle of actinides, the GFR minimizes the production of long-lived radioactive waste and makes it possible to use existing fissile and fertile materials (including depleted uranium) more efficiently than existing thermal spectrum gas reactors. The prominent GFR design features a 'pancake' style core (H/D ∼ 1.7/2.9 m) that produces 600 MW of thermal power with an average power density of 55 MW/m3. The core is comprised of SiC-coated UPuC spheres that are collected in channels to form a prismatic, hexagonal fuel assembly or coagulated to form fuel pebbles. The 11 m3 core is enveloped by TiN reflectors and stainless steel shields in both the radial and axial directions. The initial GFR design used He gas at a pressure of 7 MPa and an outlet temperature of 850 deg. C, however the design has been expanded to consider supercritical CO2 (S-CO) gas at a pressure of 19 MPa and an outlet temperature of 550 - 650 deg. C. The higher density S-CO has advantageous characteristics during off-normal low flow and pressure conditions. One of the strengths of the Generation IV reactor concepts is their inherent safety and extensive use of passive safety systems. This paper discusses an analysis performed to study the GFR's response during a severe off-normal scenario. The loss of coolant accident was chosen because it will be one of the more severe challenges to the reactors decay heat removal system

  17. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  18. Validation methodology for the evaluation of thermal-hydraulic sub-channel codes devoted to LOCA simulations

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, N.; Ruyer, P.; Biton, B., E-mail: nathalie.seiler@irsn.fr, E-mail: pierre.ruyer@irsn.fr [IRSN/DPAM/SEMCA/LEMAR, CE Cadarache, Saint Paul lez Durance (France)

    2011-07-01

    This study focuses on thermal-hydraulic simulations, at sub-channel scale, of a damaged PWR reactor core during a Loss Of Coolant Accident (LOCA). The aim of this study is to accurately simulate the thermal-hydraulics to provide the thermal-mechanical code DRACCAR with an accurate wall heat transfer law. This latter code is developed by the French Safety Institute “Institut de Radioprotection et de Surete Nucleaire” (IRSN) to evaluate the thermics and deformations of fuel assemblies within the core. The present paper first describes the methodology considered to evaluate the capabilities of existing codes CATHARE-3 and CESAR to simulate dispersed droplet flows at a sub-channel scale and then provides some first evaluations of them. (author)

  19. Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

  20. Validation methodology for the evaluation of thermal-hydraulic sub-channel codes devoted to LOCA simulations

    International Nuclear Information System (INIS)

    This study focuses on thermal-hydraulic simulations, at sub-channel scale, of a damaged PWR reactor core during a Loss Of Coolant Accident (LOCA). The aim of this study is to accurately simulate the thermal-hydraulics to provide the thermal-mechanical code DRACCAR with an accurate wall heat transfer law. This latter code is developed by the French Safety Institute “Institut de Radioprotection et de Surete Nucleaire” (IRSN) to evaluate the thermics and deformations of fuel assemblies within the core. The present paper first describes the methodology considered to evaluate the capabilities of existing codes CATHARE-3 and CESAR to simulate dispersed droplet flows at a sub-channel scale and then provides some first evaluations of them. (author)

  1. Entanglement–breaking indices

    Energy Technology Data Exchange (ETDEWEB)

    Lami, L. [Scuola Normale Superiore, I-56126 Pisa (Italy); Giovannetti, V. [NEST, Scuola Normale Superiore and Istituto Nanoscienze-CNR, I-56127 Pisa (Italy)

    2015-09-15

    We study a set of new functionals (called entanglement–breaking indices) which characterize how many local iterations of a given (local) quantum channel are needed in order to completely destroy the entanglement between the system of interest over which the transformation is defined and an external ancilla. The possibility of contrasting the noisy effects introduced by the channel iterations via the action of intermediate (filtering) transformations is analyzed. We provide some examples in which our functionals can be exactly calculated. The differences between unitary and non-unitary filtering operations are analyzed showing that, at least for systems of dimension d larger than or equal to 3, the non-unitary choice is preferable (the gap between the performances of the two cases being divergent in some cases). For d = 2 (qubit case), on the contrary, no evidences of the presence of such gap is revealed: we conjecture that for this special case unitary filtering transformations are optimal. The scenario in which more general filtering protocols are allowed is also discussed in some detail. The case of a depolarizing noise acting on a two–qubit system is exactly solved in a general case.

  2. TIE breaking: Tunable interdomain egress selection

    OpenAIRE

    Teixeira, Renata; Griffin, Tim; G. C. Resende, Mauricio; Rexford, Jennifer

    2005-01-01

    International audience The separation of intradomain and interdomain routing has been a key feature of the Internet's routing architecture from the early days of the ARPAnet. However, the appropriate "division of labor" between the two protocols becomes unclear when an Autonomous System (AS) has interdomain routes to a destination prefix through multiple border routers—a situation that is extremely common to-day because neighboring domains often connect in several loca-tions. We believe th...

  3. Containment integrity analysis for the (W) advanced AP900

    International Nuclear Information System (INIS)

    The consequences of accidents that occurred at Three Mile Island and Chernobyl and the lack of interest from utilities in committing to high powered, high maintenance light water reactor designs of the past have motivated the nuclear industry to develop smaller, lower powered, simplified designs. Simplified designs must employ more passive systems that require very low maintenance and rely on the basic laws of nature. These designs must not be limited only to the reactor, but must also include the containment building. The purpose of this paper is to discuss the design concepts and to verify the feasibility of the Westinghouse Passive Containment Cooling System (PCCS) design for a 900 MWe advanced passive reactor design (AP900) by presenting its response to a design basis large break Loss of Coolant Accident (LOCA). The mass and energy releases emanating from the design basis break were developed and a representative containment model was constructed. The COMPACT computer code, which is a multi-node program that solves the complete set of mass, energy and momentum equations, was used to determine the AP900 containment response. The PCCS design concept, the computer model, the analysis inputs, and the results of the study are discussed within this paper. (author). 2 refs., 9 figs

  4. Fragmentation in DNA double-strand breaks

    International Nuclear Information System (INIS)

    DNA double strand breaks are important lesions induced by irradiations. Random breakage model or quantification supported by this concept is suitable to analyze DNA double strand break data induced by low LET radiation, but deviation from random breakage model is more evident in high LET radiation data analysis. In this work we develop a new method, statistical fragmentation model, to analyze the fragmentation process of DNA double strand breaks. After charged particles enter the biological cell, they produce ionizations along their tracks, and transfer their energies to the cells and break the cellular DNA strands into fragments. The probable distribution of the fragments is obtained under the condition in which the entropy is maximum. Under the approximation E≅E0 + E1l + E2l2, the distribution functions are obtained as exp(αl + βl2). There are two components, the one proportional to exp(βl2), mainly contributes to the low mass fragment yields, the other component, proportional to exp(αl), decreases slowly as the mass of the fragments increases. Numerical solution of the constraint equations provides parameters α and β. Experimental data, especially when the energy deposition is higher, support the statistical fragmentation model. (authors)

  5. Chemical processes of galvanized steel corrosion in the post-LOCA phase of a PWR and the prevention of sump screen clogging

    International Nuclear Information System (INIS)

    The Emergency Core Coolant System has to remove the decay heat in case of a Loss of Coolant Accident (LOCA). Therefore, the emergency core cooling pumps recirculate the fluid from the sump back into the primary circuit. Sump strainers are mounted at the pump inlets to retain particles and fibrous insulation material. A fiber bed formed on strainers may act as an additional debris filter. However, a critical increase of pressure drop generated by debris or corrosion products could cause a failure of emergency cooling. Problems of insulation materials NUKONR (fiberglass) or CalSil and Aluminium may appear if containment spray systems using alkaline additives are installed. In such cases, dissolution / precipitation reactions resulting from insulation materials were observed, which increase the risk of sump screen blockage. In German NPPs, there are no containments spray systems, and insulation consists of more resistant materials like mineral wool (rock wool) and stainless steel. However, large scale experiments from AREVA have shown that sump screen clogging may be initiated by boric acid containing For generic investigations of galvanized steel corrosion behaviour under post-LOCA conditions, the down-scaled test facility KorrVA was designed consisting of a loop with trickle section (location of LOCA), bath section (sump), horizontal strainer and circulation pump. The low coolant volume (60 L) permits an easy and efficient purification between the experiments including complete removal of corrosion products. About 90 experiments were carried out with galvanized steel gratings and galvanized steel coupons in boric acid media in order to determine corrosion mechanisms depending on different experimental conditions like temperature, water chemistry and hydrodynamic conditions (flow impact, simulated by different nozzles). Practically, the fiber bed was prepared during a preliminary stage with the aim to separate effects of fiber bed formation on sump strainer clogging

  6. Analysis on Flow Characteristics of a Two-dimensional Rectangular Water Column during Dam Breaking%二维矩形水柱溃坝时的水流特性分析

    Institute of Scientific and Technical Information of China (English)

    王茂运; 陈祥喜

    2016-01-01

    由于溃坝失事后果的严重性,文章研究了溃坝过程中的水流特性,建立了二维矩形水柱在重力作用下瞬间全溃的数学模型,基于无量纲数分析,运用数值模拟的方法对比了深宽比在1:1~8:1时不同初始水位矩形水体坝前残余水深在溃坝过程中的变化规律。结果表明:矩形水柱在溃坝过程中上游残余水深的变化规律具有一致性,并推导得到了基于深宽比的坝前残余水深函数随时间变化的统一公式,该公式在特征时间小于6,深宽比1~8时具有较高的精度。%Because of the seriousness of the consequences of dam-break, this paper studies the flow characteristics during the process of dam-break, and establishes the mathematical model of dam-break in a two-dimensional rectangular water column under gravity. The rule of residual water depth during the process of dam-break in a rectangular water column is compared at the ratio of 1∶1 to 8∶1 with the method of numerical simulation based on dimensional analysis. The results shows that the rule of residual water depth during the process of dam-break in a rectangular water column is consistent, and a uniform formula of the residual water depth in the upstream depending on time was derived based on the aspect ratio. The formula has a high accuracy when characteristic time is less than 6, and aspect ratio is 1~8.

  7. Analysis of economic break-even point of the biogas utilization for electrical power conversion: case study in a swine terminated unit

    International Nuclear Information System (INIS)

    This work aimed to develop a study to estimate the break-even point in financial units of the electrical power generation using biogas from swine wastes. The analyzed biodigester is a continuous tubular model with brick concrete duct and plastic covering with a gasometer, and where the waste of 2,300 fattening pigs are deposited daily. The initial investment estimate for the installation was R$ 51,537.17. The system annual costs were R$ 5,708.20, for maintenance, R$ 4,390.40 for depreciation and R$ 1,366.77 for interests. It was noticed that with an average of consumption of 17.1 kW.hour-1 the system presents an annual loss of R$ 1,592.14 because the consumption of 27.85 kW.hour-1 is the minimum that should be consumed to achieve a corresponded financial break-even point of R$ 15,054.40.year-1. It was concluded that the correct technical dimensioning greatly influences on the economic results. (author)

  8. Detecting multiple breaks in geodetic time series using indicator saturation.

    Science.gov (United States)

    Jackson, Luke; Pretis, Felix

    2016-04-01

    Identifying the timing and magnitude of breaks in geodetic time series has been the source of much discussion. Instruments recording different geophysical phenomena may record long term trends, quasi-periodic signals at a variety of time scales from days to decades, and sudden breaks due to natural or anthropogenic causes, ranging from instrument replacement to earthquakes. Records can not always be relied upon to be continuous in time, yet one may desire to accurately bridge gaps without performing interpolation. We apply the novel Indicator Saturation (IS) method to identify breaks in a synthetic GPS time series used for the Detection of Offsets in GPS Experiments (DOGEX). The IS approach differs from alternative break detection methods by considering every point in the time series as a break, until it is demonstrated statistically that it is not. Saturating a model with a full set of break functions and removing all but significant ones, formulates the detection of breaks as a problem of model selection. This allows multiple breaks of different forms (from impulses, to shifts in the mean, and changing trends) without requiring a minimum break-length to be detected, while simultaneously modelling any underlying variation driven by additional covariates. To address selection bias in the coefficients, we demonstrate the bias-corrected estimates of break coefficients when using step-shifts in the mean of the modelled time-series. The regimes of the time-varying mean of the time-series (the `coefficient path' of the intercept determined by the detected breaks) can be used to conduct hypothesis tests on whether subsequent shifts offset each other - for example whether a measurement change induces a temporary bias rather than a permanent one. We explore this non-classical analysis method to see if it can bring about the sub millimetre errors in long term rates of land motion currently required by the GPS community.

  9. Detecting structural breaks in time series via genetic algorithms

    DEFF Research Database (Denmark)

    Doerr, Benjamin; Fischer, Paul; Hilbert, Astrid;

    2016-01-01

    Detecting structural breaks is an essential task for the statistical analysis of time series, for example, for fitting parametric models to it. In short, structural breaks are points in time at which the behaviour of the time series substantially changes. Typically, no solid background knowledge...... of the time series under consideration is available. Therefore, a black-box optimization approach is our method of choice for detecting structural breaks. We describe a genetic algorithm framework which easily adapts to a large number of statistical settings. To evaluate the usefulness of different crossover...... operator alone. Moreover, we present a specific fitness function which exploits the sparse structure of the break points and which can be evaluated particularly efficiently. The experiments on artificial and real-world time series show that the resulting algorithm detects break points with high precision...

  10. Experimental Study of a Stoppage Natural Circulation during a Nuclear Heating Reactor LOCA

    Institute of Scientific and Technical Information of China (English)

    博金海; 张佑杰; 姜胜耀

    2001-01-01

    The 5MW nuclear heating reactor is an integral naturalcirculation reactor. The rupture of the coolant penetrating tube is a typical accident causing coolant loss. When the water level drops down to the upper edge of the inlet of the heat exchanger, the natural circulation stops. This influences the core cooling and the stability of the main loop. A series of tests showed that there is a stable drop of pressure, and the heated element temperature is not too high to cause burnout. But the backward flow or flow oscillation in the primary coolant circuit occurs when the flow breaks completely in the end. The whole flow process is described and the mechanism is discussed.

  11. Soft branes in supersymmetry-breaking backgrounds

    OpenAIRE

    McGuirk, Paul; Shiu, Gary; Ye, Fang

    2012-01-01

    We revisit the analysis of effective field theories resulting from non-supersymmetric perturbations to supersymmetric flux compactifications of the type-IIB superstring with an eye towards those resulting from the backreaction of a small number of anti-D3-branes. Independently of the background, we show that the low-energy Lagrangian describing the fluctuations of a stack of probe D3-branes exhibits soft supersymmetry breaking, despite perturbations to marginal operators that were not fully c...

  12. Relationship between Age and the Ability to Break Scored Tablets

    Science.gov (United States)

    Notenboom, Kim; Vromans, Herman; Schipper, Maarten; Leufkens, Hubert G. M.; Bouvy, Marcel L.

    2016-01-01

    Background: Practical problems with the use of medicines, such as difficulties with breaking tablets, are an often overlooked cause for non-adherence. Tablets frequently break in uneven parts and loss of product can occur due to crumbling and powdering. Health characteristics, such as the presence of peripheral neuropathy, decreased grip strength and manual dexterity, can affect a patient's ability to break tablets. As these impairments are associated with aging and age-related diseases, such as Parkinson's disease and arthritis, difficulties with breaking tablets could be more prevalent among older adults. The objective of this study was to investigate the relationship between age and the ability to break scored tablets. Methods: A comparative study design was chosen. Thirty-six older adults and 36 young adults were systematically observed with breaking scored tablets. Twelve different tablets were included. All participants were asked to break each tablet by three techniques: in between the fingers with the use of nails, in between the fingers without the use of nails and pushing the tablet downward with one finger on a solid surface. It was established whether a tablet was broken or not, and if broken, whether the tablet was broken accurately or not. Results: The older adults experienced more difficulties to break tablets compared to the young adults. On average, the older persons broke 38.1% of the tablets, of which 71.0% was broken accurately. The young adults broke 78.2% of the tablets, of which 77.4% was broken accurately. Further analysis by mixed effects logistic regression revealed that age was associated with the ability to break tablets, but not with the accuracy of breaking. Conclusions: Breaking scored tablets by hand is less successful in an elderly population compared to a group of young adults. Health care providers should be aware that tablet breaking is not appropriate for all patients and for all drugs. In case tablet breaking is unavoidable, a

  13. Calculation of Departure from Nucleate Boiling Ratio (DNBR) minimum for accident analysis of main steam line break at Angra-1; Calculo do minimo DNBR para analise do acidente de ruptura da linha principal de vapor em Angra-1

    Energy Technology Data Exchange (ETDEWEB)

    Machado, Marcio Dornellas [ELETROBRAS Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil). E-mail: mdorne@eletronuclear.gov.br

    2000-07-01

    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)

  14. Analysis of Repair Mechanisms following an Induced Double-Strand Break Uncovers Recessive Deleterious Alleles in the Candida albicans Diploid Genome

    Science.gov (United States)

    Feri, Adeline; Loll-Krippleber, Raphaël; Commere, Pierre-Henri; Maufrais, Corinne; Sertour, Natacha; Schwartz, Katja; Sherlock, Gavin; Bougnoux, Marie-Elisabeth

    2016-01-01

    ABSTRACT The diploid genome of the yeast Candida albicans is highly plastic, exhibiting frequent loss-of-heterozygosity (LOH) events. To provide a deeper understanding of the mechanisms leading to LOH, we investigated the repair of a unique DNA double-strand break (DSB) in the laboratory C. albicans SC5314 strain using the I-SceI meganuclease. Upon I-SceI induction, we detected a strong increase in the frequency of LOH events at an I-SceI target locus positioned on chromosome 4 (Chr4), including events spreading from this locus to the proximal telomere. Characterization of the repair events by single nucleotide polymorphism (SNP) typing and whole-genome sequencing revealed a predominance of gene conversions, but we also observed mitotic crossover or break-induced replication events, as well as combinations of independent events. Importantly, progeny that had undergone homozygosis of part or all of Chr4 haplotype B (Chr4B) were inviable. Mining of genome sequencing data for 155 C. albicans isolates allowed the identification of a recessive lethal allele in the GPI16 gene on Chr4B unique to C. albicans strain SC5314 which is responsible for this inviability. Additional recessive lethal or deleterious alleles were identified in the genomes of strain SC5314 and two clinical isolates. Our results demonstrate that recessive lethal alleles in the genomes of C. albicans isolates prevent the occurrence of specific extended LOH events. While these and other recessive lethal and deleterious alleles are likely to accumulate in C. albicans due to clonal reproduction, their occurrence may in turn promote the maintenance of corresponding nondeleterious alleles and, consequently, heterozygosity in the C. albicans species. PMID:27729506

  15. Supersymmetry Breaking in Warped Geometry

    OpenAIRE

    Choi, Kiwoon; Kim, Do Young; Kim, Ian-Woo; Kobayashi, Tatsuo

    2003-01-01

    We examine the soft supersymmetry breaking parameters in supersymmetric theories on a slice of AdS_5 which generate the hierarchical Yukawa couplings by dynamically localizing the bulk matter fields in extra dimension. Such models can be regarded as the AdS dual of the recently studied 4-dimensional models which contain a supersymmetric CFT to generate the hierarchical Yukawa couplings. In such models, if supersymmetry breaking is mediated by the bulk radion superfield and/or some brane chira...

  16. Proton-Induced X-ray Emission (PIXE Analysis and DNA-chain Break study in rat hepatocarcinogenesis: A possible chemopreventive role by combined supplementation of vanadium and beta-carotene

    Directory of Open Access Journals (Sweden)

    Kanjilal NB

    2005-05-01

    Full Text Available Abstract Combined effect of vanadium and beta-carotene on rat liver DNA-chain break and Proton induced X-ray emission (PIXE analysis was studied during a necrogenic dose (200 mg/kg of body weight of Diethyl Nitrosamine (DENA induced rat liver carcinogenesis. Morphological and histopathological changes were observed as an end point biomarker. Supplementation of vanadium (0.5 ppm ad libitum in drinking water and beta-carotene in the basal diet (120 mg/Kg of body weight were performed four weeks before DENA treatment and continued till the end of the experiment (16 weeks. PIXE analysis revealed the restoration of near normal value of zinc, copper, and iron, which were substantially altered when compared to carcinogen treated groups. Supplementation of both vanadium and beta-carotene four weeks before DENA injection was found to offer significant (64.73%, P

  17. An automatic system for elaboration of chip breaking diagrams

    DEFF Research Database (Denmark)

    Andreasen, Jan Lasson; De Chiffre, Leonardo

    1998-01-01

    A laboratory system for fully automatic elaboration of chip breaking diagrams has been developed and tested. The system is based on automatic chip breaking detection by frequency analysis of cutting forces in connection with programming of a CNC-lathe to scan different feeds, speeds and cutting...... depths. An evaluation of the system based on a total of 1671 experiments has shown that unfavourable snarled chips can be detected with 98% certainty which indeed makes the system a valuable tool in chip breakability tests. Using the system, chip breaking diagrams can be elaborated with a previously...

  18. Detailed analysis of a severe accident progression for an evaluation of in-vessel corium retention estimation in KSNP

    International Nuclear Information System (INIS)

    An in-vessel melt progression and a plant damage state for the IVR evaluation of the KSNP for the high pressure cases of a total loss of feed water (LOFW) and a station blackout (SBO), and for the low pressure cases of loss of coolant accident (LOCA)s without a safety injection (SI) have been estimated using the SCDAP/RELAP5/MOD3.3 computer code. The pressure boundary of the reactor coolant system did not fail before a reactor vessel failure in the high-pressure sequences of the total LOFW and the SBO transients. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of a reactor vessel failure. In all transients of the low and high pressures, the melt temperature of the relocated corium in the lower plenum at a reactor vessel failure was approximately 2,850 K. The heat generation rates of the corium in the lower plenum at a reactor vessel failure in the total LOFW and SBO, the small and large break LOCAs, and the medium break LOCAs were approximately 2-2.7 MW/m3, 3 MW/m3, and 2 MW/m3, respectively

  19. An attempt for a unified description of mechanical testing on Zircaloy-4 cladding subjected to simulated LOCA transient

    Directory of Open Access Journals (Sweden)

    Desquines Jean

    2016-01-01

    Full Text Available During a Loss Of Coolant Accident (LOCA, an important safety requirement is that the reflooding of the core by the emergency core cooling system should not lead to a complete rupture of the fuel rods. Several types of mechanical tests are usually performed in the industry to determine the degree of cladding embrittlement, such as ring compression tests or four-point bending of rodlets. Many other tests can be found in the open literature. However, there is presently no real intrinsic understanding of the failure conditions in these tests which would allow translation of the results from one kind of mechanical testing to another. The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing. To achieve this objective, the failure mechanisms of pre-oxidized and pre-hydrided cladding samples are characterized by comparing the behavior of two different mechanical tests: Axial Tensile (AT test and “C”-shaped Ring Compression Test (CCT. The failure of samples in both cases can be described by usual linear elastic fracture mechanics theory. Using interrupted mechanical tests, metallographic examinations have evidenced that a set of parallel cracks are nucleated at the inner and outer surface of the samples just before failure, crossing both the oxide layer and the oxygen rich alpha layer. The stress intensity factors for multiple crack geometry are determined for both AT and CCT samples using finite element calculations. After each mechanical test performed on high temperature steam oxidized samples, metallography is then used to individually determine the crack depth and crack spacing. Using these two important parameters and considering the applied load at fracture, the stress intensity factor at failure is derived for each tested

  20. Effects of RCP trip when recovering HPSI during LOCA in a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Montero-Mayorga, Javier, E-mail: fj.montero@alumnos.upm.es; Queral, César; Rivas-Lewicky, Julio; González-Cadelo, Juan

    2014-12-15

    Highlights: • If HPSI is recovered during SBLOCA and RCPs are tripped core damage can be reached. • If the RCPs are tripped once the accumulators have injected the damage can be avoided. • If only 2 out of 3 RCPs are tripped the damage can be also avoided. • Improvements are proposed to the EOPs in order to avoid possible damage. - Abstract: Current Westinghouse Emergency Operating Procedures (EOPs) indicate initially that the operator must keep the reactor coolant pumps (RCPs) running during a Small Break Loss of Coolant Accident (SBLOCA) if there is unavailability of high pressure safety injection (HPSI) system in order to cool the core by forced convection. However, the crew must follow different EOPs along the transient depending on its evolution. In these EOPs there are several conditions which indicate the necessity of tripping one or more RCPs when HPSI is recovered. In this paper the occurrence of a SBLOCA with unavailability of HPSI has been analyzed with a model of Almaraz Nuclear Power Plant (Westinghouse 3 Loop) for TRACE code V5.0 patch 1. Two different approaches have been considered: the first one, taking into account Optimal Recovery Guidelines (ORGs) and in the second approach, the transition to Function Restoration Guidelines (FRGs) due to inadequate core cooling (ICC) conditions is considered. Results of this paper lead to the implementation of an improvement in current EOPs regarding how many RCPs should be tripped during SBLOCA sequences.

  1. Development Program of LOCA Licensing Calculation Capability with RELAP5-3D in Accordance with Appendix K of 10 CFR 50.46

    International Nuclear Information System (INIS)

    In light water reactors, particularly the pressurized water reactors, the severity of loss-of-coolant accidents (LOCAs) will limit how high the reactor power can extend. Although the best-estimate LOCA methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during LOCA, it will take many more resources to develop and to get final approval from the licensing authority. Instead, implementation of evaluation models required by Appendix K of the Code of Federal Regulations, Title 10, Part 50 (10 CFR 50), upon an advanced thermal-hydraulic platform can also gain significant margin on the PCT calculation. A program to modify RELAP5-3D in accordance with Appendix K of 10 CFR 50 was launched by the Institute of Nuclear Energy Research, Taiwan, and it consists of six sequential phases of work. The compliance of the current RELAP5-3D with Appendix K of 10 CFR 50 has been evaluated, and it was found that there are 11 areas where the code modifications are required to satisfy the requirements set forth in Appendix K of 10 CFR 50. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effect experiments and six sets of integral-effect experiments will be adopted. Through the assessments program, all the model changes will be verified

  2. On the Application of Fishbone Diagram Analysis in Tailings Reservoir Dam Break%鱼刺图评价法在尾矿库溃坝事故中的应用

    Institute of Scientific and Technical Information of China (English)

    吴令; 周波

    2012-01-01

    根据尾矿库现场危险、有害因素辨识结果,概括出尾矿库溃坝事故发主的四个主要因素:施工、排洪、渗流和管理,对主要因素的原因进行归纳、分析,找出了重要原因,应用鱼刺图法评价得到了尾矿库溃坝事故的危险程度和等级,提出了防止这类事故发生的重点安全对策措施,提高了尾矿库的安全管理程度.%According to the identification results of dangerous and harmful factors identification in the tailings reservoir site, the four main factors: construction, drainage flood, seepage flow and management are summarized in the dam-break accident. Through induction and analysis, found out the main reasons. The risk degree and level of the tailings dam-break accident are got by using the fishbone diagram, and some safety measures are put forward. Tailing dam safety management level has been improved.

  3. Importance of water quality on plant abundance and diversity in high-alpine meadows of the Yerba Loca Natural Sanctuary at the Andes of north-central Chile Importancia de la calidad del agua sobre la abundancia y diversidad vegetal en vegas altoandinas del Santuario Natural Yerba Loca en los Andes de Chile centro-norte

    Directory of Open Access Journals (Sweden)

    ROSANNA GINOCCHIO

    2008-12-01

    Full Text Available Porphyry Cu-Mo deposits have influenced surface water quality in high-Andes of north-central Chile since the Miocene. Water anomalies may reduce species abundance and diversity in alpine meadows as acidic and metal-rich waters are highly toxic to plants The study assessed the importance of surface water quality on plant abundance and diversity in high-alpine meadows at the Yerba Loca Natural Santuary (YLNS, central Chile (33°15' S, 70°18' W. Hydrochemical and plant prospecting were carried out on Piedra Carvajal, Chorrillos del Plomo and La Lata meadows the growing seasons of 2006 and 2007. Direct gradient analysis was performed through canonical correspondence analysis (CCA to look for relationships among water chemistry and plant factors. High variability in water chemistry was found inside and among meadows, particularly for pH, sulphate, electric conductivity, hardness, and total dissolved Cu, Zn, Cd, Pb and Fe. Data on species abundance and water chemical factors suggests that pH and total dissolved Cu are very important factor determining changes in plant abundance and diversity in study meadows. For instance, Festuca purpurascens, Colobanthus quitensis, and Arenaria rivularis are abundant in habitals with Cu-rich waters while Festuca magellanica, Patosia clandestina, Plantago barbata, Werneria pygmea, and Erigeron andícola are abundant in habitals with dilute waters.Los megadepósitos de pórfidos de Cu-Mo han influido sobre la calidad de las aguas superficiales en las zonas altoandinas del centro-norte de Chile desde el Mioceno. Estas alteraciones en la calidad de las aguas podrían afectar negativamente a la vegetación presente en las vegas altoandinas, ya que las aguas acidas y ricas en metales son altamente tóxicas para las plantas. En este estudio se evaluó el efecto de la calidad de las aguas en la abundancia y diversidad florística de las vegas altoandinas del Santuario de la Naturaleza Yerba Loca (SNYL, en Chile central (33

  4. Leak-before-break analysis of a dissimilar metal welded overlay structure for connecting pipe-nozzle of nuclear reactor pressure vessel to safe end

    International Nuclear Information System (INIS)

    Background: Primary water stress corrosion cracking (PWSCC) is commonly produced in the dissimilar metal welded joints for connecting the pipe-nozzles of nuclear reactor pressure vessels to the safe ends. The technology to repair and mitigate PWSCC is usually to make the weld overlay of higher corrosion resistant Alloy52M on the pipe joints. Purpose: We need to assess the integrity of the welded overlay structures, and to make Leak-before-break (LBB) analyses. The effect of the weld overlay thickness on the LBB behavior needs to be studied. Methods: Based on the three-dimensional finite element fracture mechanics analyses, the ABAQUS software was applied to construct the LBB curves and ligament instability lines of the dissimilar metal welded overlay structures. The effects of the weld overlay thickness on the LBB curves and ligament instability lines were analyzed. Results: The results show that the LBB curves and ligament instability lines with the weld overlay are located above those without the weld overlay. With increasing weld overlay thickness, the LBB curves and ligament instability lines both shift upward. Conclusion: The weld overlay can increase the LBB safe margin of the dissimilar metal welded joints. With increasing weld overlay thickness, the LBB safe margin of the joint structure can be further increased. (authors)

  5. Pre-test analysis for the KNGR LBLOCA DVI performance test using a best estimate code MARS

    International Nuclear Information System (INIS)

    Pre-test analysis using a MARS code has been performed for the KNGR (Korean Next Generation Reactor) DVI (Direct Vessel Injection) performance test facility which is a full height and 1/24.3 volume scaled separate effects test facility focusing on the identification of multi-dimensional thermal-hydraulic phenomena in the downcomer during the reflood conditions of a large break LOCA. From the steady state analyses for various test cases at late reflood condition, the degree of major thermal-hydraulic phenomena such as ECC bypass, ECC penetration, steam condensation, and water level sweep-out are quantified. The MARS code analysis results showed that: (a) multi-dimensional flow and temperature behaviors occurred in the downcomer region as expected, (b) the proximity of ECC injection to the break caused more ECC bypass and less steam condensation efficiency, (c) increasing steam flow rate resulted in more ECC bypass and less steam condensation, (d) and the high velocity of steam flow swept-out the water in the downcomer just below the cold leg nozzle. These results are comparable with those observed in the previous tests such as UPTF and CCTF. (author)

  6. Breaking Littlewood's cipher

    OpenAIRE

    Stehlé, Damien

    2003-01-01

    In 1953, the celebrated mathematician John Edensor Littlewood proposed a stream cipher based on logarithm tables. Fifty years later, we propose the first analysis of his scheme. Littlewood suggests the idea of using real functions as tools to build cryptographic primitives. Even when considering modern security parameters, the original scheme can be broken by a simple attack based on differentiation. We generalise the scheme such that it resists this attack, but describe another attack which ...

  7. More than a break

    DEFF Research Database (Denmark)

    Villadsen, Katrine Weiersoe; Blix, Charlotte; Boisen, Kirsten A

    2015-01-01

    analysis, the following themes emerged: Recreation; Structure, participation, and motivation; and Friends and social network. The social-pedagogical approach is a combination of interpersonal relationships and individually tailored recreational activities. Even small entertaining activities changed...... their motivation to go through their treatment. The interviewees emphasized the importance of experiencing something that was worth telling their friends about to help them stay in touch. CONCLUSION: Although the young patients emphasized the recreational aspects, the time spent with the social educator...... facilitated training in social competencies as well as conversations about emotional and sensitive topics....

  8. A Review of Dangerous Dust in Fusion Reactors: from Its Creation to Its Resuspension in Case of LOCA and LOVA

    Directory of Open Access Journals (Sweden)

    Andrea Malizia

    2016-07-01

    Full Text Available The choice of materials for the future nuclear fusion reactors is a crucial issue. In the fusion reactors, the combination of very high temperatures, high radiation levels, intense production of transmuting elements and high thermomechanical loads requires very high-performance materials. Erosion of PFCs (Plasma Facing Components determines their lifetime and generates a source of impurities (i.e., in-vessel tritium and dust inventories, which cool down and dilute the plasma. The resuspension of dust could be a consequences of LOss of Coolant Accidents (LOCA and LOss of Vacuum Accidents (LOVA and it can be dangerous because of dust radioactivity, toxicity, and capable of causing an explosion. These characteristics can jeopardize the plant safety and pose a serious threat to the operators. The purpose of this work is to determine the experimental and numerical steeps to develop a numerical model to predict the dust resuspension consequences in case of accidents through a comparison between the experimental results taken from campaigns carried out with STARDUST-U and the numerical simulation developed with CFD codes. The authors in this work will analyze the candidate materials for the future nuclear plants and the consequences of the resuspension of its dust in case of accidents through the experience with STARDUST-U.

  9. Symmetry breaking in molecular ferroelectrics.

    Science.gov (United States)

    Shi, Ping-Ping; Tang, Yuan-Yuan; Li, Peng-Fei; Liao, Wei-Qiang; Wang, Zhong-Xia; Ye, Qiong; Xiong, Ren-Gen

    2016-07-11

    Ferroelectrics are inseparable from symmetry breaking. Accompanying the paraelectric-to-ferroelectric phase transition, the paraelectric phase adopting one of the 32 crystallographic point groups is broken into subgroups belonging to one of the 10 ferroelectric point groups, i.e. C1, C2, C1h, C2v, C4, C4v, C3, C3v, C6 and C6v. The symmetry breaking is captured by the order parameter known as spontaneous polarization, whose switching under an external electric field results in a typical ferroelectric hysteresis loop. In addition, the responses of spontaneous polarization to other external excitations are related to a number of physical effects such as second-harmonic generation, piezoelectricity, pyroelectricity and dielectric properties. Based on these, this review summarizes recent developments in molecular ferroelectrics since 2011 and focuses on the relationship between symmetry breaking and ferroelectricity, offering ideas for exploring high-performance molecular ferroelectrics. PMID:27051889

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-15

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

  11. Reliability analysis of the containment spray system of Angra-1 : the injection phase

    International Nuclear Information System (INIS)

    The system studied is projected to perform two basic functions : to reduce the pressure and temperature in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment and to remove the fission products, mainly the iodine of the containment atmosphere. The spray system was analyzed concerning the probability of non-acomplishment of both functions at the same time; therefore the failure of the components of the chemical aditions subsystem are included in the failure tree shown here. (E.G.)

  12. Computer-aided event tree analysis by the impact vector method

    International Nuclear Information System (INIS)

    In the development of the Probabilistic Risk Analysis of Angra I, the ' large event tree/small fault tree' approach was adopted for the analysis of the plant behavior in an emergency situation. In this work, the event tree methodology is presented along with the adaptations which had to be made in order to attain a correct description of the safety system performances according to the selected analysis method. The problems appearing in the application of the methodology and their respective solutions are presented and discussed, with special emphasis to the impact vector technique. A description of the ETAP code ('Event Tree Analysis Program') developed for constructing and quantifying event trees is also given in this work. A preliminary version of the small-break LOCA analysis for Angra 1 is presented as an example of application of the methodology and of the code. It is shown that the use of the ETAP code sigmnificantly contributes to decreasing the time spent in event tree analyses, making it viable the practical application of the analysis approach referred above. (author)

  13. Inflationary implications of supersymmetry breaking

    NARCIS (Netherlands)

    Borghese, Andrea; Roest, Diederik; Zavala, Ivonne

    2013-01-01

    We discuss a general bound on the possibility to realise inflation in any minimal supergravity with F-terms. The derivation crucially depends on the sGoldstini, the scalar field directions that are singled out by spontaneous supersymmetry breaking. The resulting bound involves both slow-roll paramet

  14. Strong coupling electroweak symmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Barklow, T.L. [Stanford Linear Accelerator Center, Menlo Park, CA (United States); Burdman, G. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Physics; Chivukula, R.S. [Boston Univ., MA (United States). Dept. of Physics

    1997-04-01

    The authors review models of electroweak symmetry breaking due to new strong interactions at the TeV energy scale and discuss the prospects for their experimental tests. They emphasize the direct observation of the new interactions through high-energy scattering of vector bosons. They also discuss indirect probes of the new interactions and exotic particles predicted by specific theoretical models.

  15. How do accretion discs break?

    Science.gov (United States)

    Dogan, Suzan

    2016-07-01

    Accretion discs are common in binary systems, and they are often found to be misaligned with respect to the binary orbit. The gravitational torque from a companion induces nodal precession in misaligned disc orbits. In this study, we first calculate whether this precession is strong enough to overcome the internal disc torques communicating angular momentum. We compare the disc precession torque with the disc viscous torque to determine whether the disc should warp or break. For typical parameters precession wins: the disc breaks into distinct planes that precess effectively independently. To check our analytical findings, we perform 3D hydrodynamical numerical simulations using the PHANTOM smoothed particle hydrodynamics code, and confirm that disc breaking is widespread and enhances accretion on to the central object. For some inclinations, the disc goes through strong Kozai cycles. Disc breaking promotes markedly enhanced and variable accretion and potentially produces high-energy particles or radiation through shocks. This would have significant implications for all binary systems: e.g. accretion outbursts in X-ray binaries and fuelling supermassive black hole (SMBH) binaries. The behaviour we have discussed in this work is relevant to a variety of astrophysical systems, for example X-ray binaries, where the disc plane may be tilted by radiation warping, SMBH binaries, where accretion of misaligned gas can create effectively random inclinations and protostellar binaries, where a disc may be misaligned by a variety of effects such as binary capture/exchange, accretion after binary formation.

  16. Sediment transport under breaking waves

    DEFF Research Database (Denmark)

    Christensen, Erik Damgaard; Hjelmager Jensen, Jacob; Mayer, Stefan

    2000-01-01

    generated at the surface where the wave breaks as well as the turbulence generated near the bed due to the wave-motion and the undertow. In general, the levels of turbulent kinetic energy are found to be higher than experiments show. This results in an over prediction of the sediment transport. Nevertheless...

  17. Small Break Air Ingress Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Soo Kim

    2011-09-01

    The small break air-ingress experiment, described in this report, is designed to investigate air-ingress phenomena postulated to occur in pipes in a very high temperature gas-cooled reactor (VHTRs). During this experiment, air-ingress rates were measured for various flow and break conditions through small holes drilled into a pipe of the experimental apparatus. The holes were drilled at right angles to the pipe wall such that a direction vector drawn from the pipe centerline to the center of each hole was at right angles with respect to the pipe centerline. Thus the orientation of each hole was obtained by measuring the included angle between the direction vector of each hole with respect to a reference line anchored on the pipe centerline and pointing in the direction of the gravitational force. Using this reference system, the influence of several important parameters on the air ingress flow rate were measured including break orientation, break size, and flow velocity . The approach used to study the influence of these parameters on air ingress is based on measuring the changes in oxygen concentrations at various locations in the helium flow circulation system as a function of time using oxygen sensors (or detectors) to estimate the air-ingress rates through the holes. The test-section is constructed of a stainless steel pipe which had small holes drilled at the desired locations.

  18. Code breaking in the pacific

    CERN Document Server

    Donovan, Peter

    2014-01-01

    Covers the historical context and the evolution of the technically complex Allied Signals Intelligence (Sigint) activity against Japan from 1920 to 1945 Describes, explains and analyzes the code breaking techniques developed during the war in the Pacific Exposes the blunders (in code construction and use) made by the Japanese Navy that led to significant US Naval victories

  19. Breaking Carbon Lock-in

    DEFF Research Database (Denmark)

    Driscoll, Patrick Arthur

    2014-01-01

    This central focus of this paper is to highlight the ways in which path dependencies and increasing returns (network effects) serve to reinforce carbon lock-in in large-scale road transportation infrastructure projects. Breaking carbon lock-in requires drastic changes in the way we plan future...

  20. 澳洲坚果破壳工艺参数优化及压缩特性的有限元分析%Optimization of technical parameters of breaking Macadamia nut shell and finite element analysis of compression characteristics

    Institute of Scientific and Technical Information of China (English)

    涂灿; 杨薇; 尹青剑; 吕俊龙

    2015-01-01

    sphere. The width diameter was near to average diameter of macadamia nut. The thickness was 4.03-4.36 mm at top of macadamia nut and 2.22-2.48 mm at the middle. That showed that the shell thickness of macadamia nut was nonuniform. The shell’s material was similar to wood, so poisson’s ratio was near to 0.3. Material properties and geometric model of macadamia nut were imported into ANSYS workbench. The distribution of stress and strain of breaking shell of macadamia nut was obtained by finite element analysis of macadamia nut. According to the finite element analysis diagram of macadamia nuts in the 3 loading directions, it could be known that cracks were easy to emerge and expand in horizontal. It was also horizontal which was the most appropriate for breaking macadamia nut shell. The simulation results were consistent with the results of experiment. The biggest shelled force was 2016 N in horizontal, and therefore the shelled force should not be less than 2500 N for ensuring that all nuts would be cracked. The loaded force should be in horizontal when macadamia nut shell was crushed. It provides the design basis for the macadamia nut shell-breaking machine.%为了提高澳洲坚果破壳整仁率,该文首先以加载速度、加载方向、果壳含水率为试验因素进行正交试验,以破壳后澳洲坚果果仁的整仁率为评价指标,优化出适宜澳洲坚果破壳的最佳工艺参数为:加载速率45 mm/min、沿水平向加载、果壳含水率6%~9%。在此破壳工艺下进行试验,得出澳洲坚果的整仁率最高可达93%。澳洲坚果果壳种脐向、宽度向、水平向的平均破壳力分别为1018、2274、1173 N;弹性模量分别为32.24、68.63、39.65 MPa。说明澳洲坚果是各向异性的,宽度向的抗压能力最强。运用有限元方法对澳洲坚果的3个加载方向进行应力和应变分析,得出较佳的施力方向为水平向,与试验结果一致。故在设计破壳机时,施加于

  1. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  2. A model of intrinsic symmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Ge, Li [Research Center for Quantum Manipulation, Department of Physics, Fudan University, Shanghai 200433 (China); Li, Sheng [Department of Physics, Zhejiang Normal University, Zhejiang 310004 (China); George, Thomas F., E-mail: tfgeorge@umsl.edu [Office of the Chancellor and Center for Nanoscience, Department of Chemistry and Biochemistry, University of Missouri-St. Louis, St. Louis, MO 63121 (United States); Department of Physics and Astronomy, University of Missouri-St. Louis, St. Louis, MO 63121 (United States); Sun, Xin, E-mail: xin_sun@fudan.edu.cn [Research Center for Quantum Manipulation, Department of Physics, Fudan University, Shanghai 200433 (China)

    2013-11-01

    Different from the symmetry breaking associated with a phase transition, which occurs when the controlling parameter is manipulated across a critical point, the symmetry breaking presented in this Letter does not need parameter manipulation. Instead, the system itself suddenly undergoes symmetry breaking at a certain time during its evolution, which is intrinsic symmetry breaking. Through a polymer model, it is revealed that the origin of the intrinsic symmetry breaking is nonlinearity, which produces instability at the instance when the evolution crosses an inflexion point, where this instability breaks the original symmetry.

  3. A Reanalysis of Interfacing Systems LOCA Frequency of Hanul Units 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong San; Park, Jin Hee; Jang, Seung Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    An interfacing system loss-of-coolant accident (ISLOCA) is a special kind of loss-of-coolant accident where a low-pressure system interfacing with the reactor coolant system (RCS) is breached due to the overpressurization caused by the failure of isolation between them. Because of its importance in the risk perspective, a probabilistic safety assessment (PSA) typically includes ISLOCA as one of the initiating events for each of which accident scenarios are derived and core damage frequency is calculated. In this paper, the ISLOCA frequency of the Hanul units 3 and 4 was reanalyzed. Considering the state-ofthe- art of ISLOCA analysis, this reanalysis used screening criteria, quantification methods and data that are different from those in the previous analysis. While the two SCS suction lines accounted for about 100% of the total ISLOCA frequency in the previous analysis, the results of this study indicate that CVCS letdown line and CCW return lines from RCP HP coolers are also important when considering the risk of ISLOCA in Hanul units 3 and 4.

  4. Applying process mapping and analysis as a quality improvement strategy to increase the adoption of fruit, vegetable, and water breaks in Australian primary schools.

    Science.gov (United States)

    Biggs, Janice S; Farrell, Louise; Lawrence, Glenda; Johnson, Julie K

    2014-03-01

    Over the past decade, public health policy in Australia has prioritized the prevention and control of obesity and invested in programs that promote healthy eating-related behaviors, which includes increasing fruit and vegetable consumption in children. This article reports on a study that used process mapping and analysis as a quality improvement strategy to improve the delivery of a nutrition primary prevention program delivered in primary schools in New South Wales, Australia. Crunch&Sip® has been delivered since 2008. To date, adoption is low with only 25% of schools implementing the program. We investigated the cause of low adoption and propose actions to increase school participation. We conducted semistructured interviews with key stakeholders and analyzed the process of delivering Crunch&Sip to schools. Interviews and process mapping and analysis identified a number of barriers to schools adopting the program. The analyses identified the need to simplify and streamline the process of delivering the program to schools and introduce monitoring and feedback loops to track ongoing participation. The combination of stakeholder interviews and process mapping and analysis provided important practical solutions to improving program delivery and also contributed to building an understanding of factors that help and hinder program adoption. The insight provided by this analysis helped identify usable routine measures of adoption, which were an improvement over those used in the existing program plan. This study contributed toward improving the quality and efficiency of delivering a health promoting program to work toward achieving healthy eating behaviors in children.

  5. Analysis of ROSA-III test RUN 704 by RELAP5/MOD0 code

    International Nuclear Information System (INIS)

    The ROSA-III test RUN 704 was analyzed for the assessment of RELAP5 code for BWR LOCA. RELAP5 is an advenced code developed to analyze thermal-hydraulic phenomena during LOCA and non-LOCA transients of LWR. It is based on a one-dimensional, nonhomogeneous, nonequilibrium two-phase flow model. The ROSA-III test RUN 704 is a standard BWR LOCA test, simulating a 200% double-ended break at the recirculation pump inlet pipe with all emergency core cooling systems activated. Large increase of core inlet flow due to lower plenum flashing and resulted rewetting of heater surface were calculated by RELAP5, indicating superior capability of RELAP5 two-phase flow model than RELAP4 phase separation model. Vapor and liquid counter-current flow was calculated at core inlet and core outlet. A small degree thermal nonequilibrium between vapor and liquid was calculated in the upper plenum after HPCS activation. However, core reflooding and quenching of heater surface were not calculated. There are still room for improvement in the interfacial drag and the heat transfer models of RELAP5/MOD0. (author)

  6. Symmetries, Symmetry Breaking, Gauge Symmetries

    CERN Document Server

    Strocchi, Franco

    2015-01-01

    The concepts of symmetry, symmetry breaking and gauge symmetries are discussed, their operational meaning being displayed by the observables {\\em and} the (physical) states. For infinitely extended systems the states fall into physically disjoint {\\em phases} characterized by their behavior at infinity or boundary conditions, encoded in the ground state, which provide the cause of symmetry breaking without contradicting Curie Principle. Global gauge symmetries, not seen by the observables, are nevertheless displayed by detectable properties of the states (superselected quantum numbers and parastatistics). Local gauge symmetries are not seen also by the physical states; they appear only in non-positive representations of field algebras. Their role at the Lagrangian level is merely to ensure the validity on the physical states of local Gauss laws, obeyed by the currents which generate the corresponding global gauge symmetries; they are responsible for most distinctive physical properties of gauge quantum field ...

  7. Renormalizable theories with symmetry breaking

    CERN Document Server

    Becchi, Carlo M

    2016-01-01

    The description of symmetry breaking proposed by K. Symanzik within the framework of renormalizable theories is generalized from the geometrical point of view. For an arbitrary compact Lie group, a soft breaking of arbitrary covariance, and an arbitrary field multiplet, the expected integrated Ward identities are shown to hold to all orders of renormalized perturbation theory provided the Lagrangian is suitably chosen. The corresponding local Ward identity which provides the Lagrangian version of current algebra through the coupling to an external, classical, Yang-Mills field, is then proved to hold up to the classical Adler-Bardeen anomaly whose general form is written down. The BPHZ renormalization scheme is used throughout in such a way that the algebraic structure analyzed in the present context may serve as an introduction to the study of fully quantized gauge theories.

  8. Breaking GSM with rainbow Tables

    CERN Document Server

    Meyer, Steven

    2011-01-01

    Since 1998 the GSM security has been academically broken but no real attack has ever been done until in 2008 when two engineers of Pico Computing (FPGA manufacture) revealed that they could break the GSM encryption in 30 seconds with 200'000$ hardware and precomputed rainbow tables. Since then the hardware was either available for rich people only or was confiscated by government agencies. So Chris Paget and Karsten Nohl decided to react and do the same thing but in a distributed open source form (on torrent). This way everybody could "enjoy" breaking GSM security and operators will be forced to upgrade the GSM protocol that is being used by more than 4 billion users and that is more than 20 years old.

  9. Breaking through the tranfer tunnel

    CERN Multimedia

    Laurent Guiraud

    2001-01-01

    This image shows the tunnel boring machine breaking through the transfer tunnel into the LHC tunnel. Proton beams will be transferred from the SPS pre-accelerator to the LHC at 450 GeV through two specially constructed transfer tunnels. From left to right: LHC Project Director, Lyn Evans; CERN Director-General (at the time), Luciano Maiani, and Director for Accelerators, Kurt Hubner.

  10. Lugar de encuentros de tópicos románicos : Doña Juana la Loca de Pradilla

    OpenAIRE

    García Melero, José Enrique

    1999-01-01

    Estudio del cuadro titulado «Doña Juana la Loca» de Pradilla, que aquí es considerado como la culminación de la pintura de historia en España y el inicio de su decadencia. Se atribuye su éxito en su época a la concurrencia en él de una serie de características románticas tanto formales como de contenido: la representación de la locura de una Reina enamorada y celosa debido a la muerte de su esposo el Rey Felipe I, tema sacado de la Historia al poco tiemp...

  11. 16th Edition Breaks New Ground

    Institute of Scientific and Technical Information of China (English)

    Zhao Fei

    2010-01-01

    @@ Breaking new ground,Intertextile Shanghai Apparel Fabrics,successfully closed its doors on October 22nd,2010at the Shanghai New International Exhibition Centre,attracting a record breaking,more than 57,000 buyers.

  12. Post-test analysis of ROSA-III test Run 701

    International Nuclear Information System (INIS)

    Post-test analysis of ROSA-III test Run 701, the first simulated BWR/LOCA experiment in the ROSA-III program, was made by a computer code, RELAP4J. Objectives of the analysis were to evaluate ability of the code to predict thermal hydraulic behavior of coolant by comparing the analytical results with the test data, to show areas of model refinement and to reveal requirements in the program. Analysis showed the following: 1) Calculation model to give a maximum flow rate at low quality is important in determining coolant behavior in the system. 2) The phase separation model in the lower plenum influences calculation of the discharge quality at the break and hence that of coolant behavior in the system. 3) Overall behavior of coolant can be well predicted. 4) Effect of ECC water cannot be well predicted by RELAP4J. Following are the requirements in future tests: 1) measurement of the initial pressure distribution in the system, 2) improvement or development of two-phase flow instrumentations, especially in flow rate measurement, 3) simplification of the feed water and steam discharge lines. In future analysis, emphasis should be placed on prediction of ECC water behavior, evaluation of heat transfer correlation and realistic evaluation of liquid level in the core. (author)

  13. Correlation analysis of symmetry breaking in the surface nanostructure ordering: case study of the ventral scale of the snake Morelia viridis

    Science.gov (United States)

    Kovalev, A.; Filippov, A.; Gorb, S. N.

    2016-03-01

    In contrast to the majority of inorganic or artificial materials, there is no ideal long-range ordering of structures on the surface in biological systems. Local symmetry of the ordering on biological surfaces is also often broken. In the present paper, the particular symmetry violation was analyzed for dimple-like nano-pattern on the belly scales of the skin of the pythonid snake Morelia viridis using correlation analysis and statistics of the distances between individual nanostructures. The results of the analysis performed on M. viridis were compared with a well-studied nano-nipple pattern on the eye of the sphingid moth Manduca sexta, used as a reference. The analysis revealed non-random, but very specific symmetry violation. In the case of the moth eye, the nano-nipple arrangement forms a set of domains, while in the case of the snake skin, the nano-dimples arrangement resembles an ordering of particles (molecules) in amorphous (glass) state. The function of the nano-dimples arrangement may be to provide both friction and strength isotropy of the skin. A simple model is suggested, which provides the results almost perfectly coinciding with the experimental ones. Possible mechanisms of the appearance of the above nano-formations are discussed.

  14. Spontaneous Breaking of the Quantum Superposition

    OpenAIRE

    Pankovic, Vladan; Predojevic, Milan

    2007-01-01

    In this work spontaneous (non-dynamical) breaking (effective hiding) of the unitary quantum mechanical dynamical symmetry (superposition) is considered. It represents an especial but very interesting case of the general formalism of the spontaneous symmetry breaking (effective hiding). Conceptual analogies with spontaneous breaking of the gauge symmetry in Weinberg-Sallam's electro-weak interaction are pointed out. Also, consequences of the spontaneous superposition breaking in the measuremen...

  15. Identification of limiting case between DBA and SBDBA (CL break area sensitivity): A new model for the boron injection system

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez Gonzalez, R., E-mail: r.gonzalez@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); D’Auria, F., E-mail: f.dauria@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); Mazzantini, O., E-mail: mazzantini@na-sa.com.ar [Nucleo-electrica Argentina Sociedad Anonima (NA-SA), Buenos Aires (Argentina)

    2014-08-15

    Atucha-2 is a Siemens-designed Pressurized Heavy Water Reactor (PHWR) reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarity (e.g. oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank. The boron clouds reconstruction is obtained using a Computational Fluid Dynamics (CFD) CFX code calculation. A complete Large Break Loss Of Coolant Accident (LBLOCA) calculation implies the application of the RELAP5-3D{sup ©} system code. Within the framework of the third Agreement “Nucleoelèctrica Argentina-Sociedad Anonima (NA-SA) – University of Pisa/GRNSPG” (Contract, 2009), a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions.

  16. BREAKING WAVE FORCES ON VERTICAL CYLINDERS

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The breaking wave forces on vertical cylinders in shallow waters were studied by means of experimental methods. The results indicate that the breaking wave pressure is distributed exponentially with respect to water depth. An experimental formula was given using the test data. Compared with test data, the calculated breaking wave forces are in good agreement with the test data.

  17. Electroweak symmetry breaking at photon colliders

    International Nuclear Information System (INIS)

    The electroweak-symmetry-breaking sector of the standard model can be weakly-coupled or can be strongly-coupled, which is characterized by some kinds of strong interaction among the Goldstone bosons of the electroweak-symmetry-breaking sector. In this paper, we summarize an investigation of probing the strong electroweak-symmetry-breaking effects at photon colliders. ((orig.))

  18. Thermal-hydraulic evaluation of passive containment cooling system of improved APR+ during LOCAs

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Byong Guk; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2014-10-15

    Highlights: • Larger air holdup tanks are confirmed to be needed for PCCS through MARS simulation. • Condensation model in MARS gives lower heat transfer than Dehbi's correlation. • Heat transfer resistance of steam–air mixture is dominant for PCCS tubes. • Heat sink structures remove dominant amount of decay heat initially. • Flow instability is eased at higher tube inlet form loss, return level, and angle. - Abstract: Various reactor concepts and technologies have been devised and evaluated to ensure the integrity of the core and the containment under a prolonged station blackout. After the successful validation of the passive auxiliary feedwater system (PAFS) of Advanced Power Reactor Plus (APR+), Korea is considering an improved APR+ with a passive containment cooling system (PCCS). In a previous paper, we suggested a PCCS design based on APR+ and performed a scoping analysis. We performed a MARS simulation for the thermal hydraulic evaluation of the system behavior, including natural circulation through the Passive Condensation Cooling Tank (PCCT) water pool and inside PCCS tubes as well as steam–air mixture condensation inside the containment. Through a simulation using the MARS system code, we investigated the effect of air holdup tanks (AHTs) on reduction of the air fraction in the containment, the effects of several steam–air mixture condensation models, the role of heat structures, and the flow instability inside the PCCS tubes. We found that the presence of AHT reduced the number of required PCCS tubes by more than half and the heat resistance of the steam–air mixture side is dominant in terms of governing the overall performance of PCCS. The embedded MARS condensation model and Uchida's correlation gave lower heat transfer coefficients than Dehbi's correlation, and heat structures removed more decay heat than PCCS tubes. Finally, intense flow instability inside the PCCS tubes was observed and was mitigated by placing

  19. Surface tension effects in wave breaking

    Science.gov (United States)

    Deike, Luc; Melville, W. K.; Popinet, Stephane

    2014-11-01

    We present a numerical study of wave breaking by solving the full Navier-Stokes equations for two-phase air-water flows using the solver Gerris. We describe a parametric study of the influence of capillary effects on wave breaking using two-dimensional simulations. The onset of wave breaking as a function of the Bond number, Bo, and the initial wave steepness S is determined and a phase diagram in terms of (S,Bo) is presented that distinguishes between non-breaking gravity waves, parasitic capillaries on a gravity wave, spilling breakers and plunging breakers. The wave energy dissipation is computed for each wave regime and is found to be in good agreement with experimental results for breaking waves. Moreover, the enhanced dissipation just by parasitic capillaries is comparable to the dissipation due to breaking. Extending the simulations to three dimensions permits studies of the generation and statistics of bubbles and spray during breaking.

  20. Directional excitation without breaking reciprocity

    Science.gov (United States)

    Ramezani, Hamidreza; Dubois, Marc; Wang, Yuan; Shen, Y. Ron; Zhang, Xiang

    2016-09-01

    We propose a mechanism for directional excitation without breaking reciprocity. This is achieved by embedding an impedance matched parity-time symmetric potential in a three-port system. The amplitude distribution within the gain and loss regions is strongly influenced by the direction of the incoming field. Consequently, the excitation of the third port is contingent on the direction of incidence while transmission in the main channel is immune. Our design improves the four-port directional coupler scheme, as there is no need to implement an anechoic termination to one of the ports.

  1. History of electroweak symmetry breaking

    CERN Document Server

    Kibble, T W B

    2015-01-01

    In this talk, I recall the history of the development of the unified electroweak theory, incorporating the symmetry-breaking Higgs mechanism, as I saw it from my standpoint as a member of Abdus Salam's group at Imperial College. I start by describing the state of physics in the years after the Second World War, explain how the goal of a unified gauge theory of weak and electromagnetic interactions emerged, the obstacles encountered, in particular the Goldstone theorem, and how they were overcome, followed by a brief account of more recent history, culminating in the historic discovery of the Higgs boson in 2012.

  2. Leaders break ground for INFINITY

    Science.gov (United States)

    2008-01-01

    Community leaders from Mississippi and Louisiana break ground for the new INFINITY at NASA Stennis Space Center facility during a Nov. 20 ceremony. Groundbreaking participants included (l to r): Gottfried Construction representative John Smith, Mississippi Highway Commissioner Wayne Brown, INFINITY board member and Apollo 13 astronaut Fred Haise, Stennis Director Gene Goldman, Studio South representative David Hardy, Leo Seal Jr. family representative Virginia Wagner, Hancock Bank President George Schloegel, Mississippi Rep. J.P. Compretta, Mississippi Band of Choctaw Indians representative Charlie Benn and Louisiana Sen. A.G. Crowe.

  3. Spontaneous Breaking of Flavor Symmetry

    CERN Document Server

    Törnqvist, N A

    1996-01-01

    It is shown that part of the quark masses of the standard model can be generated spontaneously within the strong interactions of QCD. After the breaking of U(Nf) x U(Nf) symmetry by the vacuum, also the resulting flavor symmetric, degenerate meson mass spectrum is shown to be unstable with respect to quantum loops, for rather general models. For a C-degenerate meson spectrum the stable mass spectrum obeys the Okubo-Zweig-Iizuka rule and the approximateequal spacing rule.

  4. Bone-breaking bite force of Basilosaurus isis (Mammalia, Cetacea) from the late Eocene of Egypt estimated by finite element analysis.

    Science.gov (United States)

    Snively, Eric; Fahlke, Julia M; Welsh, Robert C

    2015-01-01

    Bite marks suggest that the late Eocence archaeocete whale Basilosaurus isis (Birket Qarun Formation, Egypt) fed upon juveniles of the contemporary basilosaurid Dorudon atrox. Finite element analysis (FEA) of a nearly complete adult cranium of B. isis enables estimates of its bite force and tests the animal's capabilities for crushing bone. Two loadcases reflect different biting scenarios: 1) an intitial closing phase, with all adductors active and a full condylar reaction force; and 2) a shearing phase, with the posterior temporalis active and minimized condylar force. The latter is considered probable when the jaws were nearly closed because the preserved jaws do not articulate as the molariform teeth come into occulusion. Reaction forces with all muscles active indicate that B. isis maintained relatively greater bite force anteriorly than seen in large crocodilians, and exerted a maximum bite force of at least 16,400 N at its upper P3. Under the shearing scenario with minimized condylar forces, tooth reaction forces could exceed 20,000 N despite lower magnitudes of muscle force. These bite forces at the teeth are consistent with bone indentations on Dorudon crania, reatract-and-shear hypotheses of Basilosaurus bite function, and seizure of prey by anterior teeth as proposed for other archaeocetes. The whale's bite forces match those estimated for pliosaurus when skull lengths are equalized, suggesting similar tradeoffs of bite function and hydrodynamics. Reaction forces in B. isis were lower than maxima estimated for large crocodylians and carnivorous dinosaurs. However, comparison of force estimates from FEA and regression data indicate that B. isis exerted the largest bite forces yet estimated for any mammal, and greater force than expected from its skull width. Cephalic feeding biomechanics of Basilosaurus isis are thus consistent with habitual predation. PMID:25714832

  5. Bone-breaking bite force of Basilosaurus isis (Mammalia, Cetacea from the late Eocene of Egypt estimated by finite element analysis.

    Directory of Open Access Journals (Sweden)

    Eric Snively

    Full Text Available Bite marks suggest that the late Eocence archaeocete whale Basilosaurus isis (Birket Qarun Formation, Egypt fed upon juveniles of the contemporary basilosaurid Dorudon atrox. Finite element analysis (FEA of a nearly complete adult cranium of B. isis enables estimates of its bite force and tests the animal's capabilities for crushing bone. Two loadcases reflect different biting scenarios: 1 an intitial closing phase, with all adductors active and a full condylar reaction force; and 2 a shearing phase, with the posterior temporalis active and minimized condylar force. The latter is considered probable when the jaws were nearly closed because the preserved jaws do not articulate as the molariform teeth come into occulusion. Reaction forces with all muscles active indicate that B. isis maintained relatively greater bite force anteriorly than seen in large crocodilians, and exerted a maximum bite force of at least 16,400 N at its upper P3. Under the shearing scenario with minimized condylar forces, tooth reaction forces could exceed 20,000 N despite lower magnitudes of muscle force. These bite forces at the teeth are consistent with bone indentations on Dorudon crania, reatract-and-shear hypotheses of Basilosaurus bite function, and seizure of prey by anterior teeth as proposed for other archaeocetes. The whale's bite forces match those estimated for pliosaurus when skull lengths are equalized, suggesting similar tradeoffs of bite function and hydrodynamics. Reaction forces in B. isis were lower than maxima estimated for large crocodylians and carnivorous dinosaurs. However, comparison of force estimates from FEA and regression data indicate that B. isis exerted the largest bite forces yet estimated for any mammal, and greater force than expected from its skull width. Cephalic feeding biomechanics of Basilosaurus isis are thus consistent with habitual predation.

  6. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  7. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  8. Best estimate plus uncertainty analysis of LBLOCA for Indian PHWR

    International Nuclear Information System (INIS)

    Deterministic safety analysis is an important tool for confirming the adequacy of provisions within the defense-in-depth concept for the safety of nuclear power plants. One of the important design basis events is considered to be a complete double-ended guillotine rupture i.e. Loss of Coolant Accident (LOCA) of largest and coldest pipe in the primary coolant circuit. The present work deals with this scenario for an Indian PHWR. It highlights the identification of critical break size leading to the maximum clad temperature using the best estimate code RELAP5. Further, important parameters affecting the clad temperature are described along with results of initial sensitivity studies to select dominant uncertain parameters. For the uncertainty propagation, Latin Hypercube Sampling (LHS) is used instead of simple random sampling for Monte-Carlo simulation. The inherent characteristic of LHS is to reduce the required runs for Monte-Carlo simulation to manageable order for the current computing capability. The 95th percentile value of peak cladding temperature (PCT) is obtained by the method and compared with acceptance criteria. (orig.)

  9. DNA Double-Strand Break Analysis by {gamma}-H2AX Foci: A Useful Method for Determining the Overreactors to Radiation-Induced Acute Reactions Among Head-and-Neck Cancer Patients

    Energy Technology Data Exchange (ETDEWEB)

    Goutham, Hassan Venkatesh; Mumbrekar, Kamalesh Dattaram [Division of Radiobiology and Toxicology, Manipal Life Sciences Centre, Manipal University, Manipal, Karnataka (India); Vadhiraja, Bejadi Manjunath [Manipal Hospital, Bangalore, Karnataka (India); Fernandes, Donald Jerard; Sharan, Krishna [Department of Radiotherapy and Oncology, Shiridi Sai Baba Cancer Hospital and Research Centre, Kasturba Hospital, Manipal, Karnataka (India); Kanive Parashiva, Guruprasad; Kapaettu, Satyamoorthy [Division of Biotechnology, Manipal Life Sciences Centre, Manipal University, Manipal, Karnataka (India); Bola Sadashiva, Satish Rao, E-mail: satishraomlsc@gmail.com [Division of Radiobiology and Toxicology, Manipal Life Sciences Centre, Manipal University, Manipal, Karnataka (India)

    2012-12-01

    Purpose: Interindividual variability in normal tissue toxicity during radiation therapy is a limiting factor for successful treatment. Predicting the risk of developing acute reactions before initiation of radiation therapy may have the benefit of opting for altered radiation therapy regimens to achieve minimal adverse effects with improved tumor cure. Methods and Materials: DNA double-strand break (DSB) induction and its repair kinetics in lymphocytes of head-and-neck cancer patients undergoing chemoradiation therapy was analyzed by counting {gamma}-H2AX foci, neutral comet assay, and a modified version of neutral filter elution assay. Acute normal tissue reactions were assessed by Radiation Therapy Oncology Group criteria. Results: The correlation between residual DSBs and the severity of acute reactions demonstrated that residual {gamma}-H2AX foci in head-and-neck cancer patients increased with the severity of oral mucositis and skin reaction. Conclusions: Our results suggest that {gamma}-H2AX analysis may have predictive implications for identifying the overreactors to mucositis and skin reactions among head-and-neck cancer patients prior to initiation of radiation therapy.

  10. 某超高射速舰炮弹药筒膛内横断问题技术分析%Technical Analysis of Cartridge Breaking in Bore for Super-high Firing Rate Naval Gun

    Institute of Scientific and Technical Information of China (English)

    马献怀

    2015-01-01

    针对某超高射速舰炮及其弹道炮试验过程中出现的药筒膛内横断问题,通过现场勘测、理论分析和故障还原试验,确定了故障原因。根据故障机理,经技术攻关,提出了解决措施,经过后续多次射击试验验证,故障再未复现,证明了解决措施的可行有效。分析结果对解决速射火炮类似故障有重要的参考价值。%Aimed at the cartridge breaking problem occurred during the test of super-high firing rate naval gun and ballistic gun,the failure causes were determined by site investigation,theo-retical analysis and failure reproduction test.The solution was proposed based on the failure mechanism and technology research.The same failure hasn't occurred again during many firing test.The results show the solution is reasonable and feasible,which provides an important ref-erence value for solving similar failure of rapid-fire gun.

  11. Technical Analysis of Cartridge Breaking in Bore for Super-high Firing Rate Naval Gun%某超高射速舰炮弹药筒膛内横断问题技术分析

    Institute of Scientific and Technical Information of China (English)

    马献怀

    2015-01-01

    针对某超高射速舰炮及其弹道炮试验过程中出现的药筒膛内横断问题,通过现场勘测、理论分析和故障还原试验,确定了故障原因。根据故障机理,经技术攻关,提出了解决措施,经过后续多次射击试验验证,故障再未复现,证明了解决措施的可行有效。分析结果对解决速射火炮类似故障有重要的参考价值。%Aimed at the cartridge breaking problem occurred during the test of super-high firing rate naval gun and ballistic gun,the failure causes were determined by site investigation,theo-retical analysis and failure reproduction test.The solution was proposed based on the failure mechanism and technology research.The same failure hasn't occurred again during many firing test.The results show the solution is reasonable and feasible,which provides an important ref-erence value for solving similar failure of rapid-fire gun.

  12. 新生儿惊厥148例临床分析%Neonatal Shock Clinical Analysis of 148 Cases of Break Off

    Institute of Scientific and Technical Information of China (English)

    劳庆禄

    2012-01-01

      目的探讨新生儿惊厥病因、临床特点及急救方法,减少惊厥及后遗症的发生.方法回顾性分析我院2005年01月-2011年12月共收治的148例惊厥新生儿的临床资料.结果痊愈121例,有症状出院的19例,死亡5例,自动出院3例,20例遗留有不同程度神经系统后遗症.围产期窒息是引起惊厥的首要原因,其次是感染性疾病、糖及电解质紊乱、各类先天性遗传代谢病.发作类型以微小性发作为多见,其次为阵挛性发作.治疗重点在于止痉,首选苯巴比妥,14.9%的患儿足量用药不能控制.结论围产期窒息是导致新生儿惊厥的主要病因;苯巴比妥仍是抗惊厥的首选药物,少数患儿须加用二线药物;发作形式以微小性发作最为多见;EEG 背景波是判断预后的一个良好的指示.%  Objective Investigate the cause of neonatal convulsions、clinical characteristics and first-aid methods,reduce the incidence of seizures and the sequelae.Methods Retrospective analysis of our hospital from Janu-ary 2005 to December 2011 in the total of 148 cases of eclampsia admitted to the clinical data of newborn.Results 121 cases of ful recovery,discharged from hospital with symptoms of 19 cases,the death of five cases, three cases of automatic discharge, there are 20 cases left over from the nervous system sequelae of varying degrees.Caused by perinatal asphyxia is the primary cause of febrile seizures, folowed by infectious diseases, sugar and electrolyte disorders, various types of congenital genetic metabolic diseases. Type of attack made little as often, folowed by clonic seizures. To stop focusing on the reatment of seizures, Phenobarbital prefered ,14.9% of children with a sufficient quantiy of medicines stil con not control.Conclusions Perinatal asphyxia is a major cause of neonatal convulsions;Phenobarbital is stil the first choice anticonvulsant drug, a smal number of children with second-line drugs should be added; The

  13. A Study of the Rock Breaking Mechanism during Swirling Water Jet Drilling

    Institute of Scientific and Technical Information of China (English)

    NiHongjian; WangRuihe

    2004-01-01

    Based on an analysis of the factors affecting rock breaking and the coupling between rock and fluid during water jet drilling, the rock damage model and the damage-coupling model suitable for the whole rock breaking process under the water jet is established with continuous damage mechanics and micro-damage mechanics. The evolvement of rock damage during swirling water jet drilling is simulated on a nonlinear FEM and dynamic rock damage model, and a decoupled method is used to analyze the rock damage. The numerical results agree with the test results to a high degree, which shows the rock breaking ability of the swirling water jet is strong. This is because the jet particle velocity of the swirling water jet is three-dimensional, and its rock-breaking manner mainly has a slopping impact. Thus, the interference from returning fluid is less. All these aspects make it easy to draw and shear the rock surface. The rock breaking process is to break out an annular on the rock surface first, and then the annular develops quickly in both the radial and axial directions, the last part of the rock broken hole bottom is a protruding awl. The advantage of the swirling water jet breaking rock is the heavy breaking efficiency, large breaking area and less energy used to break rock per unite volume, so the swirling water jet can drill in a hole of a large diameter.

  14. Boundary breaking for interdisciplinary learning

    Directory of Open Access Journals (Sweden)

    Adi Kidron

    2015-10-01

    Full Text Available The purpose of this work is to contribute to the body of knowledge on processesby which students develop interdisciplinary understanding of contents, as well as to suggest technology-enhanced means for supporting them in these processes in the context of higher education. In doing so, we suggest a rethinking of three traditional practices that tend to characterise typical higher education instruction: (1 compartmentalisation of disciplines; (2 traditional pedagogy; and (3 traditional hierarchies based on levels of expertise. Our high-level conjecture was that meaningful dialogue with peers and experts supports both the deepening of ideas in one knowledge domain and the formation of connections between ideas from several domains, both of which are required for the development of interdisciplinary understanding. We developed the Boundary Breaking for Interdisciplinary Learning (BBIL model, which harnesses technology to break boundaries between disciplines, learners and organisational levels of hierarchy. Findings indicate that 36 undergraduate students who participated in an interdisciplinary online course that implemented the BBIL model have significantly improved their interdisciplinary understanding of the course contents. This study illustrates how innovative use of available, free and low-cost technology can produce a ‘positive disruption’ in higher education instruction.

  15. How often precipitation records break?

    Science.gov (United States)

    Papalexiou, Simon Michael; Oikonomou, Maria; Floutsakou, Athina; Bessas, Nikolaos; Mamassis, Nikos

    2016-04-01

    How often precipitation records break? Are there any factors that determine the average time needed for the next maximum to occur? In order to investigate these simple questions we use several hundreds of daily precipitation records (more than 100 years long each) and we study the time intervals between each successive maximum precipitation value. We investigate if the record breaking time interval is related (a) to the autocorrelation structure, (b) to probability dry, and (c) to the tail of the marginal distribution. For the last, we first, evaluate which type of tail is better fitted by choosing among three general types of tails corresponding to the distributions Pareto, Lognormal and Weibull; and second, we assess the heaviness of the tail based on the estimated shape parameter. The performance of each tail is evaluated in terms of return period values, i.e., we compare the empirical return periods of precipitation values above a threshold with the predicted ones by each of the three types of fitted tails.

  16. Improved single sector supersymmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Luty, Markus A.; Terning, John

    1998-12-09

    Building on recent work by N. Arkani-Hamed and the present authors, we construct realistic models that break supersymmetry dynamically and give rise to composite quarks and leptons, all in a single strongly-coupled sector. The most important improvement compared to earlier models is that the second-generation composite states correspond to dimension-2 ''meson'' operators in the ultraviolet. This leads to a higher scale for flavor physics, and gives a completely natural suppression of flavor-changing neutral currents. We also construct models in which the hierarchy of Yukawa couplings is explained by the dimensionality of composite states. These models provide an interesting and viable alternative to gravity- and gauge-mediated models. The generic signatures are unification of scalar masses with different quantum numbers at the compositeness scale, and lighter gaugino, Higgsino, and third-generation squark and slepton masses. We also analyze large classes of models that give rise to both compositeness and supersymmetry breaking, based on gauge theories with confining, fixed-point, or free-magnetic dynamics.

  17. Strictly Anomaly Mediated Supersymmetry Breaking

    CERN Document Server

    Hindmarsh, Mark

    2012-01-01

    We consider an MSSM extension with anomaly mediation as the source of supersymmetry-breaking, and a U(1) symmetry which solves the tachyonic slepton problem, and introduces both the see-saw mechanism for neutrino masses, and the Higgs mu-term. We compare its spectra with those from so-called minimal anomaly mediated supersymmetry breaking. We find a Standard Model-like Higgs of mass 124 GeV with a gravitino mass of 120 TeV and tan(beta)=17, while a contribution to the muon anomalous magnetic moment within 2 sigma of the discrepancy between Standard Model theory and experiment favours a slightly lower gravitino mass of around 80 TeV. The model naturally produces a period of hybrid inflation, with exit to a false vacuum characterised by large Higgs vevs, the true ground state being achieved after a period of thermal inflation. The scalar spectral index is reduced to approximately 0.975, and the correct abundance of dark matter can be produced by decays of thermally-produced gravitinos, provided the gravitino ma...

  18. Introduction to Electroweak Symmetry Breaking

    Energy Technology Data Exchange (ETDEWEB)

    Dawson,S.

    2008-10-02

    The Standard Model (SM) is the backbone of elementary particle physics-not only does it provide a consistent framework for studying the interactions of quark and leptons, but it also gives predictions which have been extensively tested experimentally. In these notes, I review the electroweak sector of the Standard Model, discuss the calculation of electroweak radiative corrections to observables, and summarize the status of SM Higgs boson searches. Despite the impressive experimental successes, however, the electroweak theory is not completely satisfactory and the mechanism of electroweak symmetry breaking is untested. I will discuss the logic behind the oft-repeated statement: 'There must be new physics at the TeV scale'. These lectures reflect my strongly held belief that upcoming results from the LHC will fundamentally change our understanding of electroweak symmetry breaking. In these lectures, I review the status of the electroweak sector of the Standard Model, with an emphasis on the importance of radiative corrections and searches for the Standard Model Higgs boson. A discussion of the special role of the TeV energy scale in electroweak physics is included.

  19. Chiral symmetry breaking and monopoles

    CERN Document Server

    Di Giacomo, Adriano; Pucci, Fabrizio

    2015-01-01

    To understand the relation between the chiral symmetry breaking and monopoles, the chiral condensate which is the order parameter of the chiral symmetry breaking is calculated in the $\\overline{\\mbox{MS}}$ scheme at 2 [GeV]. First, we add one pair of monopoles, varying the monopole charges $m_{c}$ from zero to four, to SU(3) quenched configurations by a monopole creation operator. The low-lying eigenvalues of the Overlap Dirac operator are computed from the gauge links of the normal configurations and the configurations with additional monopoles. Next, we compare the distributions of the nearest-neighbor spacing of the low-lying eigenvalues with the prediction of the random matrix theory. The low-lying eigenvalues not depending on the scale parameter $\\Sigma$ are compared to the prediction of the random matrix theory. The results show the consistency with the random matrix theory. Thus, the additional monopoles do not affect the low-lying eigenvalues. Moreover, we discover that the additional monopoles increa...

  20. Chiral symmetry breaking with the Curtis-Pennington vertex

    NARCIS (Netherlands)

    Atkinson, D.; Gusynin, V. P.; Maris, P.

    1992-01-01

    Published in: Phys. Lett. B 303 (1993) 157-162 citations recorded in [Science Citation Index] Abstract: We study chiral symmetry breaking in quenched QED$_4$, using a vertex Ansatz recently proposed by Curtis and Pennington. Bifurcation analysis is employed to establish the existence of a critical c

  1. The Analysis of the Patterns of Radiation-Induced DNA Damage Foci by a Stochastic Monte Carlo Model of DNA Double Strand Breaks Induction by Heavy Ions and Image Segmentation Software

    Science.gov (United States)

    Ponomarev, Artem; Cucinotta, F.

    2011-01-01

    To create a generalized mechanistic model of DNA damage in human cells that will generate analytical and image data corresponding to experimentally observed DNA damage foci and will help to improve the experimental foci yields by simulating spatial foci patterns and resolving problems with quantitative image analysis. Material and Methods: The analysis of patterns of RIFs (radiation-induced foci) produced by low- and high-LET (linear energy transfer) radiation was conducted by using a Monte Carlo model that combines the heavy ion track structure with characteristics of the human genome on the level of chromosomes. The foci patterns were also simulated in the maximum projection plane for flat nuclei. Some data analysis was done with the help of image segmentation software that identifies individual classes of RIFs and colocolized RIFs, which is of importance to some experimental assays that assign DNA damage a dual phosphorescent signal. Results: The model predicts the spatial and genomic distributions of DNA DSBs (double strand breaks) and associated RIFs in a human cell nucleus for a particular dose of either low- or high-LET radiation. We used the model to do analyses for different irradiation scenarios. In the beam-parallel-to-the-disk-of-a-flattened-nucleus scenario we found that the foci appeared to be merged due to their high density, while, in the perpendicular-beam scenario, the foci appeared as one bright spot per hit. The statistics and spatial distribution of regions of densely arranged foci, termed DNA foci chains, were predicted numerically using this model. Another analysis was done to evaluate the number of ion hits per nucleus, which were visible from streaks of closely located foci. In another analysis, our image segmentaiton software determined foci yields directly from images with single-class or colocolized foci. Conclusions: We showed that DSB clustering needs to be taken into account to determine the true DNA damage foci yield, which helps to

  2. Experimental Study of Wave Breaking on Gentle Slope

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    -An experimental study of regular wave and irregular wave breaking is performed on a gentle slope of 1:200. In the experiment, asymmetry of wave profile is analyzed to determine its effect on wave breaker indices and to explain the difference between Goda and Nelson about the breaker indices of regular waves on very mild slopes. The study shows that the breaker index of irregular waves is under less influence of bottom slope i, relative water depth d/ L0 and the asymmetry of wave profile than that of regular waves. The breaker index of regular waves from Goda may be used in the case of irregular waves, while the coefficient A should be 0.15. The ratio of irregular wavelength to the length calculated by linear wave theory is 0.74. Analysis is also made on the waveheight damping coefficient of regular waves after breaking and on the breaking probability of large irregular waves.

  3. NUMERICAL SIMULATION OF ROCK BREAKING MECHANISM WITH HIGH-PRESSURE WATER JET

    Institute of Scientific and Technical Information of China (English)

    NI Hong-jian

    2004-01-01

    Based on the analysis of experimental results, the rock damage model and the damage coupling model suitable for the whole rock breaking process with water jet were established with continuous damage mechanics and micro damage mechanics, and the numerical method was developed with continuum mechanics and the FEM theory. The rock breaking mechanism with water jet was studied systematically with numerical simulation for the first time in the field of water-jet rock breaking. The numerical results agree with the experimental ones which shows that the presented method is reasonable and can reflect the reality of water-jet rock breaking. The conclusion can be applied in practice.

  4. A report on the leak before break design and evaluation of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Lee, H. Y.; Joo, Y. S.; Lee, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    In this study, the necessity of leak before break application for Liquid Metal Reactors (LMR) was investigated and the outline of leak before break evaluation procedure for KALIMER reactor structures and components was proposed. In the 1st phase of this project, the theoretical background for leak before break procedure was prepared based upon the state-of-the-art technology of the advanced countries. In the 2nd phase of project, evaluation method of creep crack growth was studied and the leak before break evaluation procedure of French RCC-MRA -16 was analyzed. Also, creep-fatigue crack growth was assessed according to the Japanese JNC method and this will help to establish the KALIMER leak before break procedure. It is necessary to specify proposed KALIMER leak before break evaluation technique including high temperature crack growth evaluation, stability analysis of crack growth, leak rate evaluation method, and leak detection technology. 24 refs., 12 figs., 4 tabs. (Author)

  5. Time-symmetry breaking in turbulence

    CERN Document Server

    Jucha, Jennifer; Pumir, Alain; Bodenschatz, Eberhard

    2014-01-01

    In three-dimensional turbulent flows, the flux of energy from large to small scales breaks time symmetry. We show here that this irreversibility can be quantified by following the relative motion of several Lagrangian tracers. We find by analytical calculation, numerical analysis and experimental observation that the existence of the energy flux implies that, at short times, two particles separate temporally slower forwards than backwards, and the difference between forward and backward dispersion grows as $t^3$. We also find the geometric deformation of material volumes, surrogated by four points spanning an initially regular tetrahedron, to show sensitivity to the time-reversal with an effect growing linearly in $t$. We associate this with the structure of the strain rate in the flow.

  6. BEACON/MOD3, 1-D and 2-D 2 Phase Flow and Heat Transfer in Containment, LWR LOCA

    International Nuclear Information System (INIS)

    1 - Description of problem or function: The BEACON series of programs is designed to perform a best-estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped- parameter representations for the various parts of the system. BEACON/MOD3 contains mass and heat transfer models for wall film and for wall conduction, and is suitable for the evaluation of short- term transients in PWR dry containment systems. The capability to examine the details of a two-components, two-phase flow field in one or two dimensions under nonhomogeneous, nonequilibrium conditions (unequal velocities, unequal temperatures between the two phases) allows analysis of such problems as the calculation of jet impact forces of a fluid leaving a pipe break, the motion of a large pressure wave across a compartment, the variation in flow properties as air is displaced from a compartment by steam and water, the water entrainment or de-entrainment by a high-speed vapor flow, the flow of a flashing liquid, and many other complex nonequilibrium problems of containment system analyses. 2 - Method of solution: The basic Eulerian flow solution procedure is based on the K-FIX two-dimensional two-phase numerical method. Each phase is described by its own density, velocity, and temperature as determined by separate sets of mass, momentum, and energy equations. The two phases are coupled by exchange parameters which model the exchange of mass, momentum, and energy between the two phases. The two sets of field equations are solved with a Eulerian finite- difference technique that implicitly treats the phase transitions and inter-phasic heat transfer in the pressure iteration. The implicit solution is accomplished iteratively without linearization and allows both phases to be

  7. Multi-dimensional Analysis about UPI during the LBLOCA of KORI 1 Unit

    Energy Technology Data Exchange (ETDEWEB)

    Bae, S. W.; Chung, B. D

    2008-03-15

    A multi-dimensional transient analysis during the LBLOCA of the Kori 1 Unit has been performed by using the MARS code. 40 percent break at the cold leg and single failure of one ECC pump train is assumed. The form drag coefficients for the upper plenum and the core have been designated as 0.6 and 9.39, respectively. After 30 second elapsed from the LOCA, a liquid pool is maintained at the upper plenum. The depth of ECCS water pool is predicted as about 20 % of the total height from the upper tie plate and the center line of the hot leg pipe. The downward flow dominant region is about 32.3 % of the total upper tie plate area. The accumulated ECCS bypass ratio is predicted as 27.64 % at 300 second. The peak cladding temperature is predicted as 1236.32 K and the location is 7 ft from the bottom of fuel at the inner radial region of reactor core nodalization. It can be concluded that the multi-dimensional analysis about the ECCS behaviors in the upper plenum space during the LBLOCA of Kori 1 Unit shows a meaningful agreement in the phenomenological validity.

  8. Through analysis of LOFT L2-2 by THYDE-P code, (1)

    International Nuclear Information System (INIS)

    A Through analysis of the Test L2-2 loss-of-coolant experiment (LOCE) in the Loss-of-Fluid Test (LOFT) program was made by the THYDE-P code. LOFT Test L2-2 was the first test in the Power Ascension Test Series (Test Series L2) of nuclear full double-ended cold leg break tests. THYDE-P is a computer code to analyze both blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs) and is now under verification study and modifications. Therefore, the LOFT experimental data play an important role at the present stage of the THYDE-P code. The present analysis was performed by best estimate (BE) options as sample calculation Run 30, which is a portion of a series of THYDE-P sample calculations. In this report, the calculated results are compared with the experimental data and discussed. In the present calculation, the core nodes were completely submerged with subcooled water at 55 sec. after the test initiation. It showed a good agreement with the experimental result. (author)

  9. Evolutionary Algorithms for the Detection of Structural Breaks in Time Series

    DEFF Research Database (Denmark)

    Doerr, Benjamin; Fischer, Paul; Hilbert, Astrid;

    2013-01-01

    Detecting structural breaks is an essential task for the statistical analysis of time series, for example, for fitting parametric models to it. In short, structural breaks are points in time at which the behavior of the time series changes. Typically, no solid background knowledge of the time ser...

  10. Is there anomalous isospin breaking in the $\\pi$N$rightarrow\\pi$N S-wave?

    CERN Document Server

    Laurikainen, P

    1973-01-01

    The authors perform a phase-shift analysis of pi N to pi N within T /sub lab/=(300-600) MeV tolerating breakings of the order of 10% to find out if the present data allows for isospin breaking effects and to see in which partial waves they occur. (12 refs).

  11. Breaking the cycle of abuse.

    Science.gov (United States)

    Egeland, B; Jacobvitz, D; Sroufe, L A

    1988-08-01

    The aim of this study was to identify variables that distinguish mothers who broke the cycle of abuse from mothers who were abused as children and who also abused their own children. Based on maternal interviews and questionnaires completed over a 64-month period, measures of mothers' past and current relationship experiences, stressful life events, and personality characteristics were obtained. Abused mothers who were able to break the abusive cycle were significantly more likely to have received emotional support from a nonabusive adult during childhood, participated in therapy during any period of their lives, and to have had a nonabusive and more stable, emotionally supportive, and satisfying relationship with a mate. Abused mothers who reenacted their maltreatment with their own children experienced significantly more life stress and were more anxious, dependent, immature, and depressed. PMID:3168615

  12. Z'-mediated supersymmetry breaking.

    Science.gov (United States)

    Langacker, Paul; Paz, Gil; Wang, Lian-Tao; Yavin, Itay

    2008-02-01

    We consider a class of models in which supersymmetry breaking is communicated dominantly via a U1' gauge interaction, which also helps solve the mu problem. Such models can emerge naturally in top-down constructions and are a version of split supersymmetry. The spectrum contains heavy sfermions, Higgsinos, exotics, and Z' approximately 10-100 TeV, light gauginos approximately 100-1000 GeV, a light Higgs boson approximately 140 GeV, and a light singlino. A specific set of U1' charges and exotics is analyzed, and we present five benchmark models. The implications for the gluino lifetime, cold dark matter, and the gravitino and neutrino masses are discussed. PMID:18352261

  13. Dynamical centrosymmetry breaking in graphene

    CERN Document Server

    Carvalho, David N; Biancalana, Fabio

    2016-01-01

    We discover an unusual phenomenon that occurs when a graphene monolayer is illuminated by a short and intense pulse at normal incidence. Due to the pulse-induced oscillations of the Dirac cones, a dynamical breaking of the layer's centrosymmetry takes place, leading to the generation of second harmonic waves. We prove that this result can only be found by using the full Dirac equation and show that the widely used semiconductor Bloch equations fail to reproduce this and some other important physics of graphene. Our results open new windows in the understanding of nonlinear light-matter interactions in a wide variety of new 2D materials with a gapped or ungapped Dirac-like dispersion.

  14. Symmetry Breaking in Finite Volume

    Institute of Scientific and Technical Information of China (English)

    LIU Chuan

    2000-01-01

    Spontaneous symmetry breaking is a cooperative phenomenon for systems with infinitely many degrees of freedom and it plays an essential role in quantum field theories. Lattice O(N) model is studied within the Hamiltonian approach using an adiabatic approximation. It is shown that the low-lying spectrum of the system in the broken phase can be understood by using the adiabatic, or Born-Oppenheimer approximation, which turns out to become an expansion in the inverse power of volume. In the infinite volume limit, the symmetry is broken while in the finite volume the slow rotation of the zero-momentum mode restores the symmetry and gives rise to the rotator spectrum, which has been observed in realistic Monte Carlo simulations.

  15. Supersymmetry-breaking nonlinear sigma models

    Energy Technology Data Exchange (ETDEWEB)

    Imai, Takumi, E-mail: imai@yukawa.kyoto-u.ac.jp [Yukawa Institute for Theoretical Physics, Kyoto University, Kyoto 606-8502 (Japan); Izawa, K.-I. [Yukawa Institute for Theoretical Physics, Kyoto University, Kyoto 606-8502 (Japan); Kavli Institute for the Physics and Mathematics of the Universe, University of Tokyo (WPI), Kashiwa 277-8583 (Japan); Nakai, Yuichiro [Department of Physics, Tohoku University, Sendai 980-8578 (Japan)

    2012-10-22

    We consider a novel class of constraints on chiral superfields to obtain supersymmetric nonlinear sigma models in four spacetime dimensions, which strictly combine the internal symmetry breaking with spontaneous supersymmetry breaking. The resultant massless modes can be exclusively Nambu-Goldstone bosons without their complex partners and the goldstino that is charged under the internal symmetry. The massive modes show a peculiar relation among their masses and the scales of symmetry breakings.

  16. Mutual information and spontaneous symmetry breaking

    OpenAIRE

    Hamma, A.; Giampaolo, S. M.; Illuminati, F.

    2015-01-01

    We show that the metastable, symmetry-breaking ground states of quantum many-body Hamiltonians have vanishing quantum mutual information between macroscopically separated regions, and are thus the most classical ones among all possible quantum ground states. This statement is obvious only when the symmetry-breaking ground states are simple product states, e.g. at the factorization point. On the other hand, symmetry-breaking states are in general entangled along the entire ordered phase, and t...

  17. Performance analysis of passively safe BWR with experimental and numerical simulation

    International Nuclear Information System (INIS)

    The performance of passive safety systems of a natural circulation BWR in a Large Break Loss Of Coolant Accident (LB LOCA) is evaluated with integral tests using a scaled test facility and RELAP5 (Mod3.3) code simulation. The Main Steam Line Break (MSLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) with the initial conditions given by the code simulation. The PUMA facility is designed to reproduce thermal-hydraulic phenomena during the low-pressure blowndown and long-term cooling period of the LOCA transient. The MSLB test is initialized when Reactor Pressure Vessel (RPV) depressurizes to 1 MPa (150 psi) and lasts for 8 hours. This test aims to demonstrate the performance of passive safety systems during the LB LOCA. Test results show that core heat-up is not observed during the test transient due to the function of Emergency Core Cooling System (ECCS). The containment peak pressure and temperature are below the design limit, which is mainly contributed by the function of Passive Containment Cooling System (PCCS). The MSLB accident transient has been simulated with RELAP5 code using prototypic plant mode and test facility model. The code models give reasonably accurate predictions on most system behaviors, while having some distortions for certain local phenomena. The integral test scalability and code applicability are evaluated by comparing the test data and the code simulation results, taking into consideration of the scaling methodology and code uncertainties. (author)

  18. Determination and Analysis of the Static Breaking Agent Expansion Pressure Variation%静态破碎剂膨胀压力变化规律测定与分析

    Institute of Scientific and Technical Information of China (English)

    赵杰成

    2015-01-01

    In order to fully know the expanding pressure size and variation of the static breaking agent, using resistance strain gauge measur-ing method in the laboratory, according to the test results, the static breaking agent expansion pressure are generated within 2 hours, maximum inflation pressure up to 55MPa. Static breaking agent expansion pressure generation time is short, when using the static breaking agent to break rocks, arranging the process and steps reasonably, all procedures to be finished in short time.%为全面掌握静态破碎剂产生的膨胀压力大小及变化规律, 实验室采用电阻应变片测量法测定静态破碎剂膨胀压力, 根据试验汇总结果得出静态破碎剂膨胀压力主要产生于从拌制好药剂开始反应的2h内,最大膨胀压力可达55MPa.静态破碎剂膨胀压力产生的时间较短且集中,现场利用静态破碎剂破碎岩石或松动岩层时,要合理安排现场施工工艺和步骤,拌药、装药和封孔要在较短时间内完成.

  19. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  20. On breaks of the Indian monsoon

    Indian Academy of Sciences (India)

    Sulochana Gadgil; P V Joseph

    2003-12-01

    For over a century, the term break has been used for spells in which the rainfall over the Indian monsoon zone is interrupted. The phenomenon of `break monsoon' is of great interest because long intense breaks are often associated with poor monsoon seasons. Such breaks have distinct circulation characteristics (heat trough type circulation) and have a large impact on rainfed agriculture.Although interruption of the monsoon rainfall is considered to be the most important feature of the break monsoon, traditionally breaks have been identified on the basis of the surface pressure and wind patterns over the Indian region. We have defined breaks (and active spells) on the basis of rainfall over the monsoon zone. The rainfall criteria are chosen so as to ensure a large overlap with the traditional breaks documented by Ramamurthy (1969) and De et al (1998). We have identified these rainbreaks for 1901-89. We have also identified active spells on the basis of rainfall over the Indian monsoon zone. We have shown that the all-India summer monsoon rainfall is significantly negatively correlated with the number of rainbreak days (correlation coefficient −0.56) and significantly positively correlated with the number of active days (correlation coefficient 0.47).Thus the interannual variation of the all-India summer monsoon rainfall is shown to be related to the number of days of rainbreaks and active spells identified here. There have been several studies of breaks (and also active spells in several cases) identified on the basis of different criteria over regions differing in spatial scales (e.g., Webster et al 1998; Krishnan et al 2000; Goswami and Mohan 2000; and Annamalai and Slingo 2001). We find that there is considerable overlap between the rainbreaks we have identified and breaks based on the traditional definition. There is some overlap with the breaks identified by Krishnan et al (2000) but little overlap with breaks identified by Webster et al (1998). Further