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Sample records for break loca analysis

  1. NPP Krsko small break LOCA analysis

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  2. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

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

    International Nuclear Information System (INIS)

    Hastuti, E.P.; Kuntoro, I.; Isnaini, M. D.; Sufmawan, A.

    1998-01-01

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

  4. Analysis of a large-break LOCA at lower operational modes

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, H.Y.; Jun, H.Y.; Lee, K. [Korea Electric Power Corporation, Taejon (Korea)

    2000-10-01

    To improve Technical Specifications and Emergency Operating Guidelines (EOGs) applicable at lower operational modes it is required to perform the safety analysis reflecting the operational characteristics in those modes. Because the component availability and system configurations at lower modes are different from those of power mode, the plant safety at lower modes should be confirmed through independent analyses. In the present study, a large-break loss-of-coolant accident is analyzed to evaluate the containment pressure and temperature control function for the preparation of EOGs applicable at lower modes. To reach the required shutdown condition, the plant cool-down is controlled by the secondary steam flow and auxiliary feedwater. The mass and energy releases from primary system are obtained from RELAP5/MOD3.1 calculation and the containment pressure and temperature are evaluated with CONTEMPT-LT code. The reference plant is Korean Next Generation Reactor having 4,000 MW thermal power. Two cases of cold leg LOCA initiated at Mode 3 with and without SIT operation are calculated. At the given plant conditions, all safety injection pumps are still available. The calculation at the condition of maximum mass and energy release shows that the containment pressure and temperature can be controlled within acceptable criteria, which means the operations of 2 or 4 fan coolers are the possible success paths to achieve the containment P/T control safety function. The peak cladding temperature with minimum safety injection flow does not show remarkable excursion, which implies the lower mode LOCA at Mode 3 can be bounded by the results obtained at full power from the viewpoint of ECCS performance. (author)

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

    International Nuclear Information System (INIS)

    Papini, Davide; Grgic, Davor; Cammi, Antonio; Ricotti, Marco E.

    2011-01-01

    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. Break location effects on PWR small break LOCA phenomena

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro

    1989-01-01

    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)

  7. A Demonstration of Advanced Safety Analysis Tools and Methods Applied to Large Break LOCA and Fuel Analysis for PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Laboratory; Smith, Curtis Lee [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2016-03-01

    The U.S. Nuclear Regulatory Commission (NRC) is currently proposing a rulemaking designated as 10 CFR 50.46c to revise the loss-of-coolant accident (LOCA)/emergency core cooling system acceptance criteria to include the effects of higher burnup on fuel/cladding performance. We propose a demonstration problem of a representative four-loop PWR plant to study the impact of this new rule in the US nuclear fleet. Within the scope of evaluation for the 10 CFR 50.46c rule, aspects of safety, operations, and economics are considered in the industry application demonstration presented in this paper. An advanced safety analysis approach is used, by integrating the probabilistic element with deterministic methods for LOCA analysis, a novel approach to solving these types of multi-physics, multi-scale problems.

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

    International Nuclear Information System (INIS)

    Ammirabile, L.; Walker, S.

    2002-01-01

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

  9. The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

    Directory of Open Access Journals (Sweden)

    Jang Hyung-Wook

    2017-01-01

    Full Text Available The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during loss of coolant accident analysis. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. From original LBLOCA input deck file, the nodalization of downcomer and junction connections with 4 cold legs and direct vessel injection lines are modified for reflecting the realistic cross-flow effect and real downcomer structure. The analysis results show that the peak cladding temperature of new input deck decreases more rapidly than previous result and that the drop of peak cladding temperature was advanced by application of momentum flux term in cross-flow. Additionally, the authors developed a new input deck with multi-dimensional downcomer model and ran MARS code with multi-dimensional input deck as well. By using the modified input deck, the Emergency core cooling system by-pass flow phenomena is better characterized and found to be consistent with both experimental report and regulatory guide.

  10. Estimation of LOCA break size using cascaded Fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA), which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict the progression of LOCAs in nuclear power plants. Therefore, accurate information on the LOCA break position and size should be provided to efficiently manage the accident. In this paper, the LOCA break size is predicted using a cascaded fuzzy neural network (CFNN) model. The input data of the CFNN model are the time-integrated values of each measurement signal for an initial short-time interval after a reactor scram. The training of the CFNN model is accomplished by a hybrid method combined with a genetic algorithm and a least squares method. As a result, LOCA break size is estimated exactly by the proposed CFNN model.

  11. The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

    OpenAIRE

    Jang Hyung-Wook; Lee Sang-Yong; Oh Seung-Jong; Kim Woong-Bae

    2017-01-01

    The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during ...

  12. ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-06-01

    , mixture level, temperatur kelongsong, small break LOCA, RELAP5.   ABSTRACT ANALYSIS ON THE CORE CONDITION OF AP1000 ADVANCED POWER REACTOR DURING SMALL BREAK LOCA. Accident due to the loss of coolant from the reactor boundary is an anticipated design basis event in the design of power reactor adopting the Generation II up to IV technology. Small break LOCA leads to more significant impact on safety compared to the large break LOCA as shown in the Three-Mile Island (TMI. The focus of this paper is the small break LOCA analysis on the Generation III+ advanced power reactor of AP1000 by simulating three different initiating events, which are inadvertent opening of Automatic Depressurization System (ADS, double-ended break on one of Direct Vessel Injection (DVI pipe, and 10 inch diameter split break on one of cold leg pipe. Methodology used is by simulating the events on the AP1000 model developed using RELAP5/SCDAP/Mod3.4. The impact analyzed is the core condition during the small break LOCA consisting of the mixture level occurrence and the fuel cladding temperature transient. The results show that the mixture level for all small break LOCA events are above the active core height, which indicates no core uncovery event. The development of the mixture level affect the fuel cladding temperature transient, which shows a decreasingly trend after the break, and the effectifeness of core cooling. Those results are identical with the results of other code of NOTRUMP. The resulted core cooling is also due to the function of coolant injection from passive safety feature, which is unique in the AP1000 design. In overall, the result of analysis shows that the AP1000 model developed by the RELAP5 can be used for analysis of design basis accident considered in the AP1000 advanced power reactor. Keywords: analysis, mixture level, fuel cladding temperature, small break LOCA, RELAP5.

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

    International Nuclear Information System (INIS)

    Hein, D.

    1984-01-01

    Through reactor safety studies like WASH 1400 or the ''Deutsche Risiko-Studie'' the attention has turned from large break loss of coolant accidents to small breaks because of the high contribution of this type of accidents to core meltdown. But only after the TMI-2 accident were also the main activities in the experimental fields shifted world-wide to the small break LOCAs. Since TMI numerous research programs have either been finished or are underway. This review paper presents: a classification of the various types of transients according to break size; a discussion of major physical phenomena associated with a small break LOCA, and a description of a few selected research programs and the most important results achieved. (author)

  14. CATHARE 2 analysis of the small break LOCA experiment SP-SB-03, conducted in SPES facility

    International Nuclear Information System (INIS)

    Meloni, P.

    1995-01-01

    SPES integral test facility is a scale model of a commercial three-loop PWR plant, making the simulation of a wide range of accident scenarios possible. A Small Break Loss of Coolant test was carried out in this facility in 1991 to serve as a counterpart of tests conducted on BETHSY (France), LSTF (Japan) and LOBI (EC) facilities. A post-test analysis of this test, performed with CATHARE 2 code was realized by ENEA in the framework of the co-operation ENEA-CEA on advanced reactors. This paper presents a survey of the results of the post-test calculation. (author). 5 refs, 11 figs, 3 tabs

  15. Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

    Directory of Open Access Journals (Sweden)

    Eltayeb Yousif

    2017-01-01

    Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

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

    International Nuclear Information System (INIS)

    Bayoumi, M.H.; Muir, W.C.; Middleton, P.B.

    1996-01-01

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

  17. The large break LOCA evaluation method with the simplified statistic approach

    International Nuclear Information System (INIS)

    Kamata, Shinya; Kubo, Kazuo

    2004-01-01

    USNRC published the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology to large break LOCA which supported the revised rule for Emergency Core Cooling System performance in 1989. In USNRC regulatory guide 1.157, it is required that the peak cladding temperature (PCT) cannot exceed 2200deg F with high probability 95th percentile. In recent years, overseas countries have developed statistical methodology and best estimate code with the model which can provide more realistic simulation for the phenomena based on the CSAU evaluation methodology. In order to calculate PCT probability distribution by Monte Carlo trials, there are approaches such as the response surface technique using polynomials, the order statistics method, etc. For the purpose of performing rational statistic analysis, Mitsubishi Heavy Industries, LTD (MHI) tried to develop the statistic LOCA method using the best estimate LOCA code MCOBRA/TRAC and the simplified code HOTSPOT. HOTSPOT is a Monte Carlo heat conduction solver to evaluate the uncertainties of the significant fuel parameters at the PCT positions of the hot rod. The direct uncertainty sensitivity studies can be performed without the response surface because the Monte Carlo simulation for key parameters can be performed in short time using HOTSPOT. With regard to the parameter uncertainties, MHI established the treatment that the bounding conditions are given for LOCA boundary and plant initial conditions, the Monte Carlo simulation using HOTSPOT is applied to the significant fuel parameters. The paper describes the large break LOCA evaluation method with the simplified statistic approach and the results of the application of the method to the representative four-loop nuclear power plant. (author)

  18. Modeling study of droplet behavior during blowdown period of large break LOCA based on experimental data

    International Nuclear Information System (INIS)

    Sakaba, Hiroshi; Umezawa, Shigemitsu; Teramae, Tetsuya; Furukawa, Yuji

    2004-01-01

    During LOCA (Loss Of Coolant Accident) in PWR, droplets behavior during blowdown period is one of the important phenomena. For example, the spattering from falling liquid film that flows from upper plenum generates those droplets in core region. The behavior of droplets in such flow has strong effect for cladding temperature behavior because these droplets are able to remove heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface. For safety analysis of LOCA in PWR, it is necessary to evaluate droplet diameter precisely in order to predict fuel cladding temperature changing by the calculation code. Based on the test results, a new droplet behavior model was developed for the MCOBRA/TRC code that predicts the droplet behavior during such LOCA events. Furthermore, the verification calculations that simulated some blowdown tests were performed using by the MCOBRA/TRAC code. These results indicated the validity of this droplet model during blow down cooling period. The experiment was focused on investigating the Weber number of steady droplet in the blow down phenomenon of large break LOCA. (author)

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

    International Nuclear Information System (INIS)

    2003-01-01

    -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

  20. Sensitivity Study on Analysis of Reactor Containment Response to LOCA

    International Nuclear Information System (INIS)

    Chung, Ku Young; Sung, Key Yong

    2010-01-01

    As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

  1. Sensitivity Study on Analysis of Reactor Containment Response to LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ku Young; Sung, Key Yong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2010-10-15

    As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

  2. LOFT/L3-, Loss of Fluid Test, 7. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the seventh in the NRC L3 Series of small-break LOCA experiments. A 2.5-cm (10-in.) cold-leg non-communicative-break LOCA was simulated. The experiment was conducted on 20 June 1980

  3. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

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

    International Nuclear Information System (INIS)

    Valette, Michel; Pouvreau, Jérôme; Bestion, Dominique; Emonot, Philippe

    2011-01-01

    Highlights: ► CATHARE 3 enables a three-field analysis of a LB LOCA. ► Reflooding experiments in isolated rod bundles are satisfactory predicted. ► 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 hot legs.

  5. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

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

  6. LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The second OECD LOFT experiment was conducted on 23 June 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with early pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  7. LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The third OECD LOFT experiment was conducted on 14 July 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with delayed pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  8. TRAC-PF1/MOD2 best-estimate analysis of a large-break LOCA in a 15 x 15 generic four-loop Westinghouse nuclear power plant

    International Nuclear Information System (INIS)

    Spore, J.W.; Lin, J.C.; Schnurr, N.M.; White, J.R.; Cappiello, M.C.

    1992-01-01

    Calculations of a large-break loss-of-coolant accident (LOCA) in a 15 x 15 generic four-loop Westinghouse nuclear power plant with both the TRAC-PF1/MOD1 and TRAC-PF1/MOD2 computer codes will be presented. The Transient Reactor Analysis Code (TRAC) has been developed by Los Alamos National Laboratory to provide advanced best-estimate simulations of real postulated transients in pressurized light-water reactors (LWRs) and for many related thermal-hydraulic facilities. The latest released version of TRAC is TRAC-PF1/MOD2. Significant improvements and enhancements over the MOD1 version were implemented in the MOD2 heat-transfer and constitutive models. One of the most significant improvements in the MOD2 code has been the implementation of the two-step numerics method in the three-dimensional components, which can significantly reduce run times for long, slow transients. A very important area of improvement has been in the reflood heat-transfer models. Developmental assessment results (i.e., code comparisons with experimental data) will be discussed for several separate-effects and integral test, including analysis of the Upper Plenum Test Facility (UPTF), the Cylindrical Core Test Facility (CCTF), and the Loss-of-Fluid Test Facility (LOFT). The assessment results provide information on the anticipated accuracy for the best-estimate models in the MOD2 computer code. The MOD1 to MOD2 comparison will provide an estimate for the effect of improved heat-transfer models on predicted peak cladding temperatures

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

    International Nuclear Information System (INIS)

    Orive Moreno, Raul; Gallego Cabezon, Ines; Garcia Sedano, Pablo

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Brittain, I.; Aksan, S.N.

    1990-08-01

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

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

    International Nuclear Information System (INIS)

    Glaeser, H.G.

    2004-01-01

    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

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

    International Nuclear Information System (INIS)

    Valette, Michel; Pouvreau, Jerome; Bestion, Dominique; Emonot, Philippe

    2009-01-01

    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 an isolated rod bundle mockup 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 6-equation model is used in the other parts of the loop. A short analysis of the results is presented. (author)

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

    International Nuclear Information System (INIS)

    Sdouz, Gert

    2006-01-01

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

  14. The phenomenology of a small break LOCA in a complex thermal hydraulic loop

    International Nuclear Information System (INIS)

    Di Marzo, M.; Almenas, K.K.; Hsu, Y.Y.; Wang, Z.

    1988-01-01

    A phenomenological description of the thermal hydraulics events that take place during a simulated Small Break Loss of Coolant Accident (SB-LOCA) is presented. The SB-LOCA transient is described in detail and the various mass and energy transport modes are identified. Similar behavior is observed in other facilities designed for the simulation of this type of accidents. Previous investigations suggest a simple modelling of the phenomena based on fluid mechanic considerations. An extensive experimental program conducted at the experimental facility of the University of Maryland reveals that condensation is a dominant driving force for this type of transients. This finding has significant implications in the modelling of enthalpy transport for some of the flow modes which occur during the transient. In particular it affects the Interruption and Resumption Mode (IRM) during which enthalpy is transported by periodic flow of a two phase mixture. The efforts to predict the flow interruption based on fluid mechanic criteria of phase separation in the hot leg are shown to be misdirected since thermodynamic phenomena taking place in the horizontal portion of the cold legs and in the reactor vessel downcomer are mostly responsible for that transition. For flow resumption to occur the liquid-vapor mixture swelling in the vertical portion of the hot leg determines the occurrence of the liquid spill over the top of the candy cane. (orig.)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  16. Temporary core liquid level depression during cold-leg small-break LOCA effect of break size and power level

    International Nuclear Information System (INIS)

    Koizumi, Y.; Kumamaru, H.; Mimura, Y.; Kukita, Y.; Tasaka, K.

    1989-01-01

    Cold-leg small break LOCA experiments (0.5-10% break) were conducted at the large scale test facility (LSTF), a volumetrically-scaled (1/48) simulator of a PWR, of the ROSA-IV Program. When a break area was less than 2.5% of the scaled cold-leg flow area, the core liquid level was temporarily further depressed to the bottom elevation of the crossover leg during the loop seal clearing early in the transient only by the manometric pressure balance since no coolant remained in the upper portion of the primary system. When the break size was larger than 5%, the core liquid level was temporarily further depressed lower than the bottom elevation of the crossover leg during the loop seal clearing since coolant remained at the upper portion of the primary system; the steam generator (SG) U-tube upflow side and the SG inlet plenum, due to counter current flow limiting by updrafting steam while the coolant drained. The amount of coolant trapped there was dependent on the vapor velocity (core power); the larger the core power, the lower the minimum core liquid level. The RELAP5/MOD2 code reasonable predicted phenomena observed in the experiments. (orig./DG)

  17. LOFT/L3-6, Loss of Fluid Test, 6. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the sixth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were running. The experiment was conducted on 10 December 1980

  18. LOFT/L3-5, Loss of Fluid Test, 5. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1991-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the fifth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were shut off. The experiment was conducted on 29 September 1980

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

    International Nuclear Information System (INIS)

    1996-07-01

    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

  20. Prototype application of best estimate and uncertainty safety analysis methodology to large LOCA analysis

    International Nuclear Information System (INIS)

    Luxat, J.C.; Huget, R.G.

    2001-01-01

    Development of a methodology to perform best estimate and uncertainty nuclear safety analysis has been underway at Ontario Power Generation for the past two and one half years. A key driver for the methodology development, and one of the major challenges faced, is the need to re-establish demonstrated safety margins that have progressively been undermined through excessive and compounding conservatism in deterministic analyses. The major focus of the prototyping applications was to quantify the safety margins that exist at the probable range of high power operating conditions, rather than the highly improbable operating states associated with Limit of the Envelope (LOE) assumptions. In LOE, all parameters of significance to the consequences of a postulated accident are assumed to simultaneously deviate to their limiting values. Another equally important objective of the prototyping was to demonstrate the feasibility of conducting safety analysis as an incremental analysis activity, as opposed to a major re-analysis activity. The prototype analysis solely employed prior analyses of Bruce B large break LOCA events - no new computer simulations were undertaken. This is a significant and novel feature of the prototyping work. This methodology framework has been applied to a postulated large break LOCA in a Bruce generating unit on a prototype basis. This paper presents results of the application. (author)

  1. LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, Thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCE is expected to closely model a LPWR LOCA. 2 - Description of test: Experiment LP-LB-1 was conducted on 3 February 1984 in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Cooperation and Development. The primary objectives of Experiment LP-LB-1 were to determine system transient characteristics and to assess code predictive capabilities for design basis large-break loss-of-coolant accidents in pressurized water reactors (PWRs). This experiment simulated a double-ended offset shear of one inlet pipe in a four-loop PWR and was initiated from conditions representative of licensing limits in a PWR. Other boundary conditions for the simulation were loss of offsite power, rapid primary coolant pump coast down, and United Kingdom minimum safeguard emergency core coolant injection rates. The nuclear fuel rods were not pressurized. The transient was initiated by opening the quick-opening blowdown valves in the broken loop hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

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

    International Nuclear Information System (INIS)

    1988-12-01

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

  3. LOFT/L2-5, Loss of Fluid Test, 3. NRC L2 Large Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the third of the NRC L2 Series of nuclear large Break LOCA experiments, conducted on 16 June 1981. It simulated a 100% cold leg break with a maximum heat generation of 40 kW/m and rapid pump coast down

  4. LOFT/L2-3, Loss of Fluid Test, 2. NRC L2 Large Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the second of the NRC L2 Series of nuclear large Break LOCA experiments, and was conducted on 12 May 1979. It simulated a 100% cold leg break with a maximum heat generation of 39 kW/m

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

    International Nuclear Information System (INIS)

    2012-08-01

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

  6. LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressure injection System (HPIS)

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The sixth OECD LOFT experiment was conducted on 5 March 1984. It simulated a 1.8-in cold leg break LOCA with no HPIS available. This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncover and addressed accumulator injection at low pressure differentials. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  7. THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2001-01-01

    1 - Description of problem or function: THYDE-P1 analyzes the behaviour of LWR plants in response to various disturbances, including the thermal hydraulic transient following a break of the primary coolant pipe system, generally referred to as a loss-of-coolant-accident (LOCA). LOCA can be considered as the most critical condition for testing the methods and models for plant dynamics, since thermal hydraulic conditions in the system change drastically during the transient. THYDE-P is capable of a complete LOCA calculation from start to complete reflooding of the core by subcooled water. The program performs steady-state adjustment, which is complete in the sense that the steady state obtained is a set of exact solutions of all the transient equations without time derivatives, not only for plant hydraulics but also for all the other phenomena in the simulation of a PWR plant. THYDE-P2 contains among others the following improvements over THYDE-P1: (1) not only the mass and momentum equations but also the energy equation are included in the non-linear implicit scheme; (2) the valve model is implemented; (3) the relaxation equation for void fraction is theoretically derived; (4) vectorized programming is implemented; (5) both EM (evaluation mode) and BE (best estimate) calculations are possible. THYDE-W is an improved version of THYDE-P2 and contains the following additional features: (a) analysis of multiple number of disjoint loops is possible; (b) a control system simulation model is included; (c) the trip model has been improved; (d) heavy water is allowed as coolant; (e) the effect of drift flux is accounted for in the steady state calculation; (f) boron transport is included; (g) to obtain steady state loop heat balance, the option of adjusting the enthalpy distribution is prepared included in addition to that of adjusting heat exchanger areas; (h) to obtain steady state pressure distribution, three other options are prepared in addition to the original ones

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

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

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

  9. 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.

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

    Science.gov (United States)

    Kinoshita, Ikuo; Torige, Toshihide; Yamada, Minoru

    2014-06-01

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

  11. IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS

    Directory of Open Access Journals (Sweden)

    DONG HYUN LEE

    2014-08-01

    Full Text Available 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&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.

  12. Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

    International Nuclear Information System (INIS)

    Arshad, M. W.

    2012-01-01

    Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

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

    International Nuclear Information System (INIS)

    Wilkowski, G.; Rudland, D.; Wolterman, R.; Krishnaswamy, P.; Scott, P.; Rahman, S.; Fairbanks, C.

    2002-01-01

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

  14. Analysis of BWR high burnup fuel in LOCA conditions

    International Nuclear Information System (INIS)

    Garcia Sedano, Pablo; Dey Navarro, Jose Manuel; Gallego Cabezon, Ines; Orive Moreno, Raul

    2004-01-01

    High Burnup Fuel Behaviour has been growing in importance since middle 80's when pellet microstructure changes (rim effect) and cladding oxidation rates increase were observed. Later on, Cadarache reactivity tests revealed cladding integrity failures below safety limits. These phenomena, occurred at high burnup, stressed the necessity of having a wide experimental data base that would allow to dispose non-extrapolated data of material properties submitted to higher burnups than 40000 MWd/TM and data of new materials at the same time. One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MTU rod average burnup. The technical bases to support those high burnup levels are being developed. One of the licensing points of concern is the behaviour of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/TM and H.B. Robinson at 67 GWd/MTU. When the ANL tests have been finished, a conservative Peak Cladding Temperature/ Equivalent Cladding Reacted (PCT/ECR) limit will be determine from the residual ductility tests to be applied to the high burnup fuel. This makes necessary to determine the behaviour of the high burnup fuel in LOCA conditions and to determine the available safety margin. In licensing LOCA calculations, corresponding to present core designs and future core designs, the calculated PCT and ECR values as a function of the fuel burnup could be used to determine the relative severity of LOCA for the high burnup fuel. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations. (authors)

  15. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

    Directory of Open Access Journals (Sweden)

    Li Yuquan

    2017-02-01

    Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

  16. Analysis of factors affecting the LOCA test quality

    International Nuclear Information System (INIS)

    Wang Lu

    2008-01-01

    Localization of nuclear safety-related equipment has become an important way of nuclear power development in China. To meet this demand, the competence should be promoted in the following two areas, one is to develop the capability of R and D and manufacturing of nuclear safety-related equipment, the other is to implement equipment qualification according to relevant codes and standards. As LOCA test is one of the most important parts in the qualification test of nuclear safety-related equipment, the main factors related with the quality of the LOCA test are analyzed in this paper, and this may be a reference to improve the skills in designing, constructing and operating LOCA test devices. (authors)

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

    International Nuclear Information System (INIS)

    1998-01-01

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

  18. Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

    2012-01-01

    Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

  19. Analysis of the large break loss of coolant accidents in nuclear power plants by the computer code RELAP4/MOD6

    International Nuclear Information System (INIS)

    Gregoric, M.; Stritar, A.

    1983-01-01

    The safety analysis of the nuclear power plant Krsko by the code RELAP4/MOD6 is described. Methodology for the safety evaluation for the case of the Large LOCA is introduced. The problems encountered during the analysis of the blowdown phase of the accident are described. Some results of double ended cold leg LOCA analysis for different break sizes are shown. (author)

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

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  1. Complete BWR--EM LOCA analysis using the WRAP--EM system

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

    1979-01-01

    The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

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

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Kim, Soo Hyung; Lee, Gyu Hyeung; Lee, Won Jae

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  4. Analysis of LOCA experiments with RELAP4J code

    International Nuclear Information System (INIS)

    Mochizuki, Yooji; Sobajima, Makoto; Suzuki, Mitsuhiro.

    1978-09-01

    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)

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

    International Nuclear Information System (INIS)

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

    1993-03-01

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

  6. 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

  7. IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU

    International Nuclear Information System (INIS)

    Cunningham, Mitchel E.; Turnbull, J.A.

    2003-01-01

    Description - Objectives - Results: The U.S. Nuclear Regulatory Commission (NRC) conducted a series of thermal-hydraulic and cladding mechanical deformation tests in the National Research Universal (NRU) reactor at the Chalk River National Laboratory in Canada. The objective of these tests was to perform simulated loss-of-coolant-accident (LOCA) experiments using full-length light-water reactor fuel rods to study mechanical deformation, flow blockage, and coolability. Three phases of a LOCA (i.e., heat-up, reflood, and quench) were performed in situ using nuclear fissioning to simulate the low-level decay power during a LOCA after shutdown. All tests used PWR-type, non-irradiated fuel rods. Provided here is information on two materials tests, MT-6A and MT-4, which PNNL considers the better characterized for the purposes of setting up computer cases. The NRU reactor is a heterogeneous, thermal, tank-type research reactor. It has a power level of 135 MWth and is heavy-water moderated and cooled. The coolant has an inlet temperature of 310 K at a pressure of 0.65 MPa. The MT tests were conducted in a specially designed test train to supply the specified coolant conditions of flowing steam, stagnant steam, and then reflood. Typical instrumentation for the MT tests included fuel centerline thermocouples, cladding inner surface thermocouples, cladding outer surface thermocouples, rod internal gas pressure transducers or pressure switches, coolant channel steam probes, and self-powered neutron detectors. This instrumentation allowed for determining rupture times and cladding temperature. The test rods for the LOCA cases in the NRU reactor were irradiated in flowing steam prior to the transient, stagnant steam during the transient and prior to reflood, and then reflood conditions to complete the transient. Both cladding inner surface and outer surface temperatures were measured, in addition to coolant temperatures. However, only cladding inner surface temperatures were

  8. 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.

  9. 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.

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

    International Nuclear Information System (INIS)

    Axford, D.J.; Martin, I.C.; McAuley, S.J.

    1981-04-01

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

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

    International Nuclear Information System (INIS)

    Leach, L.P.; Ybarrondo, L.J.; Hicken, E.F.; Tasaka, K.

    1983-01-01

    The bases for loss-of-coolant accident (LOCA) safety analysis required by reactor licensing regulations in the United States of America (USA), Federal Republic of Germany (FRG), and Japan are investigated and related to new data obtained since the regulations were established. The licensing approaches used in the three countries are similar in that a conservative calculation is called for, the necessary conservatism is unspecified, and new research data have had only limited effect on changing the regulations

  12. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  13. 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)

    1992-02-01

    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

  14. User's guide for the PWR LOCA analysis capability of the WRAP-EM system

    Energy Technology Data Exchange (ETDEWEB)

    Beranek, F; Gregory, M V

    1980-02-01

    The Water Reactor Analysis Package (WRAP) has been expanded to provide the capability to analyze loss-of-coolant accidents (LOCAs) in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) by using evaluation models (EMs). The input specifications for modules in the WRAP-EM system are presented in this document along with the JOSHUA input templates. This document, along with the WRAP user's guide, provides a step-by-step procedure for setting up a PWR data base for the WRAP-EM system. 12 refs.

  15. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    1 - Description of program or function: THALES, which stands for 'Thermal Hydraulic Analysis of Loss-of-coolant, Emergency core cooling and Severe core damage', is a computer code system for analyzing progression of core melt accident of light water reactors. The code was developed for Level 2 PSA (probabilistic safety assessment) and applicable to a wide range of postulated accident scenarios. Its outcomes are thermal hydraulic conditions in the reactor coolant system and the containment which are necessary for analyzing fission product release and transport behavior during the accident. The code system consists of following three member codes: (1) THALES-PM for accident progression in the primary and the secondary system of PWRs, (2) THALES-BM for accident progression in the reactor coolant system of BWRs, and (3) THALES-CV for accident progression in the containment of PWRs and BWRs. The THALES-PM and the THALES-BM codes carry out two categories of analysis. The first one is overall thermal-hydraulic analysis in the reactor coolant system. The reactor coolant system is divided into multi-volumes and each volume is further separated into a liquid region and a gas region by a movable mixture level. System pressure, mixture level in each volume, coolant temperature in each region, flow rate between volumes, etc. are calculated. The other one is core heatup and meltdown analysis. The reactor core is radially and axially divided into many nodes. Fuel and cladding temperature, cladding oxidation rate, hydrogen generation rate, core melt fraction, etc. are calculated. The THALES-CV code is for containment response analysis. It divides the containment into multiple compartments, each of which is further separated into a liquid region and a gas region by a movable mixture level. Containment pressure, mixture level in each compartment, coolant temperature in each region, flow rate between compartments, etc. are calculated. The code can treat coolant blowdown from the

  16. 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)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2010-02-01

    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)

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

    International Nuclear Information System (INIS)

    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

    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)

  18. 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).

  19. Space-time neutronic analysis of postulated LOCA's in CANDU reactors

    International Nuclear Information System (INIS)

    Luxat, J.C.; Frescura, G.M.

    1978-01-01

    Space-time neutronic behaviour of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated LOCA'S. The approach involves the efficient use of a set of computer codes which provide a capability to perform simulations ranging from detailed, accurate 3-dimensional space-time to low-cost survey calculations using point kinetics with some ''effective'' spatial content. A new, space-time kinetics code based upon a modal expansion approach is described. This code provides an inexpensive and relatively accurate scoping tool for detailed 3-dimensional space-time simulations. (author)

  20. Analysis of the influence of steam generator tube plugging on the large break loss of coolant accident in NPP Krsko

    International Nuclear Information System (INIS)

    Bizjak, S.; Stritar, A.

    1987-01-01

    The preliminary analysis of the influence of steam generator tube plugging to the large break LOCA behaviour of the NPP Krsko was performed. If 10% of the tubes are plugged, the peak cladding temperature reached is 37 K higher than the temperature reached after LOCA if no tubes were plugged. The decrease of the maximum peaking factor from 2.34 to 2.25 would compensate the influence of 10% plugged tubes. The analysis was not fully in compliance with the requirements of the conservative methodology. (author)

  1. Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yurko, Joseph P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swindlehurst, Gregg [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

  2. Effect of fuel pin ballooning on the sub-channel thermal hydraulics during small break loca for Indian PHWRS

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, D.; Behera, G.H.; Bandopadhyay, S.K.; Gupta, S.K. [Bhabha Atomic Research Centre, Div. Reactor Safety, Bombay (India)

    2001-07-01

    Effect of fuel pin ballooning on the subchannel thermal-hydraulics during a small break (0.25%) located at the Reactor Inlet Feeder (RIF) has been studied for Indian PHWRs. The break leads to a low flow situation in the affected reactor channel along with delayed reactor trip. Higher power to flow ratio in the inner subchannels in comparison to outer subchannel of a 19 pin fuel bundle causes early 2-phase condition causing the flow to by pass from the inner ones to outer ones. This causes the fuel pins to experience different temperatures. Fuel pin ballooning causes reduction in the subchannel areas and further flow redistribution takes place. The transient subchannel thermal-hydraulic conditions along the reactor channel are very much different due to the power distribution and pressure drop. (authors)

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

    International Nuclear Information System (INIS)

    Kumamaru, Hiroshige; Nakamura, Hideo; Hirata, Kazuo

    1989-03-01

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

  4. Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code

    International Nuclear Information System (INIS)

    Perianez Alvarez, V.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Nguyen, V.T.; Kieu, N.D.

    2015-01-01

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

  6. Fuel behavior during a LOCA: LOFT experiments

    International Nuclear Information System (INIS)

    Russell, M.L.

    1980-11-01

    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

  7. Nuclear power plant emergency core cooling system reliability analysis - reliability estimation for small LOCA

    International Nuclear Information System (INIS)

    Vojnovic, Dj.

    1989-01-01

    System performance reliability depends not only on its own availability but also on requirements which are placed the system. This paper shows a way of system performance reliability estimation for a NPP Emergency Core Cooling System in case of small LOCA. The event scenario and requirements for systems are determined with event tree. Finally, the ECCS reliability estimation is performed on the basis of system requirements. (author)

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

    International Nuclear Information System (INIS)

    Kim, V.; Kuznetsov, V.; Balakan, G.; Gromov, G.; Krushynsky, A.; Sholomitsky, S.; Lola, I.

    2007-01-01

    The paper presents the results of the study conducted to support planned modernization of the South Ukraine nuclear power plant. The objective of the analysis has been to develop the automated emergency control algorithm for primary to secondary LOCA accident for SUNPP WWER-1000 safety upgrading. According to the analyses performed in the framework of safety assesment report, given accident is the most complex for control and has the largest contribution into the core damage frequency value. This is because of initial event diagnostics is difficult, emergency control is complicated for personnel, time available for decision making and actions performing is limited with coolant inventory for make-up, probability of steam dump valves on affected steam generator non-closing after opening is high, and as a consequence containment bypass, irretrievable loss of coolant and radioactive materials release into the environment are possible. Unit design modifications are directed on expansion of safety systems capabilities to overcome given accident and to facilitate the personnel actions on emergency control. Safety systems modification according to developed algorithm will allow to simplify accident control by personnel and enable to control the ECCS discharge limiting pressure below the affected steam generator steam dump valve opening pressure, and decrease the probability of the containment bypass sequences. The analysis of the primary-to-secondary LOCA thermal-hydraulics has been conducted with RELAP5/Mod 3.2, and involved development of the dedicated analytical model, calculations of various plant response accident scenarios, conducting of plant personnel intervention analyses using full-scale simulator, development and justification of the emergency control algorithm aimed on the minimization of negative consequences of the primary-to-secondary LOCA (Authors)

  9. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  10. A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

    The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

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

    International Nuclear Information System (INIS)

    Andrade, Delvonei Alves de; Sabundjian, Gaiane

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. A Study on the Development of Simplified Fuel Assembly SSE/LOCA Analysis Model using Optimization Technique

    International Nuclear Information System (INIS)

    Lee, Kyou Seok; Jeon, Sang Youn; Kim, Hyeong Koo

    2009-01-01

    Under the Safe Shutdown Earthquake (SSE) and Loss of Coolant Accident (LOCA) events, the fuel assembly deflection and impact force between fuel assemblies are obtained by the dynamic transient analysis for the reactor core model. The impact behavior between fuel assemblies shows non-linear characteristics, because fuel assembly shows non-linearly dynamic characteristics and its geometry is complicated. Furthermore, since a reactor core consists of a large number of fuel assemblies, the dynamic behavior of the core under the postulated events is very difficult to analyze. Therefore, it is necessary that fuel assembly model be simplified considering dynamic non-linear characteristics in core analysis. In this study, a simplified fuel assembly finite element model for 17 Type RFA has been developed using optimization technique. To obtain the simplified model, the optimization algorithm of ANSYS was used, and the model was verified by comparison with fuel assembly mechanical test results

  14. A Study on the Development of Simplified Fuel Assembly SSE/LOCA Analysis Model using Optimization Technique

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyou Seok; Jeon, Sang Youn; Kim, Hyeong Koo [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2009-05-15

    Under the Safe Shutdown Earthquake (SSE) and Loss of Coolant Accident (LOCA) events, the fuel assembly deflection and impact force between fuel assemblies are obtained by the dynamic transient analysis for the reactor core model. The impact behavior between fuel assemblies shows non-linear characteristics, because fuel assembly shows non-linearly dynamic characteristics and its geometry is complicated. Furthermore, since a reactor core consists of a large number of fuel assemblies, the dynamic behavior of the core under the postulated events is very difficult to analyze. Therefore, it is necessary that fuel assembly model be simplified considering dynamic non-linear characteristics in core analysis. In this study, a simplified fuel assembly finite element model for 17 Type RFA has been developed using optimization technique. To obtain the simplified model, the optimization algorithm of ANSYS was used, and the model was verified by comparison with fuel assembly mechanical test results.

  15. Sump water usability analysis following LB LOCA of CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, M.S. [Nuclear Engineering Service & Solution, Daejeon (Korea, Republic of); Kim, S.M. [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of); Moon, B.J.; Kim, S.R. [Nuclear Engineering Service & Solution, Daejeon (Korea, Republic of)

    2014-07-01

    This paper focused on the analysis of sump water usability as a source for low pressure emergency core cooling injection in CANDU 6 for large break loss of coolant accident, using GOTHIC-IST code. For a long term cooling, the operation of low pressure recirculation using an emergency core cooling pump is required. To operate an emergency core cooling pump, the net positive suction head of the pump should be satisfied. The maximum permissible temperature of sump water to meet the net positive suction head of an emergency core cooling pump is 87.73{sup o}C. In this study, the temperature and the level of sump water were monitored for the large break loss of coolant accident with malfunction of spray system and local air coolers. For all considered accident cases, the temperature of containment basement water was analyzed to be lower than 87.73{sup o}C and it was possible to operate the low pressure recirculation using an emergency core cooling pump for the most restricted scenario. (author)

  16. CANDU fuel behaviour under LOCA conditions

    International Nuclear Information System (INIS)

    Kohn, E.

    1989-07-01

    This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

  17. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  18. Post accidental small breaks analysis

    International Nuclear Information System (INIS)

    Depond, G.; Gandrille, J.

    1980-04-01

    EDF ordered to FRAMATOME by 1977 to complete post accidental long term studies on 'First Contrat-Programme' reactors, in order to demonstrate the safety criteria long term compliance, to get information on NSSS behaviour and to improve the post accidental procedures. Convenient analytical models were needed and EDF and FRAMATOME respectively developped the AXEL and FRARELAP codes. The main results of these studies is that for the smallest breaks, it is possible to manually undertake cooling and pressure reducing actions by dumping the steam generators secondary side in order to meet the RHR operating specifications and perform long term cooling through this system. A specific small breaks procedure was written on this basis. The EDF and FRAMATOME codes are continuously improved; the results of a French set of separate effects experiments will be incorporated as well as integral system verification

  19. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

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

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2017-08-01

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

  2. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

    Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

    1992-01-01

    The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox design (Oconee) and a Westinghouse 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 s for the Babcock and Wilcox and Westinghouse plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1

  3. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

    Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

    1991-10-01

    The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

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

    International Nuclear Information System (INIS)

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

    2005-06-01

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

  5. 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)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Kumamaru, Hiroshige; Anoda, Yoshinari; Yonomoto, Taisuke; Murata, Hideo; Tasaka, Kanji

    1990-03-01

    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)

  6. 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.

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

    International Nuclear Information System (INIS)

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

    2006-03-01

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

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

    Science.gov (United States)

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

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

  9. 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)

  10. Thermal-hydraulic Analysis of LOCA to Apply PSA Using MAAP and MARS codes

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yun Je; Kim, Tae Jin; Park, Goon Cherl [Seoul National University, Seoul (Korea, Republic of); Lim, Ho Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    Thermal-hydraulic analysis in Probabilistic Safety Assessment (PSA) is performed to product basic data, which are needed to analyze accident sequence, construct system fault tree and evaluate operator error. Through the thermal-hydraulic analysis, the reactor power level, the pressure and the temperature of primary side, and the pressure, the temperature and the water level of secondary side are calculated. From these data, system success criteria for construction of event tree and the allowable outrage time for human reliability analysis are determined. Until now, system codes such as MAAP, RELAP, MELCOR, RETRAN have been widely used for thermal-hydraulic analysis in PSA. The adequacy of analysis is dependent on the type of accident and the models of codes. As a first step of 'Study on Best-Estimate Thermal-Hydraulic Analysis Methodology Applicable to Probabilistic Safety Assessment', a part of National Nuclear Technology Program of Ministry of Science and Technology, it is required to compare the result of MARS analysis with that of MAAP analysis previously performed, and to evaluate its applicability to PSA.

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

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Liao, J.; Frepoli, C.; Ohkawa, K.

    2011-01-01

    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)

  13. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    Science.gov (United States)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Perianez Alvarez, V.

    2013-07-01

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

  15. Development of Non-LOCA Safety Analysis Methodology with RETRAN-3D and VIPRE-01/K

    International Nuclear Information System (INIS)

    Kim, Yo-Han; Cheong, Ae-Ju; Yang, Chang-Keun

    2004-01-01

    Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report

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

    International Nuclear Information System (INIS)

    Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

    1984-01-01

    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 code combines a transient 2-dimensional heat conduction code and a 1-dimensional mechanical model for the cladding deformation. The first sections of this report deal with the heat conduction model and the finite element discretization used for the thermal analysis. The mechanical deformation model is presented next: modelling of creep, phase change and oxidation of the zircaloy cladding is discussed in detail. A model describing the effect of oxidation and oxide cracking on the mechanical strength of the cladding is presented too. Next a mechanical restraint model, which allows the simulation of the presence of the neighbouring rods and is particularly important in assessing the amount of channel blockage during a transient, is presented. A description of the models used for the coolant conditions and for the power generation follows. The heat source can be placed either in the fuel or in the cladding, and direct or indirect clad heating by electrical power can be simulated. Then a section follows, dealing with the steady-state and transient types of calculation and with the automatic variable time step selection during the transient. The last sections deal with presentation of results, graphical output, test problems and an example of general application of the code

  17. 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.

  18. The analysis of 14.8 percent cold leg break without the application of hydroaccumulators in the PMK-NHV test facility

    International Nuclear Information System (INIS)

    Szabados, L.; Ezsoel, Gy.; Perneczky, L.

    1990-12-01

    A series of reactor safety tests have been performed in the experimental reactor simulation facility PMK-NHV of the Paks Nuclear Power Plant, Hungary, with and without the use of hydroaccumulator (SIT) operation. 14.8 percent cold leg break simulation experiments are reported without SITs in action, and the measurement results were analyzed using the RELAP5/mod2 computer code. The description of the experiment is followed by the parameter variations and their analysis, together with an interpretation of the measurement results. The lessons from the LOCA simulation tests are evaluated. (R.P.) 10 refs.; 48 figs.; 3 tabs

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-15

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

  1. Study for Relation of Pressure and Aging Degradation during LOCA Test

    International Nuclear Information System (INIS)

    Kim, Jong Seog

    2013-01-01

    As result of this test, it was found that low pressure effect in aging was not significant compared with that of temperature. If temperature profile in LOCA test can satisfy the plant LOCA profile, no further analysis of pressure profile for aging degradation is necessary. For environmental qualification of electric equipment in containment building of nuclear power plant, LOCA test should be applied. During the LOCA test, temperature and pressure of LOCA chamber shall be controlled to meet a requirement of plant specific LOCA profile. It is general to keep LOCA test temperature and pressure above the plant specific LOCA profile. If the test temperature is lower than required profile in some time zone while it is higher in other time zone, calculation of total cumulated test temperature is required to compare with that of plant profile. Arrhenius equation can be applied for calculation of total temperature accumulation. If there is a deviation of pressure between test profile and plant specific profile, can we still use the same rule of temperature? Since the Arrhenius equation can't be applied to pressure, analysis of pressure effect to aging degradation is not easy. Study for relation of pressure and aging degradation during LOCA condition is described herein. To Study an aging degradation effect of pressure during LOCA test, comparison of IR during high LOCA pressure and low LOCA pressure were implemented. We expected low IR in high pressure because it contained a high concentration of oxygen which induces high aging degradation. Contrary to our expectation, IR of low pressure was lower than that of high pressure. It is assumed that high vibration of temperature profile to maintain the low pressure at high temperature induced supply of high enthalpy steam into LOCA chamber

  2. Best estimate analysis of GDH break in RBMK-1500

    International Nuclear Information System (INIS)

    Vileiniskis, V.; Kaliatka, A.; Uspuras, E.

    2004-01-01

    The RBMK-1500 is a multichannel boiling water and graphite moderated nuclear reactor. Two identical cooling loops compose the main circulation circuit, which consists of 1661 parallel fuel channels (FC) and numerous components, such as collectors, headers, pumps, etc. There are 20 group distribution headers (GDH) in the each cooling loop. Guillotine or partial break of one GDH, which in turn feeds 38-43 FC, leads to the sharp coolant flow rate decrease in the FC connected to the affected GDH. In the case of partial GDH break the short-term flow stagnation in the group of fuel channels connected to ruptured GDH can occur. Thus, any GDH break is accompanied by fuel cladding and fuel channel temperature increase. This paper presents best estimate analysis of possible GDH breaks in Ignalina Nuclear Power Plant RBMK-1500 reactor: Guillotine break upstream check valve, Guillotine break downstream check valve, Partial break. Best estimate code RELAP5/MOD3.2 was used for deterministic analysis. According to the international practice if a best estimate code is used for the analysis, the uncertainty and sensitivity analysis should be performed. Sensitivity and uncertainty analysis was performed employing System for Uncertainty and Sensitivity Analysis (SUSA) developed at GRS (Germany). Results of the analysis shows that acceptance criterion for fuel cladding temperature is violated only short time period in the case of GDH guillotine break, in the case of partial GDH break the reactor core will be reliably cooled.(author)

  3. ICECON: a computer program used to calculate containment back pressure for LOCA analysis (including ice condenser plants)

    International Nuclear Information System (INIS)

    1976-07-01

    The ICECON computer code provides a method for conservatively calculating the long term back pressure transient in the containment resulting from a hypothetical Loss-of-Coolant Accident (LOCA) for PWR plants including ice condenser containment systems. The ICECON computer code was developed from the CONTEMPT/LT-022 code. A brief discussion of the salient features of a typical ice condenser containment is presented. Details of the ice condenser models are explained. The corrections and improvements made to CONTEMPT/LT-022 are included. The organization of the code, including the calculational procedure, is outlined. The user's manual, to be used in conjunction with the CONTEMPT/LT-022 user's manual, a sample problem, a time-step study (solution convergence) and a comparison of ICECON results with the results of the NSSS vendor are presented. In general, containment pressure calculated with the ICECON code agree with those calculated by the NSSS vendor using the same mass and energy release rates to the containment

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

    International Nuclear Information System (INIS)

    Husaini, S. Mahmood; Qashu, Riyad K.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Huang, X.G.; Yang, Y.H.; Cheng, X.; Al-Hawshabi, N.H.A.; Casey, S.P.

    2011-01-01

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

  6. Improved Methodology of MSLB M/E Release Analysis for OPR1000

    International Nuclear Information System (INIS)

    Park, Seok Jeong; Kim, Cheol Woo; Seo, Jong Tae

    2006-01-01

    A new mass and energy (M/E) release analysis methodology for the equipment environmental qualification (EEQ) on loss-of-coolant accident (LOCA) has been recently developed and adopted on small break LOCA EEQ. The new methodology for the M/E release analysis is extended to the M/E release analysis for the containment design for large break LOCA and the main steam line break (MSLB) accident, and named KIMERA (KOPEC Improved Mass and Energy Release Analysis) methodology. The computer code systems used in this methodology is RELAP5K/CONTEMPT4 (or RELAP5-ME) which couples RELAP5/MOD3.1/K with enhanced M/E model and LOCA long term model, and CONTEMPT4/ MOD5. This KIMERA methodology is applied to the MSLB M/E release analysis to evaluate the validation of KIMERA methodology for MSLB in containment design. The results are compared with the OPR 1000 FSAR

  7. Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

    2010-01-01

    To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

  8. Development and application of an uncertainty methodology for the Sizewell B large LOCA safety case

    International Nuclear Information System (INIS)

    Lightfoot, P.; Trow, M.

    1994-01-01

    This paper presents an uncertainty methodology which has been successfully applied to the licensing of Sizewell B for large break LOCA. The emphasis of this approach has been on gaining a detailed understanding of the physical process and of the sensitivity to individual phenomena. The major contributors to uncertainty have been identified, and have subsequently been included in a combined uncertainty analysis. The combined uncertainty analysis demonstrated that uncertainties did not combine in a highly non-linear manner phenomena such as the random reflood effect and clad ballooning have been treated a bounding biases in the assessment of the overall bounding peak clad temperature. The plant initial and boundary conditions have been conservatively defined for the uncertainty analysis. A better estimate calculation, which uses more realistic assumptions, shows a large benefit in the predicted peak clad temperature, thereby demonstrating the conservatism of the uncertainty analysis. The UK licensing regime is not prescriptive in terms of the approach to large LOCA analysis, and no attempt has been made to apply a formal probability or confidence limit to the final bounding peak clad temperature is conservative. The Sizewell B uncertainty analysis was completed within the timescale and resources limitations. It has been shown to be practical in its application and reductions in the required analysis scope have been identified for any future plants of similar design

  9. Analysis of chiral symmetry breaking mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Xin-Heng, Guo [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Tao, Huang [Academia Sinica, Beijing, BJ (China). Inst. of High Energy Physics; Chuang, Wang

    1997-07-01

    The renormalization group invariant quark condensate {mu} is determinate both from the consistent equation for quark condensate in the chiral limit and from the Schwinger-Dyson (SD) equation improved by the intermediate range QCD force singular like {delta} (q) which is associated with the gluon condensate. The solutions of {mu} in these two equations are consistent. We also obtain the critical strong coupling constant {alpha}c above which chiral symmetry breaks in two approaches. The nonperturbative kernel of the SD equation makes {alpha}c smaller and {mu} bigger. An intuitive picture of the condensation above {alpha}c is discussed. In addition, with the help of the Slavnov-Taylor-Ward (STW) identity we derive the equations for the nonperturbative quark propagator from SD equation in the presence of the intermediate-range force is also responsible for dynamical chiral symmetry breaking. (author) 32 refs., 2 figs.

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

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

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

  11. Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

    International Nuclear Information System (INIS)

    Griggs, D.P.

    1993-03-01

    The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

  12. Dam-Break Flood Analysis Upper Hurricane Reservoir, Hartford, Vermont

    National Research Council Canada - National Science Library

    Acone, Scott

    1995-01-01

    .... Various dam break flood conditions were modeled and inundation maps developed. Based on this analysis the dam is rated a Class 2 or significant hazard category in terms of its potential to cause downstream damage...

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

    International Nuclear Information System (INIS)

    Aksan, N.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Ferreira Guimaraes, Antonio Cesar; Franklin Lapa, Celso Marcelo; Lamego Simoes Filho, Francisco Fernando; Cabral, Denise Cunha

    2011-01-01

    Research highlights: → This article presents an uncertainty modelling study using a fuzzy approach. → The AP1000 Westinghouse NPP was used and it is provided of passive safety systems. → The use of advanced passive safety systems in NPP has limited operational experience. → Failure rates and basic events probabilities used on the fault tree analysis. → Fuzzy uncertainty approach was employed to reliability of the AP1000 large LOCA. - Abstract: This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

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

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

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

  16. 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)

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

    International Nuclear Information System (INIS)

    Murao, Yoshio; Okubo, Tsutomu; Sugimoto, Jun; Iguchi, Tadashi; Sudoh, Takashi.

    1985-02-01

    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)

  18. Dam Break Analysis of Embankment Dams Considering Breach Characteristics

    Directory of Open Access Journals (Sweden)

    Abolfazl Shamsaei

    2004-05-01

    Full Text Available The study of dam's break, needs the definition of various parameters such as the break cause, its type, its dimension and the duration of breach development. The precise forecast for different aspects of the breach is one of the most important factors for analyzing it in embankment dam. The characteristics of the breach and determination of their vulnerability has the most effect on the waves resulting from dam break. Investigating, about the parameters of the breach in "Silveh" earth dam have been determined using the suitable model. In Silve dam a trapezoid breach with side slope z=0.01m and the average base line b=80m was computed. The duration of the breaches development is 1.9 hour. Regarding the above results and the application of DAM Break software the consequences of the probable break of the dam was determined. The analysis of the results of water covering of the city of Piranshahr located 12km from silve dam confirms that in 3 hours the water will reach the height (level of 1425 meters.

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

    International Nuclear Information System (INIS)

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

    2007-03-01

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

  20. Multiaspect measurement analysis of breaking energy recovery

    International Nuclear Information System (INIS)

    Bartłomiejczyk, Mikołaj; Połom, Marcin

    2016-01-01

    Highlights: • A case study of implementation of eco energy technologies in municipal transport. • The “ready to use” methods are presented. • The “niche” ways of increasing efficiency, e.g. “intelligent heating”. • Novel multi way measurement method using GPS localization system. • Confirmation of the results by means of research and experimental measurement. - Abstract: Nowadays the issue of electric energy saving in public transport is becoming a key area of interest, which is connected both with a growth of environmental awareness in the society and an increase in the prices of fuel and electricity. That is why the reduction of energy consumption by increasing electrified urban transport, such as trams, trolleybuses, light rail and underground is becoming an increasingly important issue. Energy recovery during braking is possible in all modern electric vehicles, but in many cases this possibility is not fully taken advantage of, inter alia, because of an inadequate power supply structure. The aim of this article is to present practical examples of implementation of eco-friendly solutions in urban municipal transport. The article shows a thorough analysis of braking energy dispatch in the urban traction power supply system, which was based on extensive measurement research conducted in Gdynia trolleybus network. The authors applied multi way measurement method using Global Positioning System. The optimal conditions for implementation of several methods of energy recovery (storage energy systems, reconfiguration of supply system, using auxiliaries) have been shown. Great emphasis has been put on the confirmation of the results by means of research and experimental measurement.

  1. 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)

    Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

    1996-09-01

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

  2. Scaling effects concerning the analysis of small break experiments

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-11-01

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

  4. Debris transport evaluation during the blow-down phase of a LOCA using computational fluid dynamics

    International Nuclear Information System (INIS)

    Park, Jong Pil; Jeong, Ji Hwan; Kim, Won Tae; Kim, Man Woong; Park, Ju Yeop

    2011-01-01

    Highlights: → We conducted CFD simulation on the spreading of the coolant in the containment after a break of the hot leg. It is used to estimate the dispersion of the debris within the containment. → It was assumed that the small and fine debris is transported by the discharge flow so that a fraction of the small and fine debris transport can be estimated based on the amount of water. → The break flow was assumed to be a homogeneous two-phase mixture without phase separation. Isenthalpic expansion of the break flow was used to specify the inlet boundary condition of the break flow. → The fraction of the small and fine debris transported to the upper part is 73%; this value is close to the value calculated using 1D lumped-parameter codes by the USNRC and the KINS, respectively, while 48% more than the value shown in the NEI 04-07. - Abstract: The performance of the emergency recirculation water sump under the influence of debris accumulation following a loss-of-coolant accident (LOCA) has long been of safety concern. Debris generation and transport during a LOCA are significantly influenced by the characteristics of the ejected coolant flow. One-dimensional analyses previously have been attempted to evaluate the debris transport during the blow-down phase but the transport evaluation still has large uncertainties. In this work, a computational fluid dynamics (CFD) analysis was utilized to evaluate small and fine debris transport during the blow-down phase of a pressurized water reactor, OPR1000. The coolant ejected from the ruptured hot-leg was assumed to expand in an isenthalpic process. The transport of small and fine debris was assumed to be dominated by water-borne transport, and the transport fractions for the upper and lower parts of the containment were quantified based on the CFD analysis. It was estimated that 73% of small and fine debris is transported to the upper part of the containment. This value is close to the values estimated by nuclear

  5. Fuel relocation effects in BWR LOCA conditions

    International Nuclear Information System (INIS)

    Raul Orive Moreno; Ines Gallego Cabezon; Pablo Julio Garcia Sedano; Yolanda Tofino Gomez; Pedro Mata Alonso

    2005-01-01

    One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MT rod average burnup. One of the licensing points of concern is the behavior of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/MT and H.B. Robinson at 67 GWd/MT. ANL LOCA tests indicate potential fuel relocation during LOCA. This could result in an increase of LHGR during a real plant LOCA. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations with GE-14 fuel design for the whole exposure range, quantifying fuel relocation impact. This effect has been modeled and implemented in FRAP-T6/APK (vendor independent IBERDROLA licensing thermomechanical code), as well as the wall-to-fluid heat transfer area increase in the ballooned region. Separate and combined impacts on PCT and ECR values can be evaluated with this modified code version. A new hoop strain versus rupture temperature curve is implemented in code, starting from NUREG-0630 model data base, but with a more best-estimate fit, in order to reproduce expected experimental values. The increase of heat transfer in the ballooned region has been validated with Halden LOCA tests. Preliminary results indicate that the effect of fuel relocation is expected to be compensated by the increased heat transfer area. This effect is to be confirmed with the Halden LOCA tests in progress. (author)

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

    Science.gov (United States)

    Aleixo, Rui; Altinakar, Mustafa

    2015-04-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Lipatov, I.A.; Dremin, G.I.

    2003-01-01

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

  9. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  10. Utilizing elements of the CSAU phenomena identification and ranking table (PIRT) to qualify a PWR non-LOCA transients system code

    Energy Technology Data Exchange (ETDEWEB)

    Greene, K.R.; Fletcher, C.D.; Gottula, R.C.; Lindquist, T.R.; Stitt, B.D. [Framatome ANP, Richland, WA (United States)

    2001-07-01

    Licensing analyses of Nuclear Regulatory Commission (NRC) Standard Review Plan (SRP) Chapter 15 non-LOCA transients are an important part of establishing operational safety limits and design limits for nuclear power plants. The applied codes and methods are generally qualified using traditional methods of benchmarking and assessment, sample problems, and demonstration of conservatism. Rigorous formal methods for developing code and methodology have been created and applied to qualify realistic methods for Large Break Loss-of-Coolant Accidents (LBLOCA's). This methodology, Code Scaling, Applicability, and Uncertainty (CSAU), is a very demanding, resource intensive, process to apply. It would be challenging to apply a comprehensive and complete CSAU level of analysis, individually, to each of the more than 30 non-LOCA transients that comprise Chapter 15 events. However, certain elements of the process can be easily adapted to improve quality of the codes and methods used to analyze non- LOCA transients. One of these elements is the Phenomena Identification and Ranking Table (PIRT). This paper presents the results of an informally constructed PIRT that applies to non-LOCA transients for Pressurized Water Reactors (PWR's) of the Westinghouse and Combustion Engineering design. A group of experts in thermal-hydraulics and safety analysis identified and ranked the phenomena. To begin the process, the PIRT was initially performed individually by each expert. Then through group interaction and discussion, a consensus was reached on both the significant phenomena and the appropriate ranking. The paper also discusses using the PIRT as an aid to qualify a 'conservative' system code and methodology. Once agreement was obtained on the phenomena and ranking, the table was divided into six functional groups, by nature of the transients, along the same lines as Chapter 15. Then, assessment and disposition of the significant phenomena was performed. The PIRT and

  11. LOCA and RIA studies at JAERI

    International Nuclear Information System (INIS)

    Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi

    2004-01-01

    To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI. (Author)

  12. Breaking gold nano-junctions simulation and analysis

    DEFF Research Database (Denmark)

    Lauritzen, Kasper Primdal

    Simulating the movements of individual atoms allows us to look at and investigate the physical processes that happen in an experiment. In this thesis I use simulations to support and improve experimental studies of breaking gold nano-junctions. By using molecular dynamics to study gold nanowires, I...... can investigate their breaking forces under varying conditions, like stretching rate or temperature. This resolves a confusion in the literature, where the breaking forces of two different breaking structures happen to coincide. The correlations between the rupture and reformation of a gold junction......, to predict the structure of a gold junction just as it breaks. This method is based on artificial neural networks and can be used on experimental data, even when it is trained purely on simulated data. The method is extended to other types of experimental traces, where it is trained without the use...

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

    International Nuclear Information System (INIS)

    Lu, S.; Streit, R.D.; Chou, C.K.

    1980-01-01

    This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10 -12 ). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

  14. Numerical modelling dam break analysis for water supply project

    International Nuclear Information System (INIS)

    Lariyah, M S; Vikneswaran, M; Hidayah, B; Muda, Z C; Thiruchelvam, S; Isham, A K Abd; Rohani, H

    2013-01-01

    Dam provides many benefits to the society, but it can also cause extensive damage to downstream area when it fails. Dam failure can cause extensive damage to properties and loss of human life due to short warning time available. In general, dam spillway was designed to drain the maximum discharge from the dam during the Probable Maximum Flood (PMF). The spillway is functioned to prevent the dam from failure due to overtopping, which can lead to the dam failure. Dam failure will result in large volume of water travelling at very high velocity to the downstream area of the dam. It can cause extensive property damage, destruction of important facilities, and significant loss of human life along the way. Due to the potential of high hazard it poses to the downstream area, a dam break analysis is considered very essential. This paper focuses into the dam failure analysis for Kahang Dam by prediction of breach flow hydrographs and generation of inundation map at downstream area. From the PMF scenario simulation, the maximum inflow is 525.12 m 3 /s and peak discharge from the dam during dam failure is 6188m 3 /s. The results are able to provide information for preparation of Emergency Response Plan (PMF), in which appropriate steps can be taken by relevant authorities to avoid significant loss of human lives.

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  16. Probabilistic leak-before-break analysis with correlated input parameters

    International Nuclear Information System (INIS)

    Qian Guian; Niffenegger, Markus; Karanki, Durga Rao; Li Shuxin

    2013-01-01

    Highlights: ► The correlation of crack growth has the most significant impact on LBB behavior. ► The correlation impact increases with the correlation coefficients. ► The correlation impact increases with the number of cracks. ► Independent assumption may lead to nonconservative result. - Abstract: The paper presents a probabilistic methodology considering the correlations between the input variables for the analysis of leak-before-break (LBB) behavior of a pressure tube. A computer program based on Monte Carlo (MC) simulation with Nataf transformation has been developed to allow the proposed methodology to calculate both the time from the first leakage to unstable fracture and the time from leakage detection to unstable fracture. The results show that the correlation of the crack growth rates between different cracks has the most significant impact on the LBB behavior of the pressure tube. The impact of the parameters correlation on LBB behavior increases with the crack numbers. If the correlations between different parameters for an individual crack are not considered, the predicted results are nonconservative when the cumulative probability is below 50% and conservative when it is above 50%.

  17. Application of KIMERA Methodology to Kori 3 and 4 LBLOCA M/E Release Analysis

    International Nuclear Information System (INIS)

    Song, Jeung Hyo; Hwang, Byung Heon; Kim, Cheol Woo

    2007-01-01

    A new mass and energy (M/E) release analysis methodology called KIMERA (KOPEC Improved Mass and Energy Release Analysis) has been developed. This is a realistic evaluation methodology of the M/E release analysis for the containment design and is applicable to a LOCA and a main steam line break (MSLB) accident. This KIMERA methodology has the same engine as KREM (KEPRI Realistic Evaluation Model) which is the realistic evaluation methodology for LOCA peak clad temperature analysis. This methodology also has several supplementary conservative models for the M/E release such as break spillage model and multiplier on heat transfer coefficient (HTC). For estimating the applicability of the KIMERA methodology to the licensing analysis, the large break LOCA (LBLOCA) M/E analysis was performed for UCN 3 and 4 which is the typical plant of OPR1000 type. The results showed that the peak pressure and temperature occurred earlier and had lower values than those of UCN 3 and 4 FSAR. The KIMERA methodology takes off the over-conservatism from the FSAR results during the post blowdown period for the large break LOCA and provides more margin in containment design. In this study, the LBLOCA M/E analysis using the KIMERA methodology is to be performed for Kori 3 and 4 which is the typical plant of Westinghouse type. The results are compared with those of the Kori Nuclear Unit 3 and 4 FSAR

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  20. Comparison of the electrophoretic method with the sedimentation method for the analysis of DNA strand breaks

    International Nuclear Information System (INIS)

    Yamamoto, Osamu; Ogawa, Masaaki; Hoshi, Masaharu

    1982-01-01

    Application of electrophoresis to the analysis of DNA strand breaks was studied comparing with the sedimentation analysis. A BRL gel electrophoresis system (Type V16) was used for this study. Calf thymus DNA (1 mg/ml) irradiated with 60 Co gamma-rays in SSC solution was applied to both the electrophoretic analysis and the sedimentation analysis. Lamda phage DNA and its fragments were employed as the standard size molecules. In a range from 1 k base pairs to 6 k base pairs in length for double stranded DNA or from 2 k bases to 12 k bases for single stranded DNA, the calculated average molecular weight from the electrophoresis coincided with that from the sedimentation. Number of single strand breaks and double strand breaks were 1.34 x 10 11 breaks/mg/rad (G = 0.215) and 0.48 x 10 5 breaks/mg/rad 2 , respectively. (author)

  1. Differential diagnosis of adductor spasmodic dysphonia and muscle tension dysphonia using phonatory break analysis.

    Science.gov (United States)

    Roy, Nelson; Whitchurch, Melissa; Merrill, Ray M; Houtz, Daniel; Smith, Marshall E

    2008-12-01

    Muscle tension dysphonia (MTD) can masquerade as adductor spasmodic dysphonia (ADSD) leading to diagnostic confusion. Intraword phonatory breaks have been offered as the sine qua non of ADSD, however, little is known regarding the presence of phonatory breaks in MTD. This investigation assessed the diagnostic worth of acoustic analysis of phonatory breaks as a possible objective test to distinguish ADSD from MTD. Case-control comparison. Voice samples from patients with confirmed ADSD (n = 41) and MTD (n = 59) were analyzed acoustically to determine the presence, frequency, and duration of phonatory breaks -- defined as complete interruption of phonation within a word. Estimates of sensitivity, specificity, positive and negative predictive value, and likelihood ratios were calculated to determine the precision and worth of phonatory break analysis as a clinical diagnostic test. 1) Individuals with ADSD showed a significantly higher number of phonatory breaks as compared with MTD. 2) All measures of diagnostic precision varied according to both duration and frequency of phonatory breaks, with separation of males and females leading to different diagnostic test performance results. The results suggest that phonatory break analysis offers promise as an objective test to distinguish ADSD from MTD, with respectable diagnostic precision, especially among men. Automation of the acoustic analysis procedure should be explored.

  2. LOCA assessment experiments in a full-elevation, CANDU-typical test facility

    International Nuclear Information System (INIS)

    Ingham, P.J.; McGee, G.R.; Krishnan, V.S.

    1990-01-01

    The RD-14 thermal-hydraulics test facility, located at the Whiteshell Nuclear Research Establishment, is a full-elevation model representative of a CANDU primary heat transport system. The facility is scaled to accommodate a single, full-scale (5.0 MW, 21 kg/s), electrically heated channel per pass. The steam generators, pumps, headers, feeders and heated channels are arranged in a typical CANDU figure-of-eight geometry. The loop has an emergency coolant injection system (ECI) that may be operated in several modes, including typical features of the various ECI systems found in CANDU reactors. A series of experiments has been performed in RD-14 to investigate the thermal-hydraulic behaviour during the blowdown and injection phases of a loss-of-coolant accident (LOCA). The tests were designed to cover a full range of break sizes from feeder-sized breaks to guillotine breaks in either an inlet or an outlet header. Breaks resulting in channel flow stagnation were also investigated. This paper reviews the results of some of the LOCA tests carried out in RD-14, and discusses some of the behaviour observed. Plans for future experiments in a multiple-channel RD-14 facility, modified to contain multiple flow channels, are outlined. (orig.)

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

    International Nuclear Information System (INIS)

    Falotico, R.N.; Peddicord, K.L.; Oehlberg, R.N.

    1978-01-01

    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)

  4. Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

    International Nuclear Information System (INIS)

    Kolev, N.I.; Sybotinov, L.S.

    1984-01-01

    The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

  5. The development of new analysis procedures for reactor internals under pipe breaks

    International Nuclear Information System (INIS)

    Song, Heuy Gap; Jhung, Myung Jo; Chang, Sang Gyun; Lee, Gyu Man

    1993-04-01

    This study investigates the horizontal responses of the reactor internals due to a 14 inch safety injection nozzle break which is expected to cause the largest loads of the branch line pipe breaks defined for the YGN 3 and 4. It examines the effects of two forcing terms, RV motions and internals hydraulic loads, and suggests new procedure which can be used for the tributary pipe break analysis. The analysis result confirms the applicability of suggested procedure to a small size tributary pipe break analysis. Also, this study calculates the horizontal responses of the reactor internals due to a 3 inch pressurizer spray line nozzle break which is the only one remaining in the primary side after leak-before-break evaluation, and secondary side pipe breaks such as main steam line and economizer feedwater line. The responses are compared with those of safe shutdown earthquake(SSE) to show that SSE loads with a conservative margin may be used for the pipe break loads in the preliminary design. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  7. An analysis methodology for hot leg break mass and energy release

    International Nuclear Information System (INIS)

    Song, Jin Ho; Kwon, Young Min; Kim, Taek Mo; Chung, Hae Yong; Lee, Sang Jong

    1996-07-01

    An analysis methodology for the hot leg break mass and energy release is developed. For the blowdown period a modified CEFLASH-4A analysis is suggested. For the post-blowdown period a new computer model named COMET is developed. Differently from previous post-blowdown analysis model FLOOD3, COMET is capable of analyzing both cold leg and hot leg break cases. The cold leg break model is essentially same as that of FLOOD3 with some improvements. The analysis results by the newly proposed hot leg break model in the COMET is in the same trend as those observed in scaled-down integral experiment. And the analyses results for the UCN 3 and 4 by COMET are qualitatively and quantitatively in good agreement with those predicted by best-estimate analysis by using RELAP5/MOD3. Therefore, the COMET code is validated and can be used for the licensing analysis. 6 tabs., 82 figs., 9 refs. (Author)

  8. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  9. Chiral symmetry breaking in gauge theories from Reggeon diagram analysis

    International Nuclear Information System (INIS)

    White, A.R.

    1991-01-01

    It is argued that reggeon diagrams can be used to study dynamical properties of gauge theories containing a large number of massless fermions. SU(2) gauge theory is studied in detail and it is argued that there is a high energy solution which is analogous to the solution of the massless Schwinger model. A generalized winding-number condensate produces the massless pseudoscalar spectrum associated with chiral symmetry breaking and a ''trivial'' S-Matrix

  10. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  11. Core heatup prediction during SB LOCA with RELAP5/MOD3.2.2 Gamma

    International Nuclear Information System (INIS)

    Parzer, I.; Mavko, B.; Petelin, S.

    2001-01-01

    The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and the ability of RELAP5/MOD3.2.2 Gamma to predict core overheating. The code prediction has been compared to the three experiments, one conducted on the separate effect test facility NEPTUN in Switzerland and the other two conducted on two integral test facilities, PMK-2 in Hungary and PACTEL facility in Finland. In the case of a series of boiloff experiments performed on the NEPTUN test facility the influence of the two correlations available in MOD3.2.2 Gamma for determining interphase drag has been studied. In the case of IAEA-SPE-4 experiment simulation on PMK-2 facility the main goal of the analysis was to study the adequate modeling of the hexagonal core channel with 19-rod bundle and the phenomena during the core uncovering. The third analyzed experiment, OECD-ISP-33, was performed on PACTEL facility to study different natural circulation modes during SB LOCA. The analysis also focused on the final stage of this SB LOCA experiment, when core dryout and heatup was observed due to gradual emptying of the primary system. Following the experience the appropriate modeling options have been used to achieve better representation of the important phenomena during the SB LOCA.(author)

  12. Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design

    International Nuclear Information System (INIS)

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

    1994-01-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations

  13. Advanced Methodology for Containment M/E Release Analysis

    International Nuclear Information System (INIS)

    Kim, C. W.; Park, S. J.; Song, J. H.; Choi, H. R.; Seo, J. T.

    2006-01-01

    Recently, a new mass and energy (M/E) release analysis methodology for the equipment environmental qualification (EEQ) on loss-of-coolant accident (LOCA) has been developed and adopted on small break LOCA (SBLOCA). This new M/E release analysis methodology for EEQ is extended to the M/E release analysis for the containment design for large break LOCA (LBLOCA) and main steam line break (MSLB) accident. The advanced methodology of the M/E release analysis for the containment design includes the same engine as the M/E methodology for EEQ, however, conservative approaches for the M/E release such as break spillage model and multiplier on heat transfer coefficient (HTC) etc. are added. The computer code systems used in this methodology are RELAP5K/CONTEMPT4 (or RELAP5- ME) like KREM (KEPRI Realistic Evaluation Model) which couples RELAP5/MOD3.1/K and CONTEMPT4/ MOD5. RELAP5K is based on RELAP5/MOD3.1/K and includes conservatisms for the M/E release and long-term analysis model. The advanced methodology adopting the recent analysis technology is able to calculate the various transient stages of a LOCA in a single code system and also can calculate the M/E release analysis during the long term cooling period with the containment response. This advanced methodology for the M/E release is developed based on the LOCA and applied to the MSLB. The results are compared with the Ulchin Nuclear Unit (UCN) 3 and 4 FSAR

  14. Development of mass and energy release analysis methodology

    International Nuclear Information System (INIS)

    Kim, Cheol Woo; Song, Jeung Hyo; Park, Seok Jeong; Kim, Tech Mo; Han, Kee Soo; Choi, Han Rim

    2009-01-01

    Recently, new approaches to the accident analysis using the realistic evaluation have been attempted. These new approaches provide more margins to the plant safety, design, operation and maintenance. KREM (KEPRI Realistic Evaluation Methodology) for a large break loss-of-coolant accident (LOCA) is performed using RELAP5/MOD3 computer code including realistic evaluation models. KOPEC has developed KIMERA (KOPEC Improved Mass and Energy Release Analysis methodology) based on the realistic evaluation to improve the analysis method for the mass and energy (M/E) release and to obtain the adequate margin. KIMERA uses a simplified single code system unlike conventional M/E release analysis methodologies. This simple code system reduces the computing efforts especially for LOCA analysis. The computer code systems of this methodology are RELAP5K/CONTEMPT4 (or RELAP5-ME) like KREM methodology which couples RELAP5/MOD3.1/K and CONTEMPT4/MOD5. The new methodology, KIMERA based on the same engine as KREM, adopted conservative approaches for the M/E release such as break spillage model, multiplier on heat transfer coefficient (HTC), and long-term cooling model. KIMERA is developed based on a LOCA and applied to a main steam line break (MSLB) and approved by Korea Government. KIMERA has an ability of one-through calculation of the various transient stages of LOCAs in a single code system and calculate the M/E release analysis during the long term cooling period with the containment pressure and temperature (P/T) response. The containment P/T analysis results are compared with those of the Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) FSAR which is the OPR1000 (Optimized Power Reactor 1000) type nuclear power plant. The results of a large break LOCA and an MSLB are similar to those of FSAR for UCN 3 and 4. However, the containment pressure during the post-blowdown period of a large break LOCA has much lower second peak than the first peak. The resultant containment peak

  15. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    International Nuclear Information System (INIS)

    Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

    2015-01-01

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  16. The use of a break-even analysis: financial analysis of a fast-track program.

    Science.gov (United States)

    Saywell, R M; Cordell, W H; Nyhuis, A W; Giles, B K; Culler, S D; Woods, J R; Chu, D K; McKinzie, J P; Rodman, G H

    1995-08-01

    To calculate the financial break-even point and illustrate how changes in third-party reimbursement and eligibility could affect a program's fiscal standing. Demographic, clinical, and financial data were collected retrospectively for 446 patients treated in a fast-track program during June 1993. The fast-track program is located within the confines of the emergency medicine and trauma center at a 1,050-bed tertiary care Midwestern teaching hospital and provides urgent treatment to minimally ill patients. A financial break-even analysis was performed to determine the point where the program generated enough revenue to cover its total variable and fixed costs, both direct and indirect. Given the relatively low average collection rate (62%) and high percentage of uninsured patients (31%), the analysis showed that the program's revenues covered its direct costs but not all of the indirect costs. Examining collection rates or payer class mix without examining both costs and revenues may lead to an erroneous conclusion about a program's fiscal viability. Sensitivity analysis also shows that relatively small changes in third-party coverage or eligibility (income) requirements can have a large impact on the program's financial solvency and break-even volumes.

  17. Identification of Error of Commissions in the LOCA Using the CESA Method

    Energy Technology Data Exchange (ETDEWEB)

    Tukhbyet-olla, Myeruyert; Kang, Sunkoo; Kim, Jonghyun [KEPCO international nuclear graduate school, Ulsan (Korea, Republic of)

    2015-10-15

    An Errors of commission (EOCs) can be defined as the performance of any inappropriate action that aggravates the situation. The primary focus in current PSA is placed on those sequences of hardware failures and/or EOOs that lead to unsafe system states. Although EOCs can be treated when identified, a systematic and comprehensive treatment of EOC opportunities remains outside the scope of PSAs. However, some past experiences in the nuclear industry show that EOCs have contributed to severe accidents. Some recent and emerging human reliability analysis (HRA) methods suggest approaches to identify and quantify EOCs, such as ATHEANA, MERMOS, GRS, MDTA, and CESA. The CESA method, developed by the Risk and Human Reliability Group at the Paul Scherrer Institute, is to identify potentially risk-significant EOCs, given an existing PSA. The main idea underlying the method is to catalog the key actions that are required in the procedural response to plant events and to identify specific scenarios in which these candidate actions could erroneously appear to be required. This paper aims at identifying EOCs in the LOCA by using the CESA method. This study is focused on the identification of EOCs, while the quantification of EOCs is out of scope. Then, this paper applies the CESA method to the emergency operating procedure (EOP) of LOCA for APR1400. Finally, this study presents potential EOCs that may lead to the aggravation in the mitigation of LOCA. This study has identified the EOC events for APR1400 in the LOCA using CESA method. The result identified three candidate EOCs event using operator action catalog and RAW cutset of LOCA. These candidate EOC events are inappropriate terminations of safety injection system, safety injection tank and containment spray system. Then after reviewing top 100 accident sequences of PSA, this study finally identified one EOC scenario and EOC path, that is, inappropriate termination of safety injection system.

  18. 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)

    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

  19. REFLA-1D/MODE 1: a computer program for reflood thermo-hydrodynamic analysis during PWR-LOCA user's manual

    International Nuclear Information System (INIS)

    Murao, Yoshio; Sugimoto, Jun; Okubo, Tsutomu

    1981-01-01

    This manual describes the REFLA-1D/MODE 1 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/MODE 1 models and provides application information required to utilize REFLA-1D/MODE 1. (author)

  20. 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.

  1. Estimation of Leak Flow Rate during Post-LOCA Using Cascaded Fuzzy Neural Networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    In this study, important parameters such as the break position, size, and leak flow rate of loss of coolant accidents (LOCAs), provide operators with essential information for recovering the cooling capability of the nuclear reactor core, for preventing the reactor core from melting down, and for managing severe accidents effectively. Leak flow rate should consist of break size, differential pressure, temperature, and so on (where differential pressure means difference between internal and external reactor vessel pressure). 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 paper, a cascaded fuzzy neural network (CFNN) model is appropriately proposed to estimate 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.). The CFNN is a data-based model, it requires data to develop and verify itself. Because few actual severe accident data exist, it is essential to obtain the data required in the proposed model using numerical simulations. In this study, a CFNN model was developed to predict the leak flow rate before proceeding to severe LOCAs. The simulations showed that the developed CFNN model accurately predicted the leak flow rate with less error than 0.5%. The CFNN model is much better than FNN model under the same conditions, such as the same fuzzy rules. At the result of comparison, the RMS errors of the CFNN model were reduced by approximately 82 ~ 97% of those of the FNN model.

  2. Fluid-structure coupled dynamic response of PWR core barrel during LOCA

    International Nuclear Information System (INIS)

    Lu, M.W.; Zhang, Y.G.; Shi, F.

    1991-01-01

    This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

  3. LOCA consequence predictions in a CANDU-PHWR

    International Nuclear Information System (INIS)

    Meneley, D.A.; Hancox, W.T.

    1982-09-01

    The paper represents consequence predictions for LOCA sequences in a typical CANDU station. Starting from defined basic LOCA sequences, the importance of each engineered system is tested by assuming failure to perform its function on demand. Consequences are calculated for each failure combination. Experimental results are presented to support predictions. The overall conclusion is that public consequences are very small for LOCA sequences more probable than 10 - 7 per year. The moderator system, in assuring that no fuel can melt even if emergency coolant injection fails, is an important contributor to this very high level of public protection

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

    International Nuclear Information System (INIS)

    Brittain, I.; Bryce, W.M.; O'Mahoney, R.; Richards, C.G.; Gibson, I.H.; Porter, W.H.L.; Fell, J.

    1984-03-01

    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.)

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

    Science.gov (United States)

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

    2017-11-01

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

  6. 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...

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

    International Nuclear Information System (INIS)

    Muramatsu, Ken; Akimoto, Masayuki

    1982-08-01

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

  8. Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su [Korea Atomic Energy Research Institute, Yusung, Taejon (Korea, Republic of)] [and others

    1996-10-01

    This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties.

  9. Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su

    1996-10-01

    This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties

  10. Simulated LOCA Test and Characterization Study Related to High Burn-Up Issue

    International Nuclear Information System (INIS)

    Park, D. J.; Jung, Y. I.; Choi, B. K.; Park, S. Y.; Kim, H. G.; Park, J. Y.

    2012-01-01

    For the safety evaluation of fuel cladding during the injection of emergency core coolant, simulated Loss-of-coolant accident (LOCA) test was performed by using Zircaloy-4 fuel cladding samples. Zircaloy-4 tube samples with and without prehydring were oxidized in a steam environment with the test temperature of 1200 .deg. C. Prehydrided cladding was prepared from as-fabricated Zircaloy-4 to study the effects of hydrogen on mechanical properties of cladding during high temperature oxidation and quench conditions. In order to measure the ductility of the tube samples embrittled by quenching water, ring compression test was carried out by using 8 mm ring sample sectioned from oxidized tube sample and microstructural analysis was also performed after simulated LOCA test. The results showed that hydrogen increases oxygen solubility and pickup rate in the beta layer. This reduces ductility of prehydrided fuel cladding compared with as-fabricated cladding. Trend in ductility decrease for prehydrided sample under simulated LOCA condition was very similar with data obtained from tests conducted using irradiated high burn-up fuel claddings

  11. Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

    This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

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

    Science.gov (United States)

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

    2016-04-25

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

  13. Veracity and velocity of social media content during breaking news: analysis of November 2015 Paris shootings

    OpenAIRE

    Middleton, Stuart

    2016-01-01

    Social media sources are becoming increasingly important in journalism. Under breaking news deadlines semi-automated support for identification and verification of content is critical. We describe a large scale content-level analysis of over 6 million Twitter, You Tube and Instagram records covering the first 6 hours of the November 2015 Paris shootings. We ground our analysis by tracing how 5 ground truth images used in actual news reports went viral. We look at velocity of newsworthy conten...

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

    International Nuclear Information System (INIS)

    Lee, Sang Il

    1992-02-01

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

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

    International Nuclear Information System (INIS)

    Lee, Sang II; No, Hee Cheon

    1992-01-01

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

  16. High-Throughput Analysis of DNA Break-Induced Chromosome Rearrangements by Amplicon Sequencing.

    Science.gov (United States)

    Brown, Alexander J; Al-Soodani, Aneesa T; Saul, Miles; Her, Stephanie; Garcia, Juan C; Ramsden, Dale A; Her, Chengtao; Roberts, Steven A

    2018-01-01

    The mechanistic understanding of how DNA double-strand breaks (DSB) are repaired is rapidly advancing in part due to the advent of inducible site-specific break model systems as well as the employment of next-generation sequencing (NGS) technologies to sequence repair junctions at high depth. Unfortunately, the sheer volume of data produced by these methods makes it difficult to analyze the structure of repair junctions manually or with other general-purpose software. Here, we describe methods to produce amplicon libraries of DSB repair junctions for sequencing, to map the sequencing reads, and then to use a robust, custom python script, Hi-FiBR, to analyze the sequence structure of mapped reads. The Hi-FiBR analysis processes large data sets quickly and provides information such as number and type of repair events, size of deletion, size of insertion and inserted sequence, microhomology usage, and whether mismatches are due to sequencing error or biological effect. The analysis also corrects for common alignment errors generated by sequencing read mapping tools, allowing high-throughput analysis of DSB break repair fidelity to be accurately conducted regardless of which suite of NGS analysis software is available. © 2018 Elsevier Inc. All rights reserved.

  17. Thermal Analysis of Brazing Seal and Sterilizing Technique to Break Contamination Chain for Mars Sample Return

    Science.gov (United States)

    Bao, Xiaoqi; Badescu, Mircea; Bar-Cohen, Yoseph

    2015-01-01

    The potential to return Martian samples to Earth for extensive analysis is in great interest of the planetary science community. It is important to make sure the mission would securely contain any microbes that may possibly exist on Mars so that they would not be able to cause any adverse effects on Earth's environment. A brazing sealing and sterilizing technique has been proposed to break the Mars-to-Earth contamination chain. Thermal analysis of the brazing process was conducted for several conceptual designs that apply the technique. Control of the increase of the temperature of the Martian samples is a challenge. The temperature profiles of the Martian samples being sealed in the container were predicted by finite element thermal models. The results show that the sealing and sterilization process can be controlled such that the samples' temperature is maintained below the potentially required level, and that the brazing technique is a feasible approach to break the contamination chain.

  18. Transient deformation properties of Zircaloy for LOCA simulation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hann, C.R.; Mohr, C.L.; Busness, K.M.; Olson, N.J.; Reich, F.R.; Stewart, K.B.

    1978-03-01

    The creep/creep rupture anisotropic properties of Zircaloy were determined and compared by analytical techniques with ramp-pressure and ramp-temperature test results. Tests were performed over the temperature range of 600/sup 0/F (589/sup 0/K) to 2200/sup 0/F (1477/sup 0/K) with the emphasis on the 800/sup 0/F (700/sup 0/K) to 2000/sup 0/F (1366/sup 0/K) temperature levels in low pressure air (6.5 x 10/sup -5/ atm) and in a 1 atm mixture of 20% oxygen, 80% argon. Stress levels of 60 to 95% of the ultimate tensile stress were used for the majority of the tests at each of the temperature levels tested, with selected tests performed as low as 30% of the ultimate tensile stress. Biaxial and uniaxial testing modes were used to evaluate the anisotropic deformation behavior. The combination of test results and predictive analysis techniques developed as part of this program make it possible to predict the transient deformation of reactor fuel cladding during simulated LOCA conditions. Results include creep/creep rupture strain numerical constitutive relationships out of 120 seconds, computer codes and ramp test data.

  19. Transient deformation properties of Zircaloy for LOCA simulation. Final report

    International Nuclear Information System (INIS)

    Hann, C.R.; Mohr, C.L.; Busness, K.M.; Olson, N.J.; Reich, F.R.; Stewart, K.B.

    1978-03-01

    The creep/creep rupture anisotropic properties of Zircaloy were determined and compared by analytical techniques with ramp-pressure and ramp-temperature test results. Tests were performed over the temperature range of 600 0 F (589 0 K) to 2200 0 F (1477 0 K) with the emphasis on the 800 0 F (700 0 K) to 2000 0 F (1366 0 K) temperature levels in low pressure air (6.5 x 10 -5 atm) and in a 1 atm mixture of 20% oxygen, 80% argon. Stress levels of 60 to 95% of the ultimate tensile stress were used for the majority of the tests at each of the temperature levels tested, with selected tests performed as low as 30% of the ultimate tensile stress. Biaxial and uniaxial testing modes were used to evaluate the anisotropic deformation behavior. The combination of test results and predictive analysis techniques developed as part of this program make it possible to predict the transient deformation of reactor fuel cladding during simulated LOCA conditions. Results include creep/creep rupture strain numerical constitutive relationships out of 120 seconds, computer codes and ramp test data

  20. MELCOR ex-vessel LOCA simulations for ITER+

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

    1995-01-01

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

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

    International Nuclear Information System (INIS)

    Kim, Young In; Hwang, Young Dong; Park, Jong Kuen; Chung, Young Jong; Sim, Suk Gu

    1999-02-01

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

  2. 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.

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

    International Nuclear Information System (INIS)

    Bogatyr, S. M.; Krupkin, A. V.; Kuznetsov, V. I.; Novikov, V. V.; Petrov, O. M.; Shestopalov, A. A.

    2013-01-01

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

  4. Studies in Phebus reactor of fuel behaviour upon LOCA conditions

    International Nuclear Information System (INIS)

    Manin, A.; Del Negro, R.; Reocreux, M.

    1980-09-01

    The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

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

    International Nuclear Information System (INIS)

    Vrtilkova, V.; Novotny, L.; Hamouz, V.; Doucha, R.; Tinka, I.; Macek, J.; Lahovsky, F.

    2005-01-01

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

  6. Establishment of the operating procedure to prevent boron precipitation during Post-LOCA long term cooling for Korean Westinghouse 3-loop NPPs

    International Nuclear Information System (INIS)

    Choi, Han Rim; Kwon, Tae Soon; Ban, Chang Hwan; Jeong, Jae Hoon; Lee, Young Jin.

    1996-11-01

    During post-LOCA LTC the increase of the excess reactivity for the extended fuel cycle should require increasing the RWST boron concentration in order to ensure core subcritical state. To quantify the concentration increment, the calculation methods was developed for the post-LOCA RCS/Sump mixed mean boron concentration, which applied for Kori 3 and 4 and Ulchin 1 and 2 of the Westinghouse 3-loop nuclear power plants in Korean. From the calculation results, the minimum boric acid concentrations increased of the RWST and accumulator were determined consideration of the convenient operation for operator on reloading. Boric acid concentrations of the RWST and the accumulators for Westinghouse 3-loop type plants were increased to meet the post-LOCA shutdown requirement for the long life cycles from 12 months to 18 months. To maintain LTC capability following a LOCA, the switchover time is examined using boron code of prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results showed that hot leg recirculation switchover times were shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3 and 4 and 8 hours from 18 hours for Ulchin 1 and 2, respectively. The flow path in the mode J for Kori 3 and 4 was recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1 and 2. (author). 2 tabs., 12 figs., 13 refs

  7. Leakage rate from LOCA-aged inflatable airlock seals Pickering NGS 'B' personnel doors

    International Nuclear Information System (INIS)

    Fayle, G.W.; Cordingley, D.C.

    1985-01-01

    In order to demonstrate to the Atomic Energy Control Board that an air-lock inflatable seal will function after a LOCA exposure, an inflatable seal intended for personnel doors at the Pickering NGS 'B' was exposed to the thermal/moisture conditions of the LOCA requirement. While attending to determine the post-LOCA leakage rate it was found that additional leaks developed during each post-LOCA inflation/deflation cycle. The seal had been significantly and irreparably deteriorated by the LOCA exposure. The test has demonstrated that this type of LOCA exposed seal should not be expected to withstand either additional pressure above 207 kPa or additional inflation/deflation cycling. A higher inflation pressure and/or cycling will reduce the likelihood of a post-LOCA seal retaining an inflation pressure sufficient to prevent leakage across the seal

  8. 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.

  9. Reflooding phase of the LOCA - state of the art II. Rewetting and liquid entrainment

    International Nuclear Information System (INIS)

    Elias, E.; Yadigaroglu, G.

    1977-01-01

    Understanding the mechanisms by which hot fuel rods quench and the physics of liquid droplet entrainment is important for the analysis of the reflooding phase of the LOCA. Published models of the rewetting process include simple one-dimensional solutions. The basic physical assumptions of these models and the numerical values assigned to the various parameters, as well as empirical rewetting correlations are discussed. The various mechanisms for liquid droplet entrainment and analytical formulations of the critical gas velocity and of the droplet diameter at the onset of entrainment are reviewed

  10. PWR small-break analysis using a PDP-11/AD10 computer system

    International Nuclear Information System (INIS)

    Venhuizen, J.R.; Hyer, F.K.

    1983-01-01

    A simulation of a pressurized water test reactor was developed to predict the dynamic response of the primary coolant system to gradual voiding caused by an anticipated transient or a small break. Comparison of the simulation results with data from the LOFT test reactor at the Idaho National Engineering Laboratory was performed to verify the models. The simulation, designed to operate on a PDP-11/55 minicomputer and Applied Dynamic AD10 synchronous digital computer, was used interactively to do scoping analysis prior to running the transient at the test reactor

  11. 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.

  12. 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.

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

    International Nuclear Information System (INIS)

    1980-01-01

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

  14. Fuel Behaviour at High During RIA and LOCA Accidents

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Analysis of a gas stratification break-up by a vertical jet using the GOTHIC code

    International Nuclear Information System (INIS)

    Fernández-Cosials, Mikel Kevin; Jimenez, Gonzalo; Lopez-Alonso, Emma

    2016-01-01

    Highlights: • Study of a light gas distribution with the GOTHIC code based on the OECD/NEA IBE-3. • Sensitivity analysis on turbulence model, discretization scheme and heat transfer. • The jet erosion phenomena is captured properly with a relatively coarse mesh. • Development of a tool to evaluate the influence of each parameter on the simulation. • Several recommendation on modeling a stratification break-up are included. - Abstract: During a severe accident in light water reactor (LWR), hydrogen concentration can overpass the flammability limits locally, so the correct simulation of its behavior during a release is critical. The capability assessment of computational fluid dynamics tools to calculate the hydrogen distribution under different conditions has been the focus of intense research worldwide. In this context, the OECD/NEA conducted an international benchmark exercise (IBE-3), which was focused on the break-up of a stratified layer of a light gas by a vertical jet. The participants performed their simulations before the experiment data was released. When the data was released, it was noticed that a combination of several parameters like the mesh, turbulence model or solver controls were responsible for the broad differences between the participants’ results. To obtain information about how each parameter affects the simulation, a post-test sensitivity analysis has been done by the UPM. In this paper, the IBE-3 experiment simulation with GOTHIC 8.0 is presented along with extensive sensitivity analyses of the relevant parameters. The first objective of the work is to test the capability of GOTHIC 8.0 to simulate properly a gas stratification break-up by a vertical jet with a relatively coarse mesh. The second objective of the paper is to relate each sensitivity parameter with each other and with the experiment through the Parameter Influence Chart, a helpful tool specially designed for this purpose.

  16. Analysis of a gas stratification break-up by a vertical jet using the GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Fernández-Cosials, Mikel Kevin; Jimenez, Gonzalo, E-mail: gonzalo.jimenez@upm.es; Lopez-Alonso, Emma

    2016-02-15

    Highlights: • Study of a light gas distribution with the GOTHIC code based on the OECD/NEA IBE-3. • Sensitivity analysis on turbulence model, discretization scheme and heat transfer. • The jet erosion phenomena is captured properly with a relatively coarse mesh. • Development of a tool to evaluate the influence of each parameter on the simulation. • Several recommendation on modeling a stratification break-up are included. - Abstract: During a severe accident in light water reactor (LWR), hydrogen concentration can overpass the flammability limits locally, so the correct simulation of its behavior during a release is critical. The capability assessment of computational fluid dynamics tools to calculate the hydrogen distribution under different conditions has been the focus of intense research worldwide. In this context, the OECD/NEA conducted an international benchmark exercise (IBE-3), which was focused on the break-up of a stratified layer of a light gas by a vertical jet. The participants performed their simulations before the experiment data was released. When the data was released, it was noticed that a combination of several parameters like the mesh, turbulence model or solver controls were responsible for the broad differences between the participants’ results. To obtain information about how each parameter affects the simulation, a post-test sensitivity analysis has been done by the UPM. In this paper, the IBE-3 experiment simulation with GOTHIC 8.0 is presented along with extensive sensitivity analyses of the relevant parameters. The first objective of the work is to test the capability of GOTHIC 8.0 to simulate properly a gas stratification break-up by a vertical jet with a relatively coarse mesh. The second objective of the paper is to relate each sensitivity parameter with each other and with the experiment through the Parameter Influence Chart, a helpful tool specially designed for this purpose.

  17. Consistency of Trend Break Point Estimator with Underspecified Break Number

    Directory of Open Access Journals (Sweden)

    Jingjing Yang

    2017-01-01

    Full Text Available This paper discusses the consistency of trend break point estimators when the number of breaks is underspecified. The consistency of break point estimators in a simple location model with level shifts has been well documented by researchers under various settings, including extensions such as allowing a time trend in the model. Despite the consistency of break point estimators of level shifts, there are few papers on the consistency of trend shift break point estimators in the presence of an underspecified break number. The simulation study and asymptotic analysis in this paper show that the trend shift break point estimator does not converge to the true break points when the break number is underspecified. In the case of two trend shifts, the inconsistency problem worsens if the magnitudes of the breaks are similar and the breaks are either both positive or both negative. The limiting distribution for the trend break point estimator is developed and closely approximates the finite sample performance.

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

    Science.gov (United States)

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

    2014-01-01

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

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

    Science.gov (United States)

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

    2014-01-01

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

  20. Development of supporting software for safety analysis simulator for nuclear reactor

    International Nuclear Information System (INIS)

    Li Tonglin; Yao Qingsheng; Han Weishi

    2008-01-01

    An investigation of nuclear reactor simulator was made on its physics model, thermal hydraulics model, modeling method for assistant system and numerical calculation technology. The simultaneous equations for each module and node of the steam supply system are solved by full implicit difference method. Then the supporting and calculation software for simulation was developed based on Windows system. Calculation and comparison have been done for small break LOCA by the safety analysis simulator developed in this paper and Relap5 code. The results show that the variation of primary parameters for break LOCA of this model accord with the calculation results from Relap5, and the simulator can perform real time calculation with well stability. (authors)

  1. 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.

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

    International Nuclear Information System (INIS)

    Hoffer, Nathan V.; Anderson, Nolan A.; Sabharwall, Piyush

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  4. 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

  5. Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

    International Nuclear Information System (INIS)

    Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C.

    2012-01-01

    The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

  6. Transient deformation properties of Zircaloy for LOCA simulation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hann, C. R.; Mohr, C. L.; Busness, K. M.; Olson, N. J.; Reich, F. R.; Stewart, K. B.

    1980-05-01

    This experimental data report is Volume 4 of a series of 5 volumes describing the oxidation and deformation rate behavior of Zircaloy cladding under simulated LOCA conditions. It contains listings of strain versus stress, time, and temperature evaluated from the numerical constitutive relationships and the original data used to develop them. This volume also contains listings of the ramp load, pressure, and temperature test data from both current and previous phases of the series, as well as material describing applications of the data.

  7. Experimental study of plant specific head loss induced by LOCA-generated debris at containment sump of PWR

    International Nuclear Information System (INIS)

    Young Wook, Chung; Young Mook, Hwang; Jong Uk, Kim; Byung Gi, Park; Byung Chul, Lee; Jong Woon, Park

    2007-01-01

    A LOCA in PWR would generate debris from thermal insulation and other materials in the vicinity of the break. A fraction of the LOCA-generated debris and pre-LOCA debris will be transported into the sump and accumulated on the sump screens resulting in adverse blockage effects that are degradation or loss of NPSH (Net Positive Suction Head) margin. To assess debris-induced head loss in the sump screen, experimental studies have been widely conducted and the results have exhibited that head loss depends on amount of debris, specific surface area, mixture porosity of debris bed, debris types, and so on. Based on the experimental results, empirical correlations have been developed. NUREG/CR-6224 head loss correlation among them has been widely used to estimate the debris-induced head loss for PWR sump performance evaluation. However, in order to apply this correlation for estimating head loss in specific PWR plant, plant-specific head loss data are required because of a different composition of debris sources between PWRs. Plant specific head loss data were obtained with a test facility that is a closed-loop types. A vertical test section of test facility was fabricated with 6 inch CPVC (Chlorinated Polyvinyl Chloride) pipe. A ratio of length to diameter at the vertical test section was about 30. Experimental results exhibited that the head loss across NUKON debris bed with theoretical thickness greater than 4 inch was predicted conservatively by NUREG/CR-6224 correlation. Head loss test with debris composition of Westinghouse two loop plant showed that NUREG/CR-6224 correlation predicted higher head loss than experimentally measured head loss. (authors)

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

    Science.gov (United States)

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

    2013-04-01

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

  9. The Source of Bacteria Involved in the Break-Down of Gammarus Pulex Faecal Pellets Using Image Analysis

    Science.gov (United States)

    Joyce, P.; Wotton, R.

    2005-05-01

    Bacteria survive passage through the gut of aquatic animals and are implicated in the break-down of POM (such as faecal pellets) in aquatic systems. There is evidence that bacteria that survive gut passage are the cause of the initial break-down of faecal pellets, rather than colonisation by bacteria from the environment. Gammarus is the dominant shredder in lowland permeable catchments ("chalk streams") in England and feeds on allochthonous detritus such as fallen leaves. Gammarus faecal pellets could form important pathways for transfer of organic matter in chalk streams. We incubated Gammarus faecal pellets for 80 days in stream water using combinations of treatments (autoclaving the stream water or pellets; fresh non-autoclaved stream water or pellets; reducing bacterial activity using Gentamicin; combinations of these treatments) to assess the role of surviving and colonising bacteria on the break-down process. Break-down was measured using image analysis. Results show that treatments with fresh pellets show much higher levels of break-down than fully autoclaved controls, treatments with fresh stream water but autoclaved pellets, or treatments with Gentamicin. Bacteria surviving gut passage therefore seem to play a greater role in faecal pellet break-down than those colonising from the environment.

  10. Refined analysis of effects on the LOFT containment building resulting from main steam and feedwater breaks

    Energy Technology Data Exchange (ETDEWEB)

    Mosby, W.R.

    1979-01-16

    Dynamic one-degree-of-freedom analyses of main steam breaks 8/10B and 8/11 and feedwater break 9/40 were performed using refined jet-plus-reaction force-time functions. Feedwater breaks 10/8B and 10/11 were analyzed statically using a refined maximum jet-plus-reaction force and assuming a dynamic load factor of 2. None of the breaks were found to stress the containment shell in excess of the ASME Boiler and Pressure Vessel Code allowables.

  11. 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…

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

    International Nuclear Information System (INIS)

    Nguyen, T.H.; Collins, W.M.

    1995-01-01

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

  13. Sensitivity of break-flow-partition on the containment pressure and temperature

    International Nuclear Information System (INIS)

    Kwon, Young Min; Song, Jin Ho; Lee, Sang Yong

    1994-01-01

    For the case of RCS blowdown into the vapor region of a containment at low pressure, the blowdown mixture will start to boil at the containment pressure and liquid will separate from the flow near the break location. The flashed steam is added to the containment atmosphere and liquid is falled to the sump. Analytically, the break flow can be divided into steam and liquid in a number of ways. Discussed in this study is three partition models and Instantaneous Mixing(IM) Model employed in different containment analysis computer codes. IM model is employed in the CONTRANS code developed by ABB-CE for containment thermodynamic analysis. The various partition models were applied to the double ended discharge leg slot break (DEDLS) LOCA which is containment design base accident (CDBA) for Ulchin 3 and 4 PSAR. It was shown that IM model is the most conservative for containment design pressure analysis. Results of the CONTRANS analyses are compared with those of UCN PSAR for which CONTEMPT-LT code was used

  14. Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI Line Break

    Directory of Open Access Journals (Sweden)

    A.S. Ekariansyah

    2016-08-01

    Full Text Available Generation II Nuclear Power Plants (NPPs have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis.

  15. An Evaluation of Practical Applicability of Multi-Assortment Production Break-Even Analysis based on Mining Companies

    Science.gov (United States)

    Fuksa, Dariusz; Trzaskuś-Żak, Beata; Gałaś, Zdzisław; Utrata, Arkadiusz

    2017-03-01

    In the practice of mining companies, the vast majority of them produce more than one product. The analysis of the break-even, which is referred to as CVP (Cost-Volume-Profit) analysis (Wilkinson, 2005; Czopek, 2003) in their case is significantly constricted, given the necessity to include multi-assortment structure in the analysis, which may have more than 20 types of assortments (depending on the grain size) in their offer, as in the case of open-pit mines. The article presents methods of evaluation of break-even (volume and value) for both a single-assortment production and a multi-assortment production. The complexity of problem of break-even evaluation for multi-assortment production has resulted in formation of many methods, and, simultaneously, various approaches to its analysis, especially differences in accounting fixed costs, which may be either totally accounted for among particular assortments, relating to the whole company or partially accounted for among particular assortments and partially relating to the company, as a whole. The evaluation of the chosen methods of break-even analysis, given the availability of data, was based on two examples of mining companies: an open-pit mine of rock materials and an underground hard coal mine. The selection of methods was set by the available data provided by the companies. The data for the analysis comes from internal documentation of the mines - financial statements, breakdowns and cost calculations.

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

    International Nuclear Information System (INIS)

    Zee, Sung Kyun; Lee, Chung Chan; Lee, Chang Kue

    1996-02-01

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

  17. Review of RIA and LOCA criteria for WWER fuel

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    The RIA and LOCA fuel safety criteria are under revision in the international community of fuel suppliers, authorities and research organizations. The main criteria will be reviewed in the paper for WWER fuel. Experimental data on the fuel failure behaviour under reactivity-initiated accident (RIA) conditions produced in the last decade in French and Japanese test reactors indicated low failure enthalpy for high burnup fuel compared to fresh fuel. However the high burnup was not the only phenomenon influencing the fuel failure. The oxide scale on the external surface of the fuel rod, hydrogen content of the Zr cladding and the local hydriding seemed also be responsible for the failure at low enthalpy. Furthermore differences have been found between Western design fuel and Russian type WWER fuel. The burnup dependence of fuel failure for WWER fuel was found much less, probably due to the low oxidation during normal operational conditions compared to other PWRs. The recently published Vitanza and KAERI correlations for RIA failure enthalpy have been applied to 23 WWER tests. Experimental data from Russian IGR and BIGR reactors have been used. The calculations have shown that both burnup and cladding oxidation effects must be considered, however the pulse width dependence of failure enthalpy has not been confirmed. During loss of coolant accidents (LOCA) the peak cladding temperature and local oxidation criteria have to be met. The oxidation criterion is under discussion today in many laboratories. The AEKI carried out several experimental series with Zr1%Nb cladding used in WWER reactors. The paper will describe the main results of the tests and present the limit for ductile-brittle transition derived from ring compression test. The behaviour of Zr1%Nb (E110) and Zircaloy-4 claddings under LOCA conditions will be compared as well. (author)

  18. Comparison of models discribing cladding deformations during LOCA

    International Nuclear Information System (INIS)

    Chakraborty, A.K.; Zipper, R.

    1981-05-01

    This report compares the important models for the determination of cladding deformations during LOCA. In addition to the comparisons of underlying assumptions of different models the same is done for the coefficients applied for the models. In order to assess the predictive capability of the models the calculated results are compared with the experimental results of the individual claddings. It was found out that the results of temperature ramp tests could be calculated better than that of the pressure ramp tests. The calculations revealed that even with the simplified assumption of the model used in TESPA the agreement of the calculated results with those of model NORA was relatively good. (orig.) [de

  19. Transient analysis for total loss of feed water scenario due to postulated feed line breaks in both steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seyun; Kim, Minhee [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The total loss of feed water (TLOFW) scenario is a beyond design basis accident (DBA) for nuclear steam supply system (NSSS) but is considered in the safety analysis report (SAR) for the stress analysis of structures of KEPIC class 1 and the hydrogen generation analysis. The postulated single feed line break (FLB) scenario is DBA and is described in chapter 15.2 of SAR. To evaluate the safety impact, the integrity of plant and the coping measures, a transient of total loss of feed water due to the postulated breaks of both feed line is analyzed for OPR1000 with RELAP5 code. To evaluate the safety impact, the integrity of plant and the coping measures, a transient of total loss of feed water due to the postulated breaks of both feed line is analyzed for OPR1000 with RELAP5 code. The calculations show that the operation of safety depressurization system at 1800 seconds is an effective measure to mitigate the core damage due to the uncovery of the core according to the pressurization of primary loop. Through the sensitivity studies, it is presented that the peak cladding temperature is proportional to the break size.

  20. Inter-system LOCA risk assessment

    International Nuclear Information System (INIS)

    Galyean, W.J.; Kelly, D.L.; Schroeder, J.A.

    1991-01-01

    Inter-systems loss-of-coolant accidents (ISLOCAs) have been included in probabilistic risk assessments (PRAs) since WASH-1400. While estimated as being relatively low contributors to core damage frequency, ISLOCAs have been identified as major contributors to risk at nuclear power plants (NPPs). They have the potential to result in core melt and containment bypass, which may lead to the early release of large quantities of fission products. Recent events at several operating reactors have been identified as ISLOCA precursors. The occurrence of these events have raised concerns that the frequency of ISLOCA sequences might be underestimated in current state-of-the-art PRAs. In order to expand the current state-of-the-art, a Nuclear Regulatory Commission research program is being conducted by ED and G Idaho, Inc. at the Idaho National Engineering Laboratory. The objective of the ISLOCA research program is to generate qualitative and quantitative information on the hardware, human factors, and accident consequence issues that dominate nuclear power plant risks for ISLOCA. To meet this objective, the approach being taken includes analysis of all interfaces between the primary reactor coolant system and other, lower pressure systems. This historical experience (primarily, licensee event reports) has provided the basis for determining the scope of the analysis with respect to potential failure mechanisms of the pressure isolation boundary. It is important to note that in the vast majority of these events, the dominant failure was a human error. Because of their significance, human errors are given particular attention in the present analysis

  1. Numerical Ballooning and Burst Prediction of Fuel Cladding During LOCA Transients in LWR

    International Nuclear Information System (INIS)

    Landau, E.; Weiss, Y.; Szanto, M.

    2014-01-01

    Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis, most former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. Several codes for reactor core and fuel rod analysis exist nowadays, such as FRAPTRAN1.4 or RELAP5-3D . These are tailor-made codes, designed to predict general core behavior and fuel performance, and while they are also used in predicting core components behavior during accident conditions, including those of cladding ballooning and failure with good accuracy, they contain several limitations on modeling the full transient cladding thermo mechanical phenomena. Limitations such as mechanical models being one dimensional or in axisymmetric geometries only, relying mostly on analytical models therefore having further restricting assumptions in return for accuracy, etc. These limitations disable the simulation of several important aspects, such as modeling 3D azimuthal behavior for example. The objective of the current work is to develop a comprehensive numerical model for predicting zircalloy cladding thermo mechanical behavior during a LOCA. The model will eventually predicts full cladding ballooning and burst behavior followed by fuel relocation, for fuel rods that can be subjected to 3D distributed flux. The model is fully three dimensional and is created using the commercial FEM numerical simulation software ABAQUS© applying

  2. Hospital consultants breaking bad news with simulated patients: an analysis of communication using the Roter Interaction Analysis System.

    Science.gov (United States)

    Vail, Laura; Sandhu, Harbinder; Fisher, Joanne; Cooke, Heather; Dale, Jeremy; Barnett, Mandy

    2011-05-01

    To explore how experienced clinicians from wide ranging specialities deliver bad news, and to investigate the relationship between physician characteristics and patient centredness. Consultations involving 46 hospital consultants from 22 different specialties were coded using the Roter Interaction Analysis System. Consultants mainly focussed upon providing biomedical information and did not discuss lifestyle and psychosocial issues frequently. Doctor gender, age, place of qualification, and speciality were not significantly related to patient centredness. Hospital consultants from wide ranging specialities tend to adopt a disease-centred approach when delivering bad news. Consultant characteristics had little impact upon patient centredness. Further large-scale studies are needed to examine the effect of doctor characteristics on behaviour during breaking bad news consultations. It is possible to observe breaking bad news encounters by video-recording interactions between clinicians and simulated patients. Future training programmes should focus on increasing patient-centred behaviours which include actively involving patients in the consultation, initiating psychosocial discussion, and providing patients with opportunities to ask questions. Copyright © 2010 Elsevier Ireland Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Drubay, B.; Chapuliot, St.; Lacire, M.H.; Marie, St.; Deschanels, H.; Cambefort, P.

    2001-01-01

    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 R -Δa curve or J i /G 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 σ approach calculating stress and strain at a characteristic distance d from the crack tip; evaluation of fatigue crack growth based on da/dN-ΔK eff relationship with a ΔK eff derived from a simplified estimation of ΔJ for the cyclic load; evaluation of creep-fatigue crack growth adding the fatigue crack growth and the creep crack growth during the hold time derived from a simplified evaluation of C * ; Leak-Before-Break procedure. The fracture mechanic parameters determined in the A16 guide (K 1 , J, C * ) are derived from handbooks and formula in accordance with those proposed in the RSE-M document for in service inspection. Those are: the K 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 * formulations for mechanical and through thickness thermal load. The main part of the formula and assessment methodologies proposed in the A16 guide are included in a software, called MJSAM, developed under the MS Windows environment in support of the document. This allows a simple application of the analysis proposed in the document. (authors)

  4. Dam break analysis and flood inundation map of Krisak dam for emergency action plan

    Science.gov (United States)

    Juliastuti, Setyandito, Oki

    2017-11-01

    The Indonesian Regulation which refers to the ICOLD Regulation (International Committee on Large Dam required have the Emergency Action Plan (EAP) guidelines because of the dams have potential failure. In EAP guidelines there is a management of evacuation where the determination of the inundation map based on flood modeling. The purpose of the EAP is to minimize the risk of loss of life and property in downstream which caused by dam failure. This paper will describe about develop flood modeling and inundation map in Krisak dam using numerical methods through dam break analysis (DBA) using hydraulic model Zhong Xing HY-21. The approaches of dam failure simulation are overtopping and piping. Overtopping simulation based on quadrangular, triangular and trapezium fracture. Piping simulation based on cracks of orifice. Using results of DBA, hazard classification of Krisak dam is very high. The nearest village affected dam failure is Singodutan village (distance is 1.45 kilometer from dam) with inundation depth is 1.85 meter. This result can be used by stakeholders such as emergency responders and the community at risk in formulating evacuation procedure.

  5. 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.

  6. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  7. A New First Break Picking for Three-Component VSP Data Using Gesture Sensor and Polarization Analysis.

    Science.gov (United States)

    Li, Huailiang; Tuo, Xianguo; Shen, Tong; Wang, Ruili; Courtois, Jérémie; Yan, Minhao

    2017-09-19

    A new first break picking for three-component (3C) vertical seismic profiling (VSP) data is proposed to improve the estimation accuracy of first arrivals, which adopts gesture detection calibration and polarization analysis based on the eigenvalue of the covariance matrix. This study aims at addressing the problem that calibration is required for VSP data using the azimuth and dip angle of geophones, due to the direction of geophones being random when applied in a borehole, which will further lead to the first break picking possibly being unreliable. Initially, a gesture-measuring module is integrated in the seismometer to rapidly obtain high-precision gesture data (including azimuth and dip angle information). Using re-rotating and re-projecting using earlier gesture data, the seismic dataset of each component will be calibrated to the direction that is consistent with the vibrator shot orientation. It will promote the reliability of the original data when making each component waveform calibrated to the same virtual reference component, and the corresponding first break will also be properly adjusted. After achieving 3C data calibration, an automatic first break picking algorithm based on the autoregressive-Akaike information criterion (AR-AIC) is adopted to evaluate the first break. Furthermore, in order to enhance the accuracy of the first break picking, the polarization attributes of 3C VSP recordings is applied to constrain the scanning segment of AR-AIC picker, which uses the maximum eigenvalue calculation of the covariance matrix. The contrast results between pre-calibration and post-calibration using field data show that it can further improve the quality of the 3C VSP waveform, which is favorable to subsequent picking. Compared to the obtained short-term average to long-term average (STA/LTA) and the AR-AIC algorithm, the proposed method, combined with polarization analysis, can significantly reduce the picking error. Applications of actual field

  8. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

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

    1982-04-01

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

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

    International Nuclear Information System (INIS)

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

    1989-02-01

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

  10. Proteome analysis of Norway maple (Acer platanoides L. seeds dormancy breaking and germination: influence of abscisic and gibberellic acids

    Directory of Open Access Journals (Sweden)

    Pawłowski Tomasz A

    2009-05-01

    Full Text Available Abstract Background Seed dormancy is controlled by the physiological or structural properties of a seed and the external conditions. It is induced as part of the genetic program of seed development and maturation. Seeds with deep physiological embryo dormancy can be stimulated to germinate by a variety of treatments including cold stratification. Hormonal imbalance between germination inhibitors (e.g. abscisic acid and growth promoters (e.g. gibberellins is the main cause of seed dormancy breaking. Differences in the status of hormones would affect expression of genes required for germination. Proteomics offers the opportunity to examine simultaneous changes and to classify temporal patterns of protein accumulation occurring during seed dormancy breaking and germination. Analysis of the functions of the identified proteins and the related metabolic pathways, in conjunction with the plant hormones implicated in seed dormancy breaking, would expand our knowledge about this process. Results A proteomic approach was used to analyse the mechanism of dormancy breaking in Norway maple seeds caused by cold stratification, and the participation of the abscisic (ABA and gibberellic (GA acids. Forty-four proteins showing significant changes were identified by mass spectrometry. Of these, eight spots were identified as water-responsive, 18 spots were ABA- and nine GA-responsive and nine spots were regulated by both hormones. The classification of proteins showed that most of the proteins associated with dormancy breaking in water were involved in protein destination. Most of the ABA- and GA-responsive proteins were involved in protein destination and energy metabolism. Conclusion In this study, ABA was found to mostly down-regulate proteins whereas GA up-regulated proteins abundance. Most of the changes were observed at the end of stratification in the germinated seeds. This is the most active period of dormancy breaking when seeds pass from the quiescent

  11. Finite element analysis of laser-generated ultrasound for characterizing surface-breaking cracks

    International Nuclear Information System (INIS)

    Jeong, Hyun Jo

    2005-01-01

    A finite element method was used to simulate the wave propagation of laser-generated ultrasound and its interaction with surface breaking cracks in an elastic material. Thermoelastic laser line source on the material surface was approximated as a shear dipole and loaded as nodal forces in the plane-strain Finite Element (FE) model. The shear dipole-FE model was tested for the generation of ultrasound on the surface with no defect. The model was found to generate the Rayleigh surface wave. The model was then extended to examine the interaction of laser generated ultrasound with surface-breaking cracks of various depths. The crack-scattered waves were monitored to size the crack depth. The proposed model clearly reproduced the experimentally observed features that can be used to characterize the presence of surface-breaking cracks

  12. THE DIAGNOSTICS OF INDUCTION MOTORS ROTOR BAR BREAKS BASED ON THE ANALYSIS OF ELECTROMOTIVE FORCE IN THE STATOR WINDINGS

    Directory of Open Access Journals (Sweden)

    M.V. Zagirnyak

    2014-12-01

    Full Text Available A method for diagnostics of the induction motor rotor bar breaks, based on the wavelet-analysis of the electromotive force induced in the stator windings in the rundown mode is developed. A method for decomposition of the electromotive force of the stator winding phase to the electromotive force signals of the active sides of winding coils using Z-transformation theory is developed. The effectiveness of the proposed diagnostic method was experimentally confirmed.

  13. Analysis of structural breaks in the stock market integration of mexico into world

    OpenAIRE

    Arouri Mohamed el hédi; Jamel Jouini

    2009-01-01

    This paper studies the Mexican stock market integration process. First, we estimate the time-varying Mexican degree of market integration using an international CAPM with segmentation effects. Second, we study the structural breaks in this series. Finally, we relate the obtained results to important facts and economic events.

  14. study and analysis of asa river hypothetical dam break using hec-ras

    African Journals Online (AJOL)

    Windows User

    Impounded reservoirs provide beneficial functions such as flood control, recreation, hydropower and water supply but they also carry potential risks. Spontaneous dam break phenomenon can occur and the resultant flooding may cause substantial loss of life and property damage downstream of the dam. A hypothetical dam ...

  15. West Coast tree improvement programs: a break-even, cost-benefit analysis

    Science.gov (United States)

    F. Thomas Ledig; Richard L Porterfield

    1981-01-01

    Three tree improvement programs were analyzed by break-even, cost-benefit technique: one for ponderosa pine in the Pacific Northwest, and two for Douglas-fir in the Pacific Northwest-one of low intensity and the other of high intensity. A return of 8 percent on investment appears feasible by using short rotations or by accompanying tree improvement with thinning....

  16. Retrospective analysis of the financial break-even point for intrathecal morphine pump use in Korea.

    Science.gov (United States)

    Kim, Eun Kyoung; Shin, Ji Yeon; Castañeda, Anyela Marcela; Lee, Seung Jae; Yoon, Hyun Kyu; Kim, Yong Chul; Moon, Jee Youn

    2017-10-01

    The high cost of intrathecal morphine pump (ITMP) implantation may be the main obstacle to its use. Since July 2014, the Korean national health insurance (NHI) program began paying 50% of the ITMP implantation cost in select refractory chronic pain patients. The aims of this study were to investigate the financial break-even point and patients' satisfaction in patients with ITMP treatment after the initiation of the NHI reimbursement. We collected data retrospectively or via direct phone calls to patients who underwent ITMP implantation at a single university-based tertiary hospital between July 2014 and May 2016. Pain severity, changes in the morphine equivalent daily dosage (MEDD), any adverse events, and patients' satisfaction were determined. We calculated the financial break-even point of ITMP implantation via investigating the patient's actual medical costs and insurance information. During the studied period, 23 patients received ITMP implantation, and 20 patients were included in our study. Scores on an 11-point numeric rating scale (NRS) for pain were significantly reduced compared to the baseline value ( P break-even point was 28 months for ITMP treatment after the NHI reimbursement policy. ITMP provided effective chronic pain management with improved satisfaction and reasonable financial break-even point of 28 months with 50% financial coverage by NHI program.

  17. MATLAB: break

    OpenAIRE

    2005-01-01

    Interactive Media Element This interactive tutorial on MATLAB covers the use of the break function. An example demonstrates how the break function affects a for loop with if instruction embedded in the loop. An example is provided with step-by-step animated explanation. The interactions involve entering MATLAB instructions and observing the outcomes. Self-check questions are provided to help learners determine their level of understanding of the content presented. EC1010 Introduction...

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  19. Prediction of municipal solid waste generation using artificial neural network approach enhanced by structural break analysis.

    Science.gov (United States)

    Adamović, Vladimir M; Antanasijević, Davor Z; Ristić, Mirjana Đ; Perić-Grujić, Aleksandra A; Pocajt, Viktor V

    2017-01-01

    This paper presents the development of a general regression neural network (GRNN) model for the prediction of annual municipal solid waste (MSW) generation at the national level for 44 countries of different size, population and economic development level. Proper modelling of MSW generation is essential for the planning of MSW management system as well as for the simulation of various environmental impact scenarios. The main objective of this work was to examine the potential influence of economy crisis (global or local) on the forecast of MSW generation obtained by the GRNN model. The existence of the so-called structural breaks that occur because of the economic crisis in the studied period (2000-2012) for each country was determined and confirmed using the Chow test and Quandt-Andrews test. Two GRNN models, one which did not take into account the influence of the economic crisis (GRNN) and another one which did (SB-GRNN), were developed. The novelty of the applied method is that it uses broadly available social, economic and demographic indicators and indicators of sustainability, together with GRNN and structural break testing for the prediction of MSW generation at the national level. The obtained results demonstrate that the SB-GRNN model provide more accurate predictions than the model which neglected structural breaks, with a mean absolute percentage error (MAPE) of 4.0 % compared to 6.7 % generated by the GRNN model. The proposed model enhanced with structural breaks can be a viable alternative for a more accurate prediction of MSW generation at the national level, especially for developing countries for which a lack of MSW data is notable.

  20. An analysis of the importance of appropriate tie breaking rules in dispatch heuristics

    Directory of Open Access Journals (Sweden)

    Jorge M. S. Valente

    2006-04-01

    Full Text Available In this paper, we analyse the effect of using appropriate tie breaking criteria in dispatch rules. We consider four different dispatch procedures, and for each of these heuristics we compare two versions that differ only in the way ties are broken. The first version breaks ties randomly, while the second uses a criterion that incorporates problem-specific knowledge. The computational results show that using adequate tie breaking criteria improves the performance of the dispatch heuristics. The magnitude of the improvement is different for the four heuristics, and also depends on the characteristics of each specific instance. The use of problem-related knowledge for breaking ties should therefore be given some consideration in the implementation of dispatch rules.Neste artigo é analisado o efeito da utilização de regras de desempate apropriadas na eficácia de regras de despacho. São consideradas quatro regras de despacho diferentes, e para cada uma destas heurísticas são comparadas duas versões que diferem no modo como os empates são resolvidos. A primeira versão resolve os empates de forma aleatória, enquanto a segunda utiliza um critério que incorpora informação relativa ao problema em causa. Os resultados computacionais mostram que a utilização de critérios de desempate adequados melhora o desempenho das regras de despacho. A magnitude da melhoria é diferente para as quatro heurísticas, e depende igualmente das características específicas de cada instância. A utilização de informação relativa ao problema em causa para a resolução de empates deve assim ser considerada na implementação de regras de despacho.

  1. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Muhammad Subekti

    2009-01-01

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  2. The probability of intersystem LOCA: impact due to leak testing and operational changes. Technical report

    International Nuclear Information System (INIS)

    Rubin, M.P.

    1980-05-01

    The Reactor Safety Study (WASH-1400) identified the potential intersystem LOCA in a pressurized water reactor as a significant contributor to the risk resulting from core melt. Similar scenarios are also possible in boiling water reactors. This report evaluates various pressure isolation valve configurations used in reactors to determine the probability of intersystem LOCA. It is shown that periodic leak testing of these valves can substantially reduce intersystem LOCA probability. Specific analyses of the high pressure/low pressure interfaces in the Sequoyah (PWR) and Alan B. Barton (BWR) plants show that periodic leak testing of the pressure isolation check valves will reduce the intersystem LOCA probability to below 0.000001 per year

  3. The purchasing power parity in emerging Europe: Empirical results based on two-break analysis

    Directory of Open Access Journals (Sweden)

    Mladenović Zorica

    2013-01-01

    Full Text Available The purpose of the paper is to evaluate the validity of purchasing power parity (PPP for eight countries from the Emerging Europe: Hungary, Czech Republic, Poland, Romania, Lithuania, Latvia, Serbia and Turkey. Monthly data for euro and U.S. dollar based real exchange rate time series are considered covering the period: January, 2000 - August, 2011. Given significant changes in these economies in this sample it seems plausible to assume that real exchange time series are characterized by more than one time structural break. In order to endogenously determine the number and type of breaks while testing for the presence of unit roots we applied the Lee-Strazicich approach. For two euro based real exchange rate time series (in Hungary and Turkey the unit root hypothesis has been rejected. For the U.S. dollar based real exchange rate time series in Poland, Romania and Turkey the presence of unit root has been rejected. To assess the adjustment dynamics of those real exchange rates that were detected to be stationary with two breaks, the impulse response function is calculated and half-life is estimated. Our overall conclusion is that the persistence of real exchange rate in Emerging Europe is still substantially high. The lack of strong empirical support for PPP suggests that careful policy actions are needed in this region to prevent serious exchange rate misalignment.

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

    International Nuclear Information System (INIS)

    2013-06-01

    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

  5. Effects of Changes in Oil Prices on Russian Economy: Analysis of Cointegration with Multiple Structural Breaks and Symmetric Causality

    Directory of Open Access Journals (Sweden)

    İsmet GÖÇER

    2015-12-01

    Full Text Available In this paper, effects of changes in oil prices on Russian economy is analyzed with the help of cointegration with multiple structural breaks and symmetric causality tests, for the period of 1992Q1-2014Q3. In this context, stationarity of the series is investigated by Kapetanios (2005 unit root test with multiple structural breaks and it is found that the series are not stationary at level values, but stationary when their first differences are taken. Causality relations between series are investigated by Hacker and Hatemi-J (2012 symmetric causality test and it is seen that causality relation exists from oil prices to export, foreign trade balance and national income. Existence of cointegration relation between series is tested by Maki (2012 method of cointegration with multiple structural break and it is found that the series are cointegrated. Long run analysis is done and it is estimated that 1 percent increase in oil prices increase the export, foreign trade balance and national income of Russia by 1.01 percent, 0.27 percent and 0.13 percent respectively.

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

    Directory of Open Access Journals (Sweden)

    Omid Noori-Kalkhoran

    2016-10-01

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

  7. Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

    International Nuclear Information System (INIS)

    Bujan, A.; Adamik, V.; Misak, J.

    1986-01-01

    A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

  8. 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.

  9. The Break

    DEFF Research Database (Denmark)

    Strand, Anete Mikkala Camille

    2018-01-01

    The chapter elaborates on how to deal with one of the major challenges facing organizations worldwide; Stress. The Break enacts a quantum approach to meet the challenges by proposing a combination of three different quantum storytelling technologies; protreptic mentoring, walking and material sto...

  10. Supersymmetry breaking

    Indian Academy of Sciences (India)

    symmetry asks for the existence of a bosonic massless partner, which generically correspond to a non-compact flat direction. ... Hidden Sector nonren.int. Ti, 〈Fi〉 = 0. Let us define the strength of supersymmetry breaking by F2 = ∑ ... partners, not protected by the GIM mechanism. Solutions for FCNC problem: (i) Flavour ...

  11. 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.

  12. 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.

  13. Analysis of DNA strand break induction and repair in plants from the vicinity of Chernobyl

    International Nuclear Information System (INIS)

    Syomov, A.B.; Ptitsyna, S.N.; Sergeeva, S.A.

    1992-01-01

    For 3 years following the Chernobyl accident DNA repair efficiency was studied in irradiated and control populations of various plan species. Compared with the control populations, some irradiated populations exhibited increases in the yield of DNA single-strand breaks per unit dose of challenge radiation. The effect was registered in low-dose-rate alpha-irradiated populations, but was absent in plant populations growing in conditions of low-dose-rate beta-irradiation. The efficiency of single-strand DNA repair was identical in control and irradiated populations and approximated 100%. (author). 12 refs.; 1 fig.; 2 tabs

  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. Analysis of breaks in pipe connections to the hot leg and to the loop seal in the primary system of Ringhals 2 PWR

    International Nuclear Information System (INIS)

    Nilsson, L.; Sjoeberg, A.

    1987-01-01

    Analysis has been made of seven different cases of breaks in pipes connected to the hot leg and to the loop seal in Ringhals 2 PWR. The pipes, in which guillotine breaks are postulated, have nominal diameters ranging from 1 to 14 inches. The purpose of the analysis is to supplement the basis for a review of the inspection procedures for the safety of pressure vessels prescribed by SKI. A similar analysis already exists concerning breaks in the cold leg connections. The analysis has been made using the thermal hydraulic computer code RELAPS/MOD2 and with best estimate assumptions. This means that normal operating conditions have been adopted for the input to the calculations. However, the capacity of the safety injection system was assumed to be reduced by having one pump not operating each of the low pressure and high pressure safety injection system. The results of the analysis are presented in tables and as computer plots. The analysis shows that the consequences with respect to increased fuel rod and cladding temperatures are quite harmless. Only the two cases with the largest break sizes lead to critical heat flux (CHF) and increased temperatures for the hottest rods in the core. The peak cladding temperature was 636 degrees C for the 12 inch break. In both cases rewetting occurred within 25 s of accident initiation. In the cases with breaks in connections of 6 inch diameter and smaller the RELAP5 calculations indicated a substantial margin to CHF throughout the transient. (authors)

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

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

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

  17. Electroweak breaking and supersymmetry breaking

    Indian Academy of Sciences (India)

    We discuss the clash between the absence of fine tuning in the Higgs potential and a sufficient suppression of flavour changing neutral current transitions in supersymmetric extensions of the standard model. It is pointed out that horizontal U ( 1 ) symmetry combined with the D -term supersymmetry breaking provides a ...

  18. 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.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. Analysis OpenMP performance of AMD and Intel architecture for breaking waves simulation using MPS

    Science.gov (United States)

    Alamsyah, M. N. A.; Utomo, A.; Gunawan, P. H.

    2018-03-01

    Simulation of breaking waves by using Navier-Stokes equation via moving particle semi-implicit method (MPS) over close domain is given. The results show the parallel computing on multicore architecture using OpenMP platform can reduce the computational time almost half of the serial time. Here, the comparison using two computer architectures (AMD and Intel) are performed. The results using Intel architecture is shown better than AMD architecture in CPU time. However, in efficiency, the computer with AMD architecture gives slightly higher than the Intel. For the simulation by 1512 number of particles, the CPU time using Intel and AMD are 12662.47 and 28282.30 respectively. Moreover, the efficiency using similar number of particles, AMD obtains 50.09 % and Intel up to 49.42 %.

  1. Breaking Symmetries

    Directory of Open Access Journals (Sweden)

    Kirstin Peters

    2010-11-01

    Full Text Available A well-known result by Palamidessi tells us that πmix (the π-calculus with mixed choice is more expressive than πsep (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 offered 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 (initial 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 reasonable encoding from πmix into πsep. We indicate how the respective proofs can be adapted and exhibit the consequences of varying notions of uniformity and reasonableness. In each case, the ability to break initial symmetries turns out to be essential.

  2. Evidence-based analysis of prophylactic treatment of asymptomatic retinal breaks and lattice degeneration.

    Science.gov (United States)

    Wilkinson, C P

    2000-01-01

    To assess the quality of information in the literature regarding the benefits of prophylactic treatment of asymptomatic retinal tears and lattice degeneration. Asymptomatic retinal breaks occur in approximately 7% of patients over age 40, and lattice degeneration is present in approximately 8% of the general population. Because retinal breaks cause retinal detachment and lattice degeneration is associated with approximately 30% of retinal detachments, prophylactic treatment of these lesions has sometimes been recommended. A panel of vitreoretinal experts performed a literature review of all publications regarding prevention of retinal detachment that have been published in English. These articles were then used to prepare recommendations for patient care in an American Academy of Ophthalmology Preferred Practice Pattern (PPP). Each recommendation was rated according to: (1) its importance in the care process and (2) the strength of evidence supporting the given recommendation. Most recommendations were rated as A (most important to patient care). Only a single publication was graded as I (providing strong evidence in support of a recommendation), and this was not a prospective trial. Of the few publications rated as II (substantial evidence), most were studies documenting a lack of treatment benefit. Because of an absence of level I and level II studies in the literature, level III (consensus of expert opinion) was the basis for most recommendations in the PPP. The current literature regarding prevention of retinal detachment does not provide sufficient information to support strongly prophylactic treatment of lesions other than symptomatic flap tears. Prospective randomized trials of prophylactic therapy are indicated. Eyes highly predisposed to retinal detachment should be considered for such studies.

  3. CONTEMPT-4MOD3, LWR Containment Long-Term Pressure Distribution and Temperature Distribution in LOCA

    International Nuclear Information System (INIS)

    Lin, C.C.; Economos, C.; Lehner, J.R.; Maise, G.; Ng, K.K.; Mirsky, S.M.

    2002-01-01

    1 - Description of problem or function: CONTEMPT-4/MOD5 describes the response of multi-compartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user- supplied descriptions of compartments, inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. To accommodate degraded core type accidents, analytical models for hydrogen combustion within compartments and energy transfer due to gas radiation are also provided. CONTEMPT4/MOD6 is an update of previous CONTEMPT4 versions. Improvements in CONTEMPT4/MOD6 over CONTEMPT4/MOD3 include coding of a BWR pressure suppression system model, a hydrogen/carbon monoxide burn model, a gas radiation heat transfer model, a user specified variable junction (leakage) area as a function of pressure or time, additional heat transfer coefficient options for heat structures, generalized initial compartment conditions for inerted containment, an alternative containment spray model and spray carry-over capability. Also, the thermodynamic properties routines have been extended to accommodate the higher temperature and multicomponent gas mixtures associated with combustion. In addition, reduced running time is achieved by incorporation of an optional implicit numerical algorithm for junction flow. This makes economically feasible the analysis of very long

  4. In-pile investigations at the PHEBUS facility of the behavior of PWR-type fuel bundles in typical L.B. loca transients extended to and beyond the limits of ECCS criteria

    International Nuclear Information System (INIS)

    Duco, J.; Reocreux, M.; Tattegrain, A.; Berna, P.; Legrand, B.; Trotabas, M.

    1984-11-01

    An in-pile investigation is currently carried out at the PHEBUS facility of the behavior of .8m active height, 25-rod PWR-type fuel bundles during simulated large-break LOCA (L.B. LOCA) reactor transients. A first series of six tests using pressurized rods is to be completed by the end of 1984, relative to a conservatively calculated 2-peak cladding temperature transient at the hot point, as considered in the French 900 MW(e) PWR standard safety report. The severity of such a transient has been increased in the tests so as to check the bundle behavior at the limits of the first two NRC ECCS criteria, which were, in fact, locally exceeded in one test. Three of the tests are reported on hereunder. Short coplanar cladding balloonings were observed at the hot point level, which resulted in maximum flow blockage ratios of about 50%. Severe cladding embrittlement against thermal shock and subsequent handling was observed in the test where the criteria were exceeded. Prediction of the overall thermal-hydraulic behavior in the bundle was good, using the RELAP 4 MOD 6 code. Cladding strains are generally overevaluated by codes such as FRAPT 4 or CUPIDON, which currently do not take into account azimuthal cladding temperature gradients. Other L.B. LOCA test series are envisaged from 1986 on, based on transients calculated with ''physical'' models

  5. LOFT/LP-02-6, Loss of Fluid Test, 1. OECD Large Break Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The fourth OECD LOFT experiment was conducted on 3 October 1983. This was the first OECD LOFT large break experiment. The initial and boundary conditions were chosen to be representative of USNRC licensing limits for commercial PWRs. This included loss of off-site power coincident with LOCA initiation. This experiment included the first use in LOFT of pressurized fuel rods in the center bundle. The experiment was initiated by opening the quick-opening blow-down valves in the broken hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  6. Overview of DOE proposed loss-of-coolant accident (LOCA) rule-revision study

    International Nuclear Information System (INIS)

    Hanson, D.J.; Batt, D.L.

    1982-01-01

    In 1981, an independent review group, a subcommittee of the President's Nuclear Oversight Committee, recognized the conservatism in particular features of the LOCA rule and recommended that research data be explicitly included in the licensing process. Also, an initial review indicated that a generic study, which is the subject of this paper, should be performed which will use research data to provide the basis for proposed revisions to the LOCA rule. The objective of this study is to develop and propose revisions to the current LOCA rule that will provide a defensible technical base upon which a more realistic assessment of safety issues can be made. From this proposed new rule, potential reductions in costs can be effected while maintaining the current level of safety. A survey of three NSSS Vendors, 14 utilities, and several other industry-related organizations was conducted to assess the industry interest, attitudes, and ideas regarding changing the LOCA rule. The survey indicated there would be benefits in having a revised LOCA rule

  7. Simulation of high burn-up fuel cladding and its safety assessment under LOCA condition

    International Nuclear Information System (INIS)

    Park, Dong Jun; Won, Sung Bin; Choi, Byoung Kwon; Park, Jeong Yong; Koo, Yang Hyun

    2011-01-01

    Current LOCA safety criteria was established in the beginning of 1970s and based on the results obtained from non-irradiated Zircaloy-4 claddings. Because of major advantages in fuel-cycle costs, reactor operation, and waste management, the increase in fuel discharge burn-up is current worldwide trend in the nuclear industry. As the fuel burn-up increases, various phenomena unexpected have been reported due to changes in the condition of reactor operation and in-core environment. Since, it should be considered whether the current Loss-of-coolant accident (LOCA) criteria is suitable for high burn-up fuel cladding or not. In addition, many fuel vendors have recently developed new cladding alloys superior to Zircaloy-4 cladding. The performance of these advanced cladding alloys under LOCA, especially at high burn-up, is not well understood at this time. To better understand high burn-up effects and commercialize new cladding alloys, study of LOCA-related behavior of various types of high burn-up fuel cladding and their data base is essentially required. In this background, postulated LOCA test has been carried out with prehydrided Zircaloy-4 cladding as a surrogate for high burn-up cladding and the relevant results obtained are discussed

  8. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  9. Level-Swell Prediction With RETRAN-3D And Its Application To A BWR Steam-Line-Break Analysis

    International Nuclear Information System (INIS)

    Aounallah, Y.; Hofer, K.

    2003-01-01

    Level-swell experiments have often been simulated using system codes, such as TRAC and RELAP, but only cursory assessments have been performed with the operational-transient code RETRAN-3D, the main system code used within the STARS project. The present study, initiated in the framework of a BWR Steam-Line-Break (SLB) accident scenario, addresses this lacuna by performing RETRAN simulations of the General Electric Level-Swell experiments, and by investigating their implications on power plant accident analyses. Parameters to which the predicted level swell is sensitive have been identified, and recommendations on code options are made. The SLB analysis objective was to determine the amount of steam and liquid discharged through the break under specified boundary conditions, and to gauge the results against reference values. The impact of the nodalization of the upper part of the reactor pressure vessel was investigated and found to play an important role, whereas the level swell induced from flashing was found not to be the predominant factor for these simulations. (author)

  10. 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.

  11. Complex Langevin analysis of the spontaneous symmetry breaking in dimensionally reduced super Yang-Mills models

    Science.gov (United States)

    Anagnostopoulos, Konstantinos N.; Azuma, Takehiro; Ito, Yuta; Nishimura, Jun; Papadoudis, Stratos Kovalkov

    2018-02-01

    In recent years the complex Langevin method (CLM) has proven a powerful method in studying statistical systems which suffer from the sign problem. Here we show that it can also be applied to an important problem concerning why we live in four-dimensional spacetime. Our target system is the type IIB matrix model, which is conjectured to be a nonperturbative definition of type IIB superstring theory in ten dimensions. The fermion determinant of the model becomes complex upon Euclideanization, which causes a severe sign problem in its Monte Carlo studies. It is speculated that the phase of the fermion determinant actually induces the spontaneous breaking of the SO(10) rotational symmetry, which has direct consequences on the aforementioned question. In this paper, we apply the CLM to the 6D version of the type IIB matrix model and show clear evidence that the SO(6) symmetry is broken down to SO(3). Our results are consistent with those obtained previously by the Gaussian expansion method.

  12. Stress Analysis and Model Test of Rock Breaking by Arc Blade Wedged Hob

    Directory of Open Access Journals (Sweden)

    Ying-chao Liu

    2016-07-01

    Full Text Available Based on rock compression-shear damage theory, the mechanical characteristics of an arc blade wedged hob were analyzed to study the rock fragmentation mechanism of hob during excavation, and rock fragmentation forecasting model of the arc blade wedged hob was improved. A spoke type cutter model which is similar to the tunnel boring machine (TBM cutter head was designed to study the rock fragmentation efficiency in different cutter spacing by adjusting the bearing sleeve size to obtain different distances between the hobs. The results show that the hob-breaking rock force mainly comes from three directions. The vertical force along the direction of the tunnel excavation, which is associated with uniaxial compressive strength of rock mass, plays a key role in the process of rock fragmentation. Field project data shows that the prediction model’s results of rock fragmentation in this paper are closer to the measured results than the results of the traditional linear cutting model. The optimal cutter spacing exists among different cutter spacings to get higher rock fragmentation rate and lower energy consumption during rock fragmentation. It is of great reference significance to design the arc blade wedged hob and enhance the efficiency of rock fragmentation in rock strata.

  13. Analysis of DNA double-strand break repair pathways in mice

    International Nuclear Information System (INIS)

    Brugmans, Linda; Kanaar, Roland; Essers, Jeroen

    2007-01-01

    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

  14. Detailed analysis of open clusters: A mass function break and evidence of a fundamental plane

    Science.gov (United States)

    Bonatto, C.; Bica, E.

    2005-07-01

    We derive photometric, structural and dynamical evolution-related parameters of 11 nearby open clusters with ages in the range 70 Myr to 7 Gyr and masses in the range ≈400 M_⊙ to ≈5300 M_⊙. The clusters are homogeneously analysed by means of J, H and KS 2MASS photometry, which provides spatial coverage wide enough to properly take into account the contamination of the cluster field by Galaxy stars. Structural parameters such as core and limiting radii are derived from the background-subtracted radial density profiles. Luminosity and mass functions (MFs) are built for stars later than the turnoff and brighter than the 2MASS PSC 99.9% completeness limit. The total mass locked up in stars in the core and the whole cluster, as well as the corresponding mass densities, are calculated by taking into account the observed stars (evolved and main sequence) and extrapolating the MFs down to the H-burning mass limit, 0.08 M_⊙. We illustrate the methods by analysing for the first time in the near-infrared the populous open clusters NGC 2477 and NGC 2516. For NGC 2477 we derive an age of 1.1 ± 0.1 Gyr, distance from the Sun d_⊙=1.2 ± 0.1 kpc, core radius Rcore=1.4 ± 0.1 pc, limiting radius Rlim=11.6 ± 0.7 pc and total mass mtot≈(5.3±1.6) × 103 M_⊙. Large-scale mass segregation in NGC 2477 is reflected in the significant variation of the MF slopes in different spatial regions of the cluster, and in the large number-density of giant stars in the core with respect to the cluster as a whole. For NGC 2516 we derive an age of 160 ± 10 Myr, d_⊙=0.44 ± 0.02 kpc, Rcore=0.6 ± 0.1 pc, Rlim=6.2 ± 0.2 pc and mtot≈(1.3±0.2) × 103 M_⊙. Mass-segregation in NGC 2516 shows up in the MFs. Six of the 11 clusters present a slope break in the MF occurring at essentially the same mass as that found for the field stars in Kroupa's universal IMF. The MF break is not associated to cluster mass, at least in the clusters in this paper. In two clusters the low-mass end of

  15. 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.

  16. 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

  17. Experiment of the downcomer effective water head during a reflood phase of PWR LOCA

    International Nuclear Information System (INIS)

    Sudo, Yukio; Murao, Yoshio

    1978-12-01

    The results and analysis are described of a downcomer effective water head experiment. Downcomer effective water head is the driving force to feed an emergency coolant to the core during a reflood phase of PWR LOCA. The test rig has dimensions of the full-scale height and gap. Experimental conditions are: downcomer wall temperature = 250 0 -- 300 0 C, back pressure = 1 atm, coolant temperature = 98 0 -- 100 0 C, extraction water velocity = 0 -- 2 cm/s, and gap size = 200 mm. The effective water head histories obtained by experiment were compared with those predicted from the heat release from the downcomer walls. The heat release was calculated from the temperature histories indicated by thermocouples instrumented in and on the walls during experiment. The following were revealed: (1) The relation of heat flux and superheat (q vs ΔT sub(s)) obtained in the experiment is much different from that in pool boiling. (2) The predicted effective water head is in good agreement with the experimental one after 120 sec from the initiation of coolant injection. (3) The effect of extraction water velocity is negligible. (4) The effect of initial wall temperatures is evident. (author)

  18. Transient deformation properties of Zircaloy for LOCA simulation. Volume 2. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hann, C.R.; Mohr, C.L.; Busness, K.M.; Olson, N.J.; Reich, F.R.; Stewart, K.B.

    1978-03-01

    The creep/creep rupture anisotropic properties of Zircaloy were determined and compared by analytical techniques with ramp-pressure and ramp-temperature test results. Tests were performed over the temperature range of 600/sup 0/F (589 K) to 2200/sup 0/F (1477 K), with the emphasis on the 800/sup 0/F (700 K) to 2000/sup 0/F (1366 K) temperature levels in low pressure air (6.5 x 10/sup -5/ atm) and in a 1 atm mixture of 20% oxygen, 80% argon. Stress levels of 60 to 95% of the ultimate tensile stress were used for the majority of the tests at each of the temperature levels tested, with selected tests performed as low as 30% of the ultimate tensile stress. Biaxial and uniaxial testing modes were used to evaluate the anisotropic deformation behavior. The combination of test results and predictive-analysis techniques developed as part of this program make it possible to predict the transient deformation of reactor fuel cladding during simulated LOCA conditions. Results include creep/creep rupture strain numerical constitutive relationships out to 120 seconds, computer codes and ramp test data.

  19. Transient deformation properties of Zircaloy for LOCA simulation. Final report, Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Hann, C.R.; Mohr, C.L.; Busness, K.M.; Olson, N.J.; Reich, F.R.; Stewart, K.B.

    1978-03-01

    The creep/creep rupture anisotropic properties of Zircaloy were determined and compared by analytical techniques with ramp-pressure and ramp-temperature test results. Tests were performed over the temperature range of 600/sup 0/F (589/sup 0/K) to 2200/sup 0/F (1477/sup 0/K), with the emphasis on the 800/sup 0/F (700/sup 0/K) to 2000/sup 0/F (1366/sup 0/K) temperature levels in low pressure air (6.5 x 10/sup -5/ atm) and in a 1 atm mixture of 20 percent oxygen, 80 percent argon. Stress levels of 60 to 95 percent of the ultimate tensile stress were used for the majority of the tests at each of the temperature levels tested, with selected tests performed as low as 30 percent of the ultimate tensile stress. Biaxial and uniaxial testing modes were used to evaluate the anisotropic deformation behavior. The combination of test results and predictive-analysis techniques developed as part of this program make it possible to predict the transient deformation of reactor fuel cladding during simulated LOCA conditions. Results include creep/creep rupture strain numerical constitutive relationships out to 120 seconds, computer codes and ramp test data.

  20. Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term

    Directory of Open Access Journals (Sweden)

    Sunarko

    2017-01-01

    Full Text Available Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2 PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region.

  1. Design of Test Facility to Evaluate Boric Acid Precipitation Following a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong-Kwan; Song, Yong-Jae [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The U.S.NRC has identified a concern that debris associated with generic safety issue (GSI) - 191 may affect the potential precipitation of boric acid due to one or more of the following phenomena: - Reducing mass transport (i.e. mixing) between the core and the lower plenum (should debris accumulate at the core inlet) - Reduced lower plenum volume (should debris settle in the lower plenum), and, - Increased potential for boric acid precipitation (BAP) in the core (should debris accumulate in suspension in the core) To address these BAP issues, KHNP is planning to conduct validation tests by constructing a BAP test facility. This paper describes the design of test facility to evaluate BAP following a LOCA. The design of BAP test facility has been developed by KHNP. To design the test facility, test requirements and success criteria were established, and scaling analysis of power-to-volume method, Ishii-Kataoka method, and hierarchical two-tiered method were investigated. The test section is composed of two fuel assemblies with half of full of prototypic FA height. All the fuel rods are heated by the electric power supplier. The BAP tests in the presence of debris, buffering agents, and boron will be performed following the test matrix.

  2. A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

    Directory of Open Access Journals (Sweden)

    Jae-Yong Kim

    2014-01-01

    Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

  3. Effects of post-LOCA conditions on a protective coating (paint) for the Nuclear Power Industry

    International Nuclear Information System (INIS)

    Loyola, V.M.; Womelsduff, J.E.

    1985-03-01

    When corrosion protection of steel cannot be achieved by galvanizing due to size, use, or other restrictions, the steel is frequently protected by the application of a suitable corrosion-inhibiting paint. A widely accepted corrosion inhibiting coating is one in which finely powdered zinc metal is dispersed in an organic polymer matrix and applied to steel as a paint. This system is often used with a non-zinc bearing topcoat for enhanced protection. We have studied the oxidation of zinc in a zinc-rich coating used in the nuclear power industry and have measured the rates of hydrogen generation from these coatings due to zinc oxidation at temperatures of up to 175 0 C. The results suggest that the real-time rates of hydrogen generation are considerably higher than previously believed. A second concern involves the generation of debris or solid reaction products which could cause plugging or fouling of the recirculation pumps, spray nozzles, and/or heat exchangers. Coatings are observed to fail at post-LOCA conditions which are well within the limits predicted by Design Basis Accident analysis. The failures involve cracking and/or delamination of the topcoat and production of solid corrosion products involving the zinc-rich primer. 22 refs., 10 figs., 6 tabs

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

    International Nuclear Information System (INIS)

    Cartmill, C.E.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Buckalew, W.H.

    1989-07-01

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

  6. D3.5 Formalized stepwise approach for implementing logistical concepts using BeWhere and LocaGIStics

    NARCIS (Netherlands)

    Annevelink, E.; Elbersen, B.; Leduc, S.; Staritsky, I.G.

    2016-01-01

    This deliverable describes a formaliz
    logistical concepts in the practical
    chains and for assessing thei
    BeWhere and LocaGIStics. It describes
    these two logistical assessment tools
    interlinked so that LocaGIStics can further refine and detail the outcomes of the
    BeWhere model

  7. Symmetry-breaking analysis for the general Helmholtz-Duffing oscillator

    International Nuclear Information System (INIS)

    Cao Hongjun; Seoane, Jesus M.; Sanjuan, Miguel A.F.

    2007-01-01

    The symmetry breaking phenomenon for a general Helmholtz-Duffing oscillator as a function of a symmetric parameter in the nonlinear force is investigated. Different values of this parameter convert the general oscillator into either the Helmholtz or the Duffing oscillator. Due to the variation of the symmetric parameter, the phase space patterns of the unperturbed Helmholtz-Duffing oscillator will cause a huge difference between the left-hand homoclinic orbit and the right-hand one. In particular, the area of the left-hand homoclinic orbits is a strictly monotonously decreasing function, while the area of the right-hand homoclinic orbit varies only in a very small range. There exist distinct local supercritical and subcritical saddle-node bifurcations at two different centers. The left-hand and the right-hand existing regions of the harmonic solutions of the Helmholtz-Duffing oscillator created by the left-hand and the right-hand saddle-node bifurcation curves will lead to different transition in the amplitude-frequency plane. There exists also a critical frequency which has the effect that the left-hand homoclinic bifurcation value is equal to the right-hand homoclinic bifurcation value. And, if the amplitude coefficient of the Helmholtz-Duffing oscillator is used as the control parameter, and it is larger than the same left-hand and right-hand homoclinic bifurcation, then the global stability of the system will be destroyed at a lowest cost. Besides this critical frequency, the left-hand and the right-hand homoclinic bifurcations are not only unequal, but also their effects for the system's stability are different. Among them, the effect resulting from the small homoclinic bifurcation for the system's stability is local and negligible, while the effect from the large homoclinic bifurcation is global but this is accomplished at a quite larger cost

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

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Joo S.; Diamond, David

    2016-12-06

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

  9. 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

  10. Analysis of in vivo and in vitro DNA strand breaks from trihalomethane exposure

    Directory of Open Access Journals (Sweden)

    DeAngelo Anthony

    2004-01-01

    Full Text Available Abstract Background Epidemiological studies have linked the consumption of chlorinated surface waters to an increased risk of two major causes of human mortality, colorectal and bladder cancer. Trihalomethanes (THMs are by-products formed when chlorine is used to disinfect drinking water. The purpose of this study was to examine the ability of the THMs, trichloromethane (TCM, bromodichloromethane (BDCM, dibromochloromethane (DBCM, and tribromomethane (TBM, to induce DNA strand breaks (SB in (1 CCRF-CEM human lymphoblastic leukemia cells, (2 primary rat hepatocytes (PRH exposed in vitro, and (3 rats exposed by gavage or drinking water. Methods DNA SB were measured by the DNA alkaline unwinding assay (DAUA. CCRF-CEM cells were exposed to individual THMs for 2 hr. Half of the cells were immediately analyzed for DNA SB and half were transferred into fresh culture medium and incubated for an additional 22 hr before testing for DNA SB. PRH were exposed to individual THMs for 4 hr then assayed for DNA SB. F344/N rats were exposed to individual THMs for 4 hr, 2 weeks, and to BDCM for 5 wk then tested for DNA SB. Results CCRF-CEM cells exposed to 5- or 10-mM brominated THMs for 2 hr produced DNA SB. The order of activity was TBM>DBCM>BDCM; TCM was inactive. Following a 22-hr recovery period, all groups had fewer SB except 10-mM DBCM and 1-mM TBM. CCRF-CEM cells were found to be positive for the GSTT1-1 gene, however no activity was detected. No DNA SB, unassociated with cytotoxicity, were observed in PRH or F344/N rats exposed to individual THMs. Conclusion CCRF-CEM cells exposed to the brominated THMs at 5 or 10 mM for 2 hr showed a significant increase in DNA SB when compared to control cells. Additionally, CCRF-CEM cells exposed to DBCM and TBM appeared to have compromised DNA repair capacity as demonstrated by an increased amount of DNA SB at 22 hr following exposure. CCRF-CEM cells were found to be positive for the GSTT1-1 gene, however no activity

  11. 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)

    Borges, Eduardo M.; Sabundjian, Gaiane

    2015-01-01

    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 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)

  12. 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)

  13. 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.

  14. 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.

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

    International Nuclear Information System (INIS)

    Rosso, Ricardo D.; Ventura, Mirta A.

    2002-01-01

    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 cm 2 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 T HPI 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 T HPI increase, the conditions suitable for PTS occurrence in a LOCA accident tend to diminish. The maximum value to the T HPI may be fixed from the maximum temperature allowed to preserve the structural integrity of the fuel cladding. (author)

  16. Realistic assessment of break size in simulated BWR loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Andersen, J.G.M.; Heck, C.L.; Klebanov, L.A.; Shiralkar, B.S.

    2009-01-01

    Realistic loss-of-coolant accident (LOCA) licensing methodologies offer an alternative to conservative methodologies that have been used historically. Less realistic analyses based on conservative assumptions and/or inputs may potentially misrepresent or conceal complicated interactions between competing phenomenological processes. The calculated results from the historical methods for the key LOCA licensing parameters including the most important parameter peak clad temperature (PCT) are generally expected to be conservative but may be non-conservative for some scenarios. In contrast, realistic methodologies attempt to model all important phenomena without making any grossly conservative assumptions that will distort the results. The paper demonstrates that realistic calculations must accurately account for the complex balance between the heat production and heat removal processes. Specific elements such as plant and bundle design, plant initial conditions and boundary conditions, and details of the LOCA scenario that include equipment availability and break size and location influence the balances between competing processes and thereby determine the calculated value of the PCT. For example, the break size and location influences the depressurization rate, the distribution of fluid inventory, and the timing of flashing in different parts of the system. The Automatic Depressurization System (ADS) also affects the depressurization rate (especially for small breaks). The Emergency Core Cooling System (ECCS) triggered by depressurization delivers fluid to the core to make up for lost inventory. The paper illustrates and discusses the relative importance of these phenomena and mechanisms, their timing and interactions, and their impact on the calculated PCT for different break sizes. Realistic calculations made with TRACG reveal that the maximum PCT for BWR plants with jet pumps occur at a smaller break size than the traditional design-basis accident (DBA) break size

  17. 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.

  18. Evolutionary analysis of tomato Sw-5 resistance-breaking isolates of Tomato spotted wilt virus.

    Science.gov (United States)

    López, Carmelo; Aramburu, José; Galipienso, Luis; Soler, Salvador; Nuez, Fernando; Rubio, Luis

    2011-01-01

    Tomato spotted wilt virus (TSWV) causes severe economic losses in many crops worldwide and often overcomes resistant cultivars used for disease control. Comparison of nucleotide and amino acid sequences suggested that tomato resistance conferred by the gene Sw-5 can be overcome by the amino acid substitution C to Y at position 118 (C118Y) or T120N in the TSWV movement protein, NSm. Phylogenetic analysis revealed that substitution C118Y has occurred independently three times in the studied isolates by convergent evolution, whereas the substitution T120N was a unique event. Analysis of rates of non-synonymous and synonymous changes at individual codons showed that substitution C118Y was positively selected.

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

    International Nuclear Information System (INIS)

    1996-01-01

    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

  20. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    Energy Technology Data Exchange (ETDEWEB)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described.

  1. CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA

    International Nuclear Information System (INIS)

    Hargroves, D.W.; Metcalfe, L.J.; Cheng, Teh-Chin; Wheat, L.L.; Mings, W.J.

    1991-01-01

    1 - Description of problem or function: CONTEMPT-LT was developed to predict the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided for fan cooler and cooling spray engineered safety systems. One to four compartments can be modeled, and any compartment except the reactor system may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. The user determines the compartments to be used, specifies input mass and energy additions, defines heat structure and leakage systems, and prescribes the time advancement and output control. CONTEMPT-LT/28-H (NESC0433/08) includes also models for hydrogen combustion. 2 - Method of solution: The initial conditions of the containment atmosphere are calculated from input values, and the initial temperature distributions through the containment structures are determined from the steady-state solution of the heat conduction equations. A time advancement proceeds as follows. The input water and energy rates are evaluated at the midpoint of a time interval and added to the containment system. Pressure suppression, spray system effects, and fan cooler effects are calculated using conditions at the beginning of a time-step. Leakage and heat losses or gains, extrapolated from the last time-step, are added to the containment system. Containment volume pressure and temperature are estimated by solving the mass, volume, and energy balance equations. Using these results as boundary conditions, the heat conduction equations

  2. Analysis of homogeneity of mixed oxide (MOX) fuel surface by laser induced break down spectroscopy (LIBS)

    International Nuclear Information System (INIS)

    Detalle, V.; Lacour, J.L.; Mauchien, P.; Wagner, J.F.

    2000-01-01

    In the nuclear fuel cycle, plutonium is recycled to produce MOX fuel for use in PWR reactors. Surface inspections of mixed PuO 2 /UO 2 pellets is very important in the process. The performance of Laser Induced Breakdown Spectroscopy (LIBS) or Laser Ablation Optical Emission Spectroscopy (LA-OES) was therefore assessed for surface analysis of simulated MOX pellets containing a mixture of UO 2 and CeO 2 , and compared with results obtained with the standard Castaing microprobe analyzer technique. In LIBS, a material is ablated by focussing a laser beam, and the emission from neutral and ionized atoms can be used to determine the composition. An original experimental set up was developed to obtain a LIBS microprobe system for microanalysis of the sample surface. The instrument has three main components : a laser (quadrupled Nd YAG), an ablation head (using a microscope lens) and a detection unit (spectrometer combined with an ICCD Intensified Charge Coupled Device camera). The LIBS technique has well-known advantages particularly for nuclear applications: (1) it requires no sample preparation ; (2) only a small amount of material (craters 7 or 3 μm wide and 1 to 3 μm deep) is needed for the analysis; (3) analysis can be performed remotely via optical fiber, allowing measurements in a hostile environment and at atmospheric pressure. The experimental set-up developed demonstrated that the LIBS microprobe system can be used for surface analysis of UO 2 /CeO 2 pellets. Figure 1 shows the calibration curve obtained, with the Ce/U ratio versus Ce/U concentration. Good linearity was found and a relative standard deviation of 5 % was determined for 100 single shots. A qualitative comparison of the LIBS microanalysis set up and the Castaing microprobe analyzer is shown in Figure 2, which reveals the same features. Thus, both analytical techniques can identify the surface non-homogeneity of the pellet. While LIBS is destructive, it requires no sample preparation, is faster (2

  3. High-performance analysis of single interphase cells with custom DNA probes spanning translocation break points

    Science.gov (United States)

    Weier, Heinz-Ulli G.; Munne, S.; Lersch, Robert A.; Marquez, C.; Wu, J.; Pedersen, Roger A.; Fung, Jingly

    1999-06-01

    The chromatin organization of interphase cell nuclei, albeit an object of intense investigation, is only poorly understood. In the past, this has hampered the cytogenetic analysis of tissues derived from specimens where only few cells were actively proliferating or a significant number of metaphase cells could be obtained by induction of growth. Typical examples of such hard to analyze cell systems are solid tumors, germ cells and, to a certain extent, fetal cells such as amniocytes, blastomeres or cytotrophoblasts. Balanced reciprocal translocations that do not disrupt essential genes and thus do not led to disease symptoms exit in less than one percent of the general population. Since the presence of translocations interferes with homologue pairing in meiosis, many of these individuals experience problems in their reproduction, such as reduced fertility, infertility or a history of spontaneous abortions. The majority of translocation carriers enrolled in our in vitro fertilization (IVF) programs carry simple translocations involving only two autosomes. While most translocations are relatively easy to spot in metaphase cells, the majority of cells biopsied from embryos produced by IVF are in interphase and thus unsuitable for analysis by chromosome banding or FISH-painting. We therefore set out to analyze single interphase cells for presence or absence of specific translocations. Our assay, based on fluorescence in situ hybridization (FISH) of breakpoint-spanning DNA probes, detects translocations in interphase by visual microscopic inspection of hybridization domains. Probes are prepared so that they span a breakpoint and cover several hundred kb of DNA adjacent to the breakpoint. On normal chromosomes, such probes label a contiguous stretch of DNA and produce a single hybridization domain per chromosome in interphase cells. The translocation disrupts the hybridization domain and the resulting two fragments appear as physically separated hybridization domains in

  4. Dynamic analysis of elastic rubber tired car wheel breaking under variable normal load

    Science.gov (United States)

    Fedotov, A. I.; Zedgenizov, V. G.; Ovchinnikova, N. I.

    2017-10-01

    The purpose of the paper is to analyze the dynamics of the braking of the wheel under normal load variations. The paper uses a mathematical simulation method according to which the calculation model of an object as a mechanical system is associated with a dynamically equivalent schematic structure of the automatic control. Transfer function tool analyzing structural and technical characteristics of an object as well as force disturbances were used. It was proved that the analysis of dynamic characteristics of the wheel subjected to external force disturbances has to take into account amplitude and phase-frequency characteristics. Normal load variations impact car wheel braking subjected to disturbances. The closer slip to the critical point is, the higher the impact is. In the super-critical area, load variations cause fast wheel blocking.

  5. Shorter lunch breaks lead secondary-school students to make less healthy dietary choices: multilevel analysis of cross-sectional national survey data.

    Science.gov (United States)

    Townsend, Nicholas

    2015-06-01

    At the time of the study a number of schools within Wales had shortened the amount of time they allow for lunch break. The study investigated the association between length of lunch break and the dietary choices of students in secondary schools. Student-level data, collected through anonymised questionnaires, included reported dietary choices and correlates of these; data on school approaches to food were collected through postal surveys. Multilevel analysis was used to study the independent association between lunch-break length and student dietary choice. Data were collected from secondary schools in Wales that were part of the 2005/2006 Health Behaviour in School-aged Children (HBSC) study. The final sample for analysis included data from 6693 students aged 11-16 years and 289 teachers from sixty-four secondary schools in Wales. Once controlling for many individual-level and school-level factors, the length of time allowed for lunch across the range for schools included in the study (minimum =25 min, maximum =62.5 min) was associated with higher odds of students eating fruit for lunch (2.20; 95% CI 1.18, 4.11) and fruit and vegetables on a daily basis (2.15; 95% CI 1.33, 3.47) but lower odds of eating unhealthy foods on a daily basis (0.44; 95% CI 0.24, 0.80). Shorter lunch breaks are associated with less healthy dietary choices by students. Schools should consider the impact that lunch-break length has on student dietary choice as well as on other behaviours. Policy makers should work with schools in encouraging them to maintain lunch breaks of a length that allow pupils to make healthy choices.

  6. Analysis of temperature data over semi-arid Botswana: trends and break points

    Science.gov (United States)

    Mphale, Kgakgamatso; Adedoyin, Akintayo; Nkoni, Godiraone; Ramaphane, Galebonwe; Wiston, Modise; Chimidza, Oyapo

    2017-06-01

    Climate change is a global challenge which impacts negatively on sustainable rural livelihoods, public health and economic development, more especially for communities in Southern Africa. Assessment of indices that signify climate change can inform formulation of relevant adaptation strategies and policies for the communities. Diurnal temperature range (DTR) is acknowledged as an expedient measure of the scourge as it is sensitive to variations in radiative energy balance. In this study, a long-term (1961-2010) daily temperature data obtained from nine (9) synoptic stations in Botswana were analyzed for monotonic trends and epochal changes in annual maximum (T max), minimum (T min) temperatures and DTR time series. Most of the considered stations were along the Kalahari Transect, a region which is at high risk of extensive environmental change due to climate change. Mann-Kendall trend and Lepage tests were applied for trend and change point analysis, respectively. The statistical analysis shows that stations in the southern part of the country experienced significant negative trends in annual DTR at the rate of -0.09 to -0.30 °C per decade due to steeper warming rates in annual T min than annual T max trends. On the contrary, stations in the northern part of the country experienced positive trends in annual DTR brought about by either a decreasing annual T min trend which outstripped annual T max or annual T max which outpaced annual T min. The increasing trends in DTR varied from 0.25 to 0.67 °C per decade. For most of the stations, the most significant annual DTR trends change point was in 1982 which coincided with the reversal of atmospheric circulation patterns.

  7. SB LOCA thermal-hydraulic analyses for Krsko Full Scope Simulator validation

    International Nuclear Information System (INIS)

    Prosek, A.; Parzer, I.

    2005-01-01

    The Krsko nuclear power plant (NPP), which is a two-loop pressurized water reactor, Westinghouse type, before modernization in 2000 obtained plant specific full scope simulator. The purpose of the presented analyses was to perform Small Break Loss of Coolant Accident (SBLOCA) reference calculations for KFSS validation in 2004. In addition, the thermal-hydraulic response of the reactor coolant system (RCS) was studied in detail. For the thermal-hydraulic analysis the RELAP5/MOD3.3 code and input model delivered from Krsko NPP were used. The RELAP5 calculated reference results showed that the plant system response to breaks with small break area is slower compared to breaks with larger break area. The comparison of the KFSS data with calculated results suggest that the simulator validation testing in the year 2004 for this kind of accident was successful. Nevertheless, when comparing the physical phenomena and processes, the RELAP5/MOD3.3 predicted smaller core uncovery compared to the KFSS measurement. One reason is different core cycles. Finally, this finding suggests that even for simulator reference calculations the quantification of model uncertainties would be useful. (author)

  8. Breaking the Carnot limit without violating the second law: A thermodynamic analysis of off-resonant quantum light generation

    Science.gov (United States)

    Boukobza, E.; Ritsch, H.

    2013-06-01

    The Carnot limit, formulated in 1824, represents the maximal efficiency of a classical heat engine. In this work we present a thermodynamical analysis of a light amplifier based on a three-level atom coupled off-resonantly to a single quantized cavity mode and to two heat reservoirs with positive temperatures. Based on standard work and heat flow equilibrium, we show that for a cavity blue-detuned with respect to the atomic resonance, the system can surpass the Carnot limit. Nevertheless, the second law of thermodynamics is still obeyed, as the total entropy always increases. By analyzing a semiclassical version of the model, we derive a formula for the critical frequency for which the Carnot limit is broken and a formula for the amplifier efficiency which agrees with its quantum counterpart. In the semiclassical regime, however, the second law is not satisfied and hence it does not offer a physically acceptable description of the system. Finally, we show that breaking the Carnot limit occurs also in a blue-detuned quantum amplifier with output coupling, which represents a realistic model of a laser or maser.

  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)

    Petruzzi, A.; D'Auria, F.; Crecy, Agnes de; Bazin, P.; Borisov, S.; Skorek, T.; Glaeser, H.; Benoit, J. P.; Chojnacki, E.; Fujioka, K.; Inoue, S.; Chung, B.D.; Trosztel, I.; Toth, I.; Oh, D. Y.; Pernica, R.; Kyncl, M.; Macek, J.; Macian, R.; Tanker, E.; Soyer, A. E.; Ozdere, O.; Perez, M.; Reventos, F.

    2005-11-01

    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. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  11. Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

    2018-01-01

    Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

  12. Development of a multi-dimensional two phase thermal hydraulic analysis code, MEDUSA, and its application to feed line break analysis

    International Nuclear Information System (INIS)

    Park, Chan Eok; Lee, Sang Il; Lee, Sang Yong; Kim, Shin Whan; Seo, Jong Tae

    2004-01-01

    A general purpose multi-dimensional two phase thermal hydraulic analysis code, MEDUSA has been developed utilizing two fluid three field governing equations. The flexible noding system of MEDUSA allows users to model the wide variety of geometry encountered in nuclear reactor system. The specific models of reactor kinetics, pump, valves, and separators has been also developed to simulate the core power and various hydraulic components of primary or secondary systems of nuclear power plant. In this paper, the main feature of the MEDUSA code is briefly described, and its application results to feed line break(FLB) accident of Korean Standard Nuclear Power Plant(KSNP) are compared with those of the licensing analysis code, CESEC-III

  13. In-core LOCA-s: analytical solution for the delayed mixing model for moderator poison concentration

    International Nuclear Information System (INIS)

    Firla, A.P.

    1995-01-01

    Solutions to dynamic moderator poison concentration model with delayed mixing under single pressure tube / calandria tube rupture scenario are discussed. Such a model is described by a delay differential equation, and for such equations the standard ways of solution are not directly applicable. In the paper an exact, direct time-domain analytical solution to the delayed mixing model is presented and discussed. The obtained solution has a 'marching' form and is easy to calculate numerically. Results of the numerical calculations based on the analytical solution indicate that for the expected range of mixing times the existing uniform mixing model is a good representation of the moderator poison mixing process for single PT/CT breaks. However, for postulated multi-pipe breaks ( which is very unlikely to occur ) the uniform mixing model is not adequate any more; at the same time an 'approximate' solution based on Laplace transform significantly overpredicts the rate of poison concentration decrease, resulting in excessive increase in the moderator dilution factor. In this situation the true, analytical solution must be used. The analytical solution presented in the paper may also serve as a bench-mark test for the accuracy of the existing poison mixing models. Moreover, because of the existing oscillatory tendency of the solution, special care must be taken in using delay differential models in other applications. (author). 3 refs., 3 tabs., 8 figs

  14. Modelling the association of dengue fever cases with temperature and relative humidity in Jeddah, Saudi Arabia-A generalised linear model with break-point analysis.

    Science.gov (United States)

    Alkhaldy, Ibrahim

    2017-04-01

    The aim of this study was to examine the role of environmental factors in the temporal distribution of dengue fever in Jeddah, Saudi Arabia. The relationship between dengue fever cases and climatic factors such as relative humidity and temperature was investigated during 2006-2009 to determine whether there is any relationship between dengue fever cases and climatic parameters in Jeddah City, Saudi Arabia. A generalised linear model (GLM) with a break-point was used to determine how different levels of temperature and relative humidity affected the distribution of the number of cases of dengue fever. Break-point analysis was performed to modelled the effect before and after a break-point (change point) in the explanatory parameters under various scenarios. Akaike information criterion (AIC) and cross validation (CV) were used to assess the performance of the models. The results showed that maximum temperature and mean relative humidity are most probably the better predictors of the number of dengue fever cases in Jeddah. In this study three scenarios were modelled: no time lag, 1-week lag and 2-weeks lag. Among these scenarios, the 1-week lag model using mean relative humidity as an explanatory variable showed better performance. This study showed a clear relationship between the meteorological variables and the number of dengue fever cases in Jeddah. The results also demonstrated that meteorological variables can be successfully used to estimate the number of dengue fever cases for a given period of time. Break-point analysis provides further insight into the association between meteorological parameters and dengue fever cases by dividing the meteorological parameters into certain break-points. Copyright © 2016 Elsevier B.V. All rights reserved.

  15. Política habitacional e locação social em Salvador

    Directory of Open Access Journals (Sweden)

    Nelson Baltrusis

    Full Text Available Este artigo tem como objetivo analisar o mercado imobiliário de locação em Salvador. Num primeiro momento, caracterizaremos o problema habitacional em Salvador, para o que nos apoiaremos nas diretrizes e ações previstas no Plano Municipal de Habitação de Interesse Social (PMHIS. Em seguida, trataremos das políticas implantadas pelos governos do estado e federal, destacando a experiência do Programa de Arrendamento Residencial (PAR e incorporando algumas considerações sobre o Programa Minha Casa, Minha Vida. Também será abordada a questão do mercado de locação em Salvador a partir do perfil de moradores e da dinâmica do mercado.

  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 investigation on the causes for pellet fragmentation under LOCA conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bianco, A., E-mail: andrea.bianco@eon.com [E.ON Kernkraft GmbH, Tresckowstraße 5, 30457 Hannover (Germany); Technische Universität München, Lehrstuhl für Nukleartechnik, Boltzmannstr. 15, D-85748 Garching (Germany); Vitanza, C. [Institutt For Energiteknikk, Os Alle 5, NO-1777 Halden (Norway); Seidl, M. [E.ON Kernkraft GmbH, Tresckowstraße 5, 30457 Hannover (Germany); Technische Universität München, Lehrstuhl für Nukleartechnik, Boltzmannstr. 15, D-85748 Garching (Germany); Wensauer, A.; Faber, W. [E.ON Kernkraft GmbH, Tresckowstraße 5, 30457 Hannover (Germany); Macián-Juan, R. [Technische Universität München, Lehrstuhl für Nukleartechnik, Boltzmannstr. 15, D-85748 Garching (Germany)

    2015-10-15

    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. - Highlights: • LOCA separate effects test performed in the hot cell of the Halden Reactor Project. • Cladding ballooning enhances fuel pellet cracks and fragmentation. • The occurrence of burst enhances fuel pellet cracks and fragmentation. • Cladding ballooning and burst increase the transient fission gas release.

  18. Effective water cooling of very hot surfaces during the LOCA accident.

    Czech Academy of Sciences Publication Activity Database

    Štepánek, J.; Bláha, V.; Dostál, V.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 1211-1214 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : LOCA * Quenching * Divertor cooling * Heat transfer * Rewetting Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617303733

  19. Simulating a partial LOCA in a narrow channel using the DSNP simulating system

    International Nuclear Information System (INIS)

    Saphier, D.

    2007-01-01

    A partial LOCA accident in a pool type research reactor was investigated. A new MTR type fuel channel model for the DSNP simulation system was developed; permitting detailed axial and radial temperature distribution. New and older heat transfer correlations were incorporated in the model. Simulation for water levels of 14 and 35 cm in a 62 cm channel were performed. The resulting maximum temperatures remain significantly below the aluminium melting point, and no damage to the core will take place under these conditions

  20. Simulation of the effects of the extend fuel rod burn-up under LOCA scenario

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira; Abe, Alfredo, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br, E-mail: ayabe@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Due to the high burn-up imposed to the nuclear fuel in the last recent years, new challenges become important, including a deep review of the fuel performance under accident conditions. In this sense, available data in the open literature show that some experiments were carried out in order to study the behavior of fuel rods under LOCA (Loss of Coolant Accident) scenario. For instance, a series of experiments, designated IFA-650 series, performed in the Halden reactor in 2010 present data related to zircaloy fuel rods submitted to LOCA conditions. In the tests were addressed issues such as fuel fragmentation, relocation and dispersal for an extended irradiation cycle. In the studied case (IFA-650.5), the LOCA scenario was evaluated after a burnup of 83.4 MWd/kg. The aim of this paper is to compare the experimental data to the fuel performance obtained applying the codes FRAPCON and FRAPTRAN. Different phenomena were evaluated, such as ballooning, burst, cladding oxidation and fuel relocation. Also, the cladding metallurgical phase transformation was considered. The obtained results reproduced in a good way the experimental data, showing that the adopted models are representative of the observed phenomena. (author)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  2. A new high temperature deformation model for Zircaloy clad ballooning under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Brzoska, B.; Cheliotis, G.; Kunick, A.; Senski, G.

    1977-01-01

    Assuming Zircaloy clad ballooning occurs predominantly by thermal activated secondary creep, generally a power law is applied to describe the creep rate analytically. According to Norton the creep rate is taken as a power function of the cladding hoop stress multiplied by a numerical constant which is determined by the cladding structural properties and a Boltzmann factor including the creep activation energy, the gas constant and the cladding temperature, respectively. As is well known, the stress exponent is not a constant value in the total range of LOCA stresses, but increases steadily with stress. This difficulty is avoided by introducing into the Norton law a plastic flow-factor including a limiting stress, which was derived by G. Senski using plastic crack models from Dugdale and Irwin. For LOCA applications the limiting stress is identified with the burst stress, which is experimentally determined. A total number of about 290 directly heated KWU burst tests including two types of experiments: (1) controlled temperature transient tests, (ii) creep rupture tests, are used to fit the burst stress of KWU Zircaloy tubes simulating the whole range of LOCA temperatures, heating rates and creep times. (Auth.)

  3. Test facility simulation of WWER fuel rods behaviour under initial stage of LB LOCA

    International Nuclear Information System (INIS)

    Deniskin, V.; Konstantinov, V.; Nalivaev, V.; Parshin, N.; Fedik, I.; Semishkin, V.; Shumsky, A.

    2003-01-01

    The calculation and experimental method has been developed to calculate and to study the possibility of simulating the main parameters at the initial stage of an accident of LB LOCA-type for WWER reactor on the experimental facility. They are velocities and the way of heating of FR claddings, the levels of cladding temperatures, the rate of the external pressure drop and the creation of the natural change of pressure on the claddings at the maximum temperature. Series of experimental tests of FR simulators of WWER was carried out at the initial stage of LB LOCA. They resulted in a potential possibility of the ballooning and the depressurization of maximally heated FRs even on this stage of the accident. The results of these tests allow to study the structural changes of FR materials and to enlarge the knowledge about deformation and possible depressurization of FR claddings. To study more carefully the problems of the ballooning and a possible depressurization at the initial stage of the accident it is necessary to continue the researches and to develop scenarios of this accident for FRs with different energy release levels and to test the initial stage of LB LOCA on the experimental facility and under reactor conditions

  4. Public health and pipe breaks in water distribution systems: analysis with internet search volume as a proxy.

    Science.gov (United States)

    Shortridge, Julie E; Guikema, Seth D

    2014-04-15

    Drinking water distribution infrastructure has been identified as a factor in waterborne disease outbreaks and improved understanding of the public health risks associated with distribution system failures has been identified as a priority area for research. Pipe breaks may pose a risk, as their occurrence and repair can result in low or negative pressure, potentially allowing contamination of drinking water from adjacent soils. However, measuring this phenomenon is challenging because the most likely health impact is mild gastrointestinal (GI) illness, which is unlikely to result in a doctor or hospital visit. Here we present a novel method that uses data mining techniques and internet search volume to assess the relationship between pipe breaks and symptoms of GI illness in two U.S. cities. Weekly search volume for the terms diarrhea and vomiting was used as the response variable with the number of pipe breaks in each city as a covariate as well as additional covariates to control for seasonal patterns, search volume persistence, and other sources of GI illness. The fit and predictive accuracy of multiple regression and data mining techniques were compared, with the best performance obtained using random forest and bagged regression tree models. Pipe breaks were found to be an important and positively correlated predictor of internet search volume in multiple models in both cities, supporting previous investigations that indicated an increased risk of GI illness from distribution system disturbances. Copyright © 2014 Elsevier Ltd. All rights reserved.

  5. Analysis of the effects of postulated pipe breaks on the LOFT containment building and on building TAN 650

    Energy Technology Data Exchange (ETDEWEB)

    Mosby, W.R.

    1978-08-31

    This report presents the results of analyses of the consequences of pipe whips and jets occurring as a result of pipe breaks postulated to occur in the LOFT main steam and feedwater lines both inside and adjacent to the LOFT Containment Vessel and Building TAN 650. Pipe whip and jet cases resulting in breach of containment or damage to Building TAN 650 are identified.

  6. Containment parametric analysis for loss of coolant accident

    International Nuclear Information System (INIS)

    Fabjan, L.

    1985-01-01

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

  7. Experimental analysis of the power curve sensitivity test series at ROSA-III

    International Nuclear Information System (INIS)

    Koizumi, Y.; Iriko, M.; Yonomoto, T.; Tasaka, K.

    1985-01-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral LOCA and ECCS tests. Seven recirculation pump suction line break LOCA experiments were conducted at the ROSA-III facility in order to examine the effect of the initial stored heat of a fuel rod on the peak cladding temperature (PCT). The break size was changed from 200% to 5% in the test series and a failure of a high pressure core spray (HPCS) diesel generator was assumed. Three power curves which represented conservative, realistic and zero initial stored heat, respectively, were used. In a large break LOCA such as 200% or 50% breaks, the initial stored heat in a fuel rod has a large effect on the cladding surface temperature because core uncovery occurs before all the initial stored heat is released, whereas in a small break LOCA such as a 5% break little effect is observed because core uncovery occurs after the initial stored heat is released. The maximum PCTs for the conservative initial stored heat case was 925 K, obtained in the 50% break experiment, and that for the realistic initial stored heat case was 835 K, obtained in the 5% break experiment. (orig./HP)

  8. Uncertainty analysis of the FRAP code

    International Nuclear Information System (INIS)

    Peck, S.O.

    1978-01-01

    A user oriented, automated uncertainty analysis capability has been built into the FRAP code (Fuel Rod Analysis Program) and applied to a PWR fuel rod undergoing a LOCA. The method of uncertainty analysis is the Response Surface Method (RSM). (author)

  9. Break-even analysis of costs for controlling Toxoplasma gondii infections in slaughter pigs via a serological surveillance program in the Netherlands.

    Science.gov (United States)

    van Asseldonk, M; van Wagenberg, C P A; Wisselink, H J

    2017-03-01

    Toxoplasma gondii (T. gondii) is a food safety hazard which causes a substantial human disease burden and cost-of-illness. Infected pig meat is a common source of toxoplasmosis. A break-even analysis was conducted to estimate the point for which the intervention cost at fattening pig farms equaled the cost of averted human disease burden and cost-of-illness minus the costs of a T. gondii surveillance program. The surveillance program comprised serological testing of blood samples taken at slaughter. Break-even points were determined given alternative levels of the effectiveness of the intervention program (10% up to 90% in steps of 10%), the value of an averted DALY (20,000, 50,000 and 80,000 Euro), and threshold of sample prevalence for a farm to be under intervention (5% up to 50% out of 20 samples in steps of 5%). Since test characteristics are a determining factor in the break-even analysis, and literature is inconclusive concerning sensitivity (se) and specificity (sp) of the serological test kit used, two alternative sets of assumptions were analysed. The estimated maximum costs of an intervention if only benefits for domestic consumers were accounted amounted approximately 2981 Euro (se=98.9% and sp=92.7%) versus 4389 Euro (se=65.2% and sp=97.4%) per year per fattening pig farm under intervention assuming an effectiveness of 50%, 50,000 Euro per averted DALY and threshold T. gondii sample prevalence of 5% for a farm to be under intervention. Since almost 80% of the gross domestic production is exported corresponding break-even values increased up to 12,034 Euro and 18,366 Euro if benefits for consumers abroad were included as well. Empirical research to strengthen the knowledge about the efficacy of a farm intervention measures is recommended. Copyright © 2017 Elsevier B.V. All rights reserved.

  10. 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. (c) 2016 APA, all rights reserved).

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

    International Nuclear Information System (INIS)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

    1993-01-01

    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)

  12. Failure Mode and Effects Analysis (FMEA) of the Emergency Core Cooling System (ECCS) for a Westinghouse type 312, three loop pressurized water reactor

    International Nuclear Information System (INIS)

    Shopsky, W.E.

    1977-01-01

    The Emergency Core Cooling System (ECCS) is a Safeguards System designed to cool the core in the unlikely event of a Loss-of-Coolant Accident (LOCA) in the primary reactor coolant system as well as to provide additional shutdown capability following a steam break accident. The system is designed for a high reliability of providing emergency coolant and shutdown reactivity to the core for all anticipated occurrences of such accidents. The ECCS by performing its intended function assures that fuel and clad damage is minimized during accident conditions thus reducing release of fission products from the fuel. The ECCS is designed to perform its function despite sustaining a single failure by the judicious use of equipment and flow path redundancy within and outside the containment structure. By the use of an analytic tool, a Failure Mode and Effects Analysis (FMEA), it is shown that the ECCS is in compliance with the Single Failure Criterion established for active failures of fluid systems during short and long term cooling of the reactor core following a LOCA or steam break accident. An analysis was also performed with regards to passive failure of ECCS components during long-term cooling of the core following an accident. The design of the ECCS was verified as being able to tolerate a single passive failure during long-term cooling of the reactor core following an accident. The FMEA conducted qualitatively demonstrates the reliability of the ECCS (concerning active components) to perform its intended safety function

  13. 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

  14. Cooling the intact loop of primary heat transport system using shut down cooling system after events such as LOCA

    International Nuclear Information System (INIS)

    Icleanu, D.L.

    2015-01-01

    The purpose of this paper is to model the Shutdown Cooling System operation for CANDU 6 NPP in case of LOCA accident, using Flowmaster calculation code by delimiting models and setting calculation assumptions and input data for hydraulic analysis, and and assumptions for the calculation and input data for calculating thermal performance check heat exchangers that are part of this system. The Flowmaster V7.8 code provides system engineers with a powerful tool to investigate pressure surge, pressure drop, flow rate, temperature and system response times - removing the uncertainty from fluid flow systems. Flowmaster is a one-dimensional thermal-hydraulic calculation code for dimensioning, analyzing and verifying the pipeline systems operation. Each component of Flowmaster is a mathematical model for an equipment that is included in a facility. Selected components are connected via nodes in order to form a network, which constitutes a computerized model of the system. Analyzing the parameters of the cooling system for all cooling processes considered it was found that the values obtained for thermal-hydraulic parameters, as well as the duration up to reaching specified limits fall within the design values of the system. This document is made up of an abstract and the slides of the presentation

  15. A Theoretical Model for the Prediction of Siphon Breaking Phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Youngmin; Kim, Young-In; Seo, Jae-Kwang; Kim, Keung Koo; Yoon, Juhyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A siphon phenomenon or siphoning often refers to the movement of liquid from a higher elevation to a lower one through a tube in an inverted U shape (whose top is typically located above the liquid surface) under the action of gravity, and has been used in a variety of reallife applications such as a toilet bowl and a Greedy cup. However, liquid drainage due to siphoning sometimes needs to be prevented. For example, a siphon breaker, which is designed to limit the siphon effect by allowing the gas entrainment into a siphon line, is installed in order to maintain the pool water level above the reactor core when a loss of coolant accident (LOCA) occurs in an open-pool type research reactor. In this paper, we develop a theoretical model to predict the siphon breaking phenomenon. In this paper, a theoretical model to predict the siphon breaking phenomenon is developed. It is shown that the present model predicts well the fundamental features of the siphon breaking phenomenon and undershooting height.

  16. Evaluation of simulated-LOCA tests that produced large fuel cladding ballooning

    International Nuclear Information System (INIS)

    Powers, D.A.; Meyer, R.O.

    1979-02-01

    A description is given of the NRC review and evaluation of simulated-LOCA tests that produced large axially extended ballooing in Zircaloy fuel cladding. Technical summaries are presented on the likelihood of the transient that was used in the tests, the effects of temperature variations on strain localization, and the results of other similar experiments. It is concluded that (a) the large axially extended deformations were an artifact of the experimental technique, (b) current NRC licensing positions are not invalidated by this new information, and (c) no new research programs are needed to study this phenomenon

  17. Breaking bad news in the emergency department: a comparative analysis among residents, patients and family members' perceptions.

    Science.gov (United States)

    Toutin-Dias, Gabriela; Daglius-Dias, Roger; Scalabrini-Neto, Augusto

    2018-02-01

    Our main objective was to assess patient and family members' perception of bad news communication in the emergency department (ED) and compare these with physicians' perceptions. This is a cross-sectional study carried out at the ED of a tertiary teaching hospital. To compare physicians' and receivers' (patient and/or family member) perceptions, we created a survey based on the six attributes derived from the SPIKES protocol. The surveys were applied immediately after communication of bad news occurred in the ED. We analyzed agreement among participants using κ-statistics and the χ-test to compare proportions. A total of 73 bad news communication encounters were analyzed. The survey respondents were 73 physicians, 69 family members, and four patients. In general, there is a low level of agreement between physicians' and receivers' perceptions of how breaking bad news transpired. The satisfaction level of receivers, in terms of breaking bad news by doctors, presented a mean of 3.7±0.6 points. In contrast, the physicians' perception of the communication was worse (2.9±0.6 points), with P value less than 0.001. Doctors and receivers disagree in relation to what transpired throughout bad news communications. Discrepancies were more evident in issues involving emotion, invitation, and privacy. An important agreement between perceptions was found in technical and knowledge-related aspects of the communication.

  18. 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.

  19. 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.

  20. 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.

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

    International Nuclear Information System (INIS)

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

    1994-04-01

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

  2. Water volume available for ECCS sump recirculation mode following a LOCA

    International Nuclear Information System (INIS)

    Riekert, T.; Rebohm, H.; Huber, J.; Brandes, F.

    2006-01-01

    In this paper we describe the reviews performed in Germany on the water level in the containment sump after a LOCA and the derived actions. Our view on the issue is from the perspective of the independent safety experts - i.e. TUV SUD Industrie Service (TUV SUD IS), TUV SUD Energietechnik GmbH Baden-Wuerttemberg (TUV SUD ET), TUV NORD EnSys Hannover and TUV NORD SysTec -, which reviewed the analyses of the utilities on behalf of the responsible supervising authorities. Between these expert organizations information were exchanged via the steering committee on nuclear technology of the association of the TUVs (VdTUV). In our paper we describe the analyses on the two safety issues relevant in the connection with the water level in the containment sump: the necessary minimum coverage of suction pipes to avoid inadmissible entrainment of air and the water retention inside the containment after a LOCA. Our description concentrates on PWRs because of the more complex conditions in comparison to BWRs. In conclusion it can be stated that due to the thorough evaluation of operating experience, optimization measures could be derived. In addition, the analyses served the purpose of know-how maintenance. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vigil, R.A. [Science & Engineering Associates, Inc., Albuquerque, NM (United States); Jacobus, M.J. [Sandia National Labs., Albuquerque, NM (United States)

    1994-04-01

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

  4. 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.

  5. A main steam line break experiment at ROSA III - RUN 952

    International Nuclear Information System (INIS)

    Kawaji, Masahiro; Nakamura, Hideo; Suzuki, Mitsuhiro; Tasaka, Kanji; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Murata, Hideo; Shiba, Masayoshi

    1984-12-01

    This report presents the experimental data for RUN 952, a 100% Main Steam Line (MSL) break experiment performed at the ROSA-III test facility. The ROSA-III facility is a volumetrically scaled (1/424) system of a BWR/6 used for integral BWR LOCA simulation experiments. RUN 952 is a reference MSL break test performed with a 100% break upstream of the main steam isolation valve (MSIV) and full ECCS actuation logic. The MSL break is characterized by a relatively slower depressurization of the system due to a break flow of high mass quality, in comparison with a recirculation line break test, RUN 901. Continuous flashing of the fluid in the pressure vessel was observed and a slow decrease in the downcomer water level eventually led to actuation of HPCS but not LPCS and LPCI. About 2/3 of the core was uncovered, however, the coolant level recovered quickly following the HPCS injection. The peak cladding temperature reached was 752 K, which is 28 K lower than that obtained in RUN 901. (author)

  6. 'BREAKS' Protocol for Breaking Bad News.

    Science.gov (United States)

    Narayanan, Vijayakumar; Bista, Bibek; Koshy, Cheriyan

    2010-05-01

    Information that drastically alters the life world of the patient is termed as bad news. Conveying bad news is a skilled communication, and not at all easy. The amount of truth to be disclosed is subjective. A properly structured and well-orchestrated communication has a positive therapeutic effect. This is a process of negotiation between patient and physician, but physicians often find it difficult due to many reasons. They feel incompetent and are afraid of unleashing a negative reaction from the patient or their relatives. The physician is reminded of his or her own vulnerability to terminal illness, and find themselves powerless over emotional distress. Lack of sufficient training in breaking bad news is a handicap to most physicians and health care workers. Adherence to the principles of client-centered counseling is helpful in attaining this skill. Fundamental insight of the patient is exploited and the bad news is delivered in a structured manner, because the patient is the one who knows what is hurting him most and he is the one who knows how to move forward. Six-step SPIKES protocol is widely used for breaking bad news. In this paper, we put forward another six-step protocol, the BREAKS protocol as a systematic and easy communication strategy for breaking bad news. Development of competence in dealing with difficult situations has positive therapeutic outcome and is a professionally satisfying one.

  7. Perspectives on the application of order-statistics in best-estimate plus uncertainty nuclear safety analysis

    International Nuclear Information System (INIS)

    Martin, Robert P.; Nutt, William T.

    2011-01-01

    Research highlights: → Historical recitation on application of order-statistics models to nuclear power plant thermal-hydraulics safety analysis. → Interpretation of regulatory language regarding 10 CFR 50.46 reference to a 'high level of probability'. → Derivation and explanation of order-statistics-based evaluation methodologies considering multi-variate acceptance criteria. → Summary of order-statistics models and recommendations to the nuclear power plant thermal-hydraulics safety analysis community. - Abstract: The application of order-statistics in best-estimate plus uncertainty nuclear safety analysis has received a considerable amount of attention from methodology practitioners, regulators, and academia. At the root of the debate are two questions: (1) what is an appropriate quantitative interpretation of 'high level of probability' in regulatory language appearing in the LOCA rule, 10 CFR 50.46 and (2) how best to mathematically characterize the multi-variate case. An original derivation is offered to provide a quantitative basis for 'high level of probability.' At root of the second question is whether one should recognize a probability statement based on the tolerance region method of Wald and Guba, et al., for multi-variate problems, one explicitly based on the regulatory limits, best articulated in the Wallis-Nutt 'Testing Method', or something else entirely. This paper reviews the origins of the different positions, key assumptions, limitations, and relationship to addressing acceptance criteria. It presents a mathematical interpretation of the regulatory language, including a complete derivation of uni-variate order-statistics (as credited in AREVA's Realistic Large Break LOCA methodology) and extension to multi-variate situations. Lastly, it provides recommendations for LOCA applications, endorsing the 'Testing Method' and addressing acceptance methods allowing for limited sample failures.

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Reliability assessment and enhancement of pressure and differential pressure transmitter subjected to LOCA environment in nuclear power plants

    International Nuclear Information System (INIS)

    Kulkarni, R.D.; Bora, J.S.; Prakash, Ravi; Agarwal, Vivek; Sundersingh, V.P.

    2002-01-01

    Full text: In nuclear power plant, the safety and safety-related instrument viz. differential pressure transmitter is used for measurement of PHT pump room pressure to actuate containment isolation whereas pressure transmitter is used for monitoring PHT pressure to control emergency core cooling system (ECCS) actuation during LOCA condition. These instruments has to withstand the gamma radiation dose occurred during LOCA to maintain the safety as desired. The existing silicon devices in the signal processing circuit of these instruments are not qualified to work under the scenario of dosage due to LOCA event. Hence the alternative approaches like separating the transmitter sensor module from electronic PCB by using appropriate shielded cable, design of appropriate complete enclosure Igloo with lead as shield and seal to accommodate the transmitter, etc. has been worked out and subsequently the various experiments has been performed to find out the suitability of the schemes. The experimental results has been presented in the paper and the appropriate modifications in these schemes has been proposed to qualify these instrument for LOCA environment in the nuclear power plant. The suggested schemes enhances the overall reliability of the safety and safety-related equipment/ instruments in nuclear power plant

  10. Influence of hydrogen simulating burn-up effects on the metallurgical and thermal-mechanical behaviour of M5TM and zircaloy-4 alloys under LOCA conditions

    International Nuclear Information System (INIS)

    Mardon, J.P.; Brachet, J.C.; Portier, L.; Maillot, V.; Forgeron, T.; Lesbros, A.; Waeckel, N.

    2005-01-01

    layers) and their specific chemical composition - especially their oxygen content which is known to influence strongly the residual mechanical properties. Also, fractography analysis has been applied on failed samples to get a better knowledge of the failure mechanism as a function of the materials and of the hydrogen content, for different oxidation conditions. So, in the second and major part of the presentation, we will focus on LOCA post-quenched mechanical behavior of as-received and pre-hydrided Zircaloy-4 and M5 TM cladding tubes for typical hydrogen contents ranging from ∼100 up to ∼600 wt-ppm depending on the alloy. Ring compression, impact, and bending tests at room temperature and at 135 degree C have been performed for different oxidation conditions. The mechanical results will be presented and discussed vs. load, strain or energy, taking into account the metallurgical analysis (resultant phase morphology and thickness, chemical composition-oxygen contents, failure mode). (authors)

  11. In-pile experiments on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Karb, E.; Pruessmann, M.; Sepold, L.

    1980-05-01

    This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

  12. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    International Nuclear Information System (INIS)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-01-01

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  13. 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)

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

    International Nuclear Information System (INIS)

    Sweet, D.W.; Gibson, I.H.; Fell, J.

    1982-12-01

    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.)

  15. The analysis of PCCS heat removal performance for the 1300MWe simplified BWR

    Energy Technology Data Exchange (ETDEWEB)

    Yoshioka, Yuzuru [Japan Atomic Power Co., Tokyo (Japan); Arai, Kenji

    1997-12-31

    A passive containment cooling system (PCCS) is to remove the decay heat during and/or after LOCA. And in the 670MWe SBWR design 3 PCC units have been adopted. In order to study the number of PCC units for the 1300MWe SBWR, we carried out the containment response analysis by using TRAC, which has been qualified based on GIRAFFE tests. After we set 5 PCC units for 1300MWe SBWR plant, the TRAC performance has been analyzed for the main steam line break accident. The post-LOCA D/W pressure is maintained below the maximum allowable pressure (0.37MPa) for 3 days. The PCCS concept requires a lot of water in the top of reactor building. To allow the PCCS heat tubes uncovered is one of the effective method in order to reduce the pool volume. The target of reduced PCC pool size was set at 2/3 pool size of the present design. The number of the PCC heat exchanger was assumed to be 6 units by adding 1 unit. After that a TRAC analysis has been performed. Even with tubes uncovered, it is shown the degradation of PCCS performance is relatively small. The D/W peak pressure became about 0.36MPa below the maximum allowable pressure. (author)

  16. Chiral symmetry and chiral-symmetry breaking

    International Nuclear Information System (INIS)

    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

  17. LOCA, LOFA and LOVA analyses pertaining to NET/ITER safety design guidance

    International Nuclear Information System (INIS)

    Ebert, E.; Raeder, J.

    1991-01-01

    The analyses presented pertain to loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA). These types of accidents may jeopardise components and plasma vessel integrity and cause radioactivity mobilisation. The analyses reviewed have been performed under the assumption that the plasma facing components are protected by a carbon based armour. Accidental temperatures and pressure transients are quantified, the possibility of reaction products combustion is investigated and worst case accidental public doses are assessed. On this basis, design recommendations are given and design features such as low plasma facing components armour temperatures (on almost the entire surface) and inert gas adjacent to the vacuum vessel have been implemented. (orig.)

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  19. Prophylactic treatment of retinal breaks

    DEFF Research Database (Denmark)

    Blindbæk, Søren Leer; Grauslund, Jakob

    2015-01-01

    Prophylactic treatment of retinal breaks has been examined in several studies and reviews, but so far, no studies have successfully applied a systematic approach. In the present systematic review, we examined the need of follow-up after posterior vitreous detachment (PVD) - diagnosed by slit...... published before 2012. Four levels of screening identified 13 studies suitable for inclusion in this systematic review. No meta-analysis was conducted as no data suitable for statistical analysis were identified. In total, the initial examination after symptomatic PVD identified 85-95% of subsequent retinal......-47% of cases, respectively. The cumulated incidence of RRD despite prophylactic treatment was 2.1-8.8%. The findings in this review suggest that follow-up after symptomatic PVD is only necessary in cases of incomplete retinal examination at presentation. Prophylactic treatment of symptomatic retinal breaks...

  20. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    Energy Technology Data Exchange (ETDEWEB)

    Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  1. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    International Nuclear Information System (INIS)

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

    2017-01-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  2. The applicability of CFD to simulate and study the mixing process and the thermo-hydraulic consequences of a main steam line break in PWR model

    Directory of Open Access Journals (Sweden)

    Farkas Istvan

    2017-01-01

    Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

  3. Differential gene expression in a DNA double-strand-break repair mutant XRS-5 defective in Ku80. Analysis by cDNA microarray

    Energy Technology Data Exchange (ETDEWEB)

    Chan, John Y.H.; Chen, Lung-Kun; Chang, Jui-Feng [National Yang Ming Univ., Taipei, Taiwan (China). Inst. of Radiological Sciences] (and others)

    2001-12-01

    The ability of cells to rejoin DNA double-strand breaks (DSBs) usually correlates with their radiosensitivity. This correlation has been demonstrated in radiosensitive cells, including the Chinese hamster ovary mutant XRS-5. XRS-5 is defective in a DNA end-binding protein, Ku80, which is a component of a DNA-dependent protein kinase complex used for joining strand breaks. However, Ku80-deficient cells are known to be retarded in cell proliferation and growth as well as other yet to be identified defects. Using custom-made 600-gene cDNA microarray filters, we found differential gene expressions between the wild-type and XRS-5 cells. Defective Ku80 apparently affects the expression of several repair genes, including topoisomerase-I and -IIA, ERCC5, MLH1, and ATM. In contrast, other DNA repair-associated genes, such as GADD45A, EGR1 MDM2 and p53, were not affected. In addition, for large numbers of growth-associated genes, such as cyclins and clks, the growth factors and cytokines were also affected. Down-regulated expression was also found in several categories of seemingly unrelated genes, including apoptosis, angiogenesis, kinase and signaling, phosphatase, stress protein, proto-oncogenes and tumor suppressors, transcription and translation factors. A RT-PCR analysis confirmed that the XRS-5 cells used were defective in Ku80 expression. The diversified groups of genes being affected could mean that Ku80, a multi-functional DNA-binding protein, not only affects DNA repair, but is also involved in transcription regulation. Our data, taken together, indicate that there are specific genes being modulated in Ku80- deficient cells, and that some of the DNA repair pathways and other biological functions are apparently linked, suggesting that a defect in one gene could have global effects on many other processes. (author)

  4. SPES-99 IBLOCA analysis with the RELAP5 Mod3.2 code

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.; Bianchi, F.; Meloni, P.; Ferri, R.; Cattadori, G.

    2000-01-01

    SPES is an experimental facility which allows the assessment of thermalhydraulic codes through the simulation of a wide range of physical phenomena characterising different accident scenarios in PWRs. In 1999, in view of a possible proposal for an international cooperative programme, ENEA provided the SIET company with limited funding to restore the facility after five-year shutdown and ordered SIET a ''demonstration'' experiment where the new SPES configuration and the test conditions were selected to exploit in a valuable way the unique and outstanding characteristics of the facility. The test, defined by a working group composed of ENEA, ANPA, JRC Ispra, ANSALDO, Pisa University and SIET, consisted of a 10'' equivalent break in Cold Leg, starting from full power and full pressure conditions. This document deals with the analysis of the comparison between calculated results and experimental data. Moreover, it underlines the SPES facility capability to simulate IB-LOCAs in addition to SB-LOCAs and other kind of transients for which it was designed. (author)

  5. Key Impact Factors on Dam Break Fatalities

    Science.gov (United States)

    Huang, D.; Yu, Z.; Song, Y.; Han, D.; Li, Y.

    2016-12-01

    Dam failures can lead to catastrophes on human society. However, there is a lack of research about dam break fatalities, especially on the key factors that affect fatalities. Based on the analysis of historical dam break cases, most studies have used the regression analysis to explore the correlation between those factors and fatalities, but without implementing optimization to find the dominating factors. In order to understand and reduce the risk of fatalities, this study has proposed a new method to select the impact factors on the fatality. It employs an improved ANN (Artificial Neural Network) combined with LOOCV (Leave-one-out cross-validation) and SFS (Stepwise Forward Selection) approach to explore the nonlinear relationship between impact factors and life losses. It not only considers the factors that have been widely used in the literature but also introduces new factors closely involved with fatalities. Dam break cases occurred in China from 1954 to 2013 are summarized, within which twenty-five cases are selected with a comprehensive coverage of geographic position and temporal variation. Twelve impact factors are taken into account as the inputs, i.e., severity of dam break flood (SF), population at risk (PR), public understanding of dam break (UB), warning time (TW), evacuation condition (EC), weather condition during dam break (WB), dam break mode (MB), water storage (SW), building vulnerability (VB), dam break time (TB), average distance from the affected area to the dam (DD) and preventive measures by government (PG).From those, three key factors of SF, MB and TB are chosen. The proposed method is able to extract the key factors, and the derived fatality model performs well in various types of dam break conditions.

  6. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    International Nuclear Information System (INIS)

    Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

    2016-01-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  7. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    Energy Technology Data Exchange (ETDEWEB)

    Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-12-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  8. A Monte Carlo model of DNA double-strand break clustering and rejoining kinetics for the analysis of pulsed-field gel electrophoresis data.

    Science.gov (United States)

    Pinto, M; Prise, K M; Michael, B D

    2004-10-01

    In studies of radiation-induced DNA fragmentation and repair, analytical models may provide rapid and easy-to-use methods to test simple hypotheses regarding the breakage and rejoining mechanisms involved. The random breakage model, according to which lesions are distributed uniformly and independently of each other along the DNA, has been the model most used to describe spatial distribution of radiation-induced DNA damage. Recently several mechanistic approaches have been proposed that model clustered damage to DNA. In general, such approaches focus on the study of initial radiation-induced DNA damage and repair, without considering the effects of additional (unwanted and unavoidable) fragmentation that may take place during the experimental procedures. While most approaches, including measurement of total DNA mass below a specified value, allow for the occurrence of background experimental damage by means of simple subtractive procedures, a more detailed analysis of DNA fragmentation necessitates a more accurate treatment. We have developed a new, relatively simple model of DNA breakage and the resulting rejoining kinetics of broken fragments. Initial radiation-induced DNA damage is simulated using a clustered breakage approach, with three free parameters: the number of independently located clusters, each containing several DNA double-strand breaks (DSBs), the average number of DSBs within a cluster (multiplicity of the cluster), and the maximum allowed radius within which DSBs belonging to the same cluster are distributed. Random breakage is simulated as a special case of the DSB clustering procedure. When the model is applied to the analysis of DNA fragmentation as measured with pulsed-field gel electrophoresis (PFGE), the hypothesis that DSBs in proximity rejoin at a different rate from that of sparse isolated breaks can be tested, since the kinetics of rejoining of fragments of varying size may be followed by means of computer simulations. The problem of how

  9. Strong Electroweak Symmetry Breaking

    CERN Document Server

    Grinstein, Benjamin

    2011-01-01

    Models of spontaneous breaking of electroweak symmetry by a strong interaction do not have fine tuning/hierarchy problem. They are conceptually elegant and use the only mechanism of spontaneous breaking of a gauge symmetry that is known to occur in nature. The simplest model, minimal technicolor with extended technicolor interactions, is appealing because one can calculate by scaling up from QCD. But it is ruled out on many counts: inappropriately low quark and lepton masses (or excessive FCNC), bad electroweak data fits, light scalar and vector states, etc. However, nature may not choose the minimal model and then we are stuck: except possibly through lattice simulations, we are unable to compute and test the models. In the LHC era it therefore makes sense to abandon specific models (of strong EW breaking) and concentrate on generic features that may indicate discovery. The Technicolor Straw Man is not a model but a parametrized search strategy inspired by a remarkable generic feature of walking technicolor,...

  10. 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.

  11. Evaluation of uncertainty in dam-break analysis resulting from dynamic representation of a reservoir; Evaluation de l'incertitude due au modele de representation du reservoir dans les analyses de rupture de barrage

    Energy Technology Data Exchange (ETDEWEB)

    Tchamen, G.W.; Gaucher, J. [Hydro-Quebec Production, Montreal, PQ (Canada). Direction Barrage et Environnement, Unite Barrages et Hydraulique

    2010-08-15

    Owners and operators of high capacity dams in Quebec have a legal obligation to conduct dam break analysis for each of their dams in order to ensure public safety. This paper described traditional hydraulic methodologies and models used to perform dam break analyses. In particular, it examined the influence of the reservoir drawdown submodel on the numerical results of a dam break analysis. Numerical techniques from the field of fluid mechanics and aerodynamics have provided the basis for developing effective hydrodynamic codes that reduce the level of uncertainties associated with dam-break analysis. A static representation that considers the storage curve was compared with a dynamic representation based on Saint-Venant equations and the real bathymetry of the reservoir. The comparison was based on breach of reservoir, maximum water level, flooded area, and wave arrival time in the valley downstream. The study showed that the greatest difference in attained water level was in the vicinity of the dam, and the difference decreased as the distance from the reservoir increased. The analysis showed that the static representation overestimated the maximum depth and inundated area by as much as 20 percent. This overestimation can be reduced by 30 to 40 percent by using dynamic representation. A dynamic model based on a synthetic trapezoidal reconstruction of the storage curve was used, given the lack of bathymetric data for the reservoir. It was concluded that this model can significantly reduce the uncertainty associated with the static model. 7 refs., 9 tabs., 7 figs.

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

    International Nuclear Information System (INIS)

    Faidy, C.; Gilles, P.

    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

  13. 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.

  14. Assessment of Environmental Qualification Practices and Condition Monitoring Techniques for Low-Voltage Electric Cables: LOCA Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Lofaro, R. [Brookhaven National Lab. (BNL), Upton, NY (United States); Grove, E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Villaran, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Soo, P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hsu, F. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2001-02-01

    This report documents the results of a research program addressing issues related to the qualification process for low-voltage instrumentation and control (I&C) electric cables used in commercial nuclear power plants. Three commonly used types of I&C cable were tested: Cross-Linked Polyethylene (XLPE) insulation with a Neoprene® jacket, Ethylene Propylene Rubber (EPR) insulation with an unbonded Hypalon® jacket, and EPR with a bonded Hypalon® jacket. Each cable type received accelerated aging to simulate 20, 40, and 60 years of qualified life. In addition, naturally aged cables of the same types were obtained from decommissioned nuclear power plants and tested. The cables were subjected to simulated loss-of-coolant-accident (LOCA) conditions, which included the sequential application of LOCA radiation followed by exposure to steam at high temperature and pressure, as well as to chemical spray. Periodic condition monitoring (CM) was performed using nine different techniques to obtain data on the condition of the cable, as well as to evaluate the effectiveness of those CM techniques for in situ monitoring of cables. Volume 1 of this report presents the results of the LOCA tests, and Volume 2 discusses the results of the condition monitoring tests.

  15. RELAP/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)

    Peeler, G.B.; McDonald, T.A.; Kennedy, M.F.

    1984-01-01

    RELAP/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

  16. Drought-breaking love: An analysis of the moral values implied in ‘Drought’ by Jan Rabie

    Directory of Open Access Journals (Sweden)

    C. N. van der Merwe

    1996-05-01

    Full Text Available In this article the tension in 20th century literary theory between absolutism and relativism is discussed. It is argued that, in spite of a movement from absolutism towards relativism, the age-old “absolute” values of truth, beauty and goodness have never been totally forsaken in the creation and the contemplation of literature. In an analysis of “Drought” by Jan Rabie, it is indicated how these values are implied and invoked in Rabie's short story. In conclusion, the fundamental value of love or charity is discussed, a value which contains and supersedes the values of truth, beauty and goodness, and reconciles the tension between absolutism and relativism.

  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...... pedagogical elements on which the education in entrepreneurship rests....

  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. An analysis of reactor pit pressurization and forces applied on reactor vessel

    International Nuclear Information System (INIS)

    Wang Rongzhong; Li Feng

    1997-12-01

    The pressure and temperature transients with the time of the reactor pit during LOCA have been analyzed by using Catem computer code for Qinshan-2 nuclear power plant. The force and bending moment on the inlet and outlet nozzles of the reactor vessel also have been calculated by using Wformom code. Qinshan-2 NPP is a two-loop nuclear power plant. The cold water of the accumulators are directly injected into the downcomer of reactor vessel. Injection line of accumulators is located at the same level with the inlet and outlet nozzles. These geometry characteristics have been taken into account in the circumferential vessel pit nodding using five volumes around the vessel. The assumptions used in the analysis and calculation results have been presented. Many sensitive calculations have been performed for different break size and circumferential nodding

  20. Physiological characterization of leaf and internode after bud break in Japanese indigenous Koshu grape by comparative RNA sequencing analysis

    Science.gov (United States)

    Enoki, Shinichi; Hamaguchi, Yu; Suzuki, Shunji; Fujisawa, Hiroyuki; Hattori, Tomoki; Arita, Kayo; Yamaguchi, Chiho; Mikami, Masachika; Nagasaka, Shu; Tanaka, Keisuke

    2018-01-01

    Koshu is indigenous to Japan and considered the most important wine grape in Japan. Koshu grape berry possesses characteristics that make it unique from European V. vinifera as wine grape. However, the physiological characteristics of Koshu leaf and internode remain unknown. An understanding of those characteristics would contribute to improvements in Koshu cultivation, thereby enhancing grape berry and wine quality. To identify the genes responsible for the physiological characteristics of Koshu, we comprehensively analyzed leaf and internode differences at the transcriptome level between Koshu and Pinot Noir by RNA sequencing. A total of 248 and 131 differentially expressed genes (DEGs) were detected in leaves and internodes, respectively. Gene Ontology (GO) and Kyoto Encyclopedia of Genes and Genomes (KEGG) pathway enrichment analyses of these DEGs revealed that “flavonoid biosynthesis” and “glutathione metabolism” pathways were significantly enriched in Koshu leaves. On the other hand, when internodes were compared, “flavonoid”-related GO terms were specifically detected in Koshu. KEGG pathway enrichment analysis suggested that the expression of such genes as leucoanthocyanidin reductase and flavonol synthase in the flavonoid biosynthesis pathway was higher in Koshu than Pinot Noir. Measurement of the relative expression levels of these genes by RT-qPCR validated the results obtained by RNA sequencing. The characteristics of Koshu leaf and internode, which are expected to produce flavonoids with antibacterial activity and UV protection function, would suit Japanese climate as a survival strategy. PMID:29566077

  1. Quantitative analysis of the thermal requirements for stepwise physical dormancy-break in seeds of the winter annual Geranium carolinianum (Geraniaceae).

    Science.gov (United States)

    Gama-Arachchige, N S; Baskin, J M; Geneve, R L; Baskin, C C

    2013-05-01

    Physical dormancy (PY)-break in some annual plant species is a two-step process controlled by two different temperature and/or moisture regimes. The thermal time model has been used to quantify PY-break in several species of Fabaceae, but not to describe stepwise PY-break. The primary aims of this study were to quantify the thermal requirement for sensitivity induction by developing a thermal time model and to propose a mechanism for stepwise PY-breaking in the winter annual Geranium carolinianum. Seeds of G. carolinianum were stored under dry conditions at different constant and alternating temperatures to induce sensitivity (step I). Sensitivity induction was analysed based on the thermal time approach using the Gompertz function. The effect of temperature on step II was studied by incubating sensitive seeds at low temperatures. Scanning electron microscopy, penetrometer techniques, and different humidity levels and temperatures were used to explain the mechanism of stepwise PY-break. The base temperature (Tb) for sensitivity induction was 17·2 °C and constant for all seed fractions of the population. Thermal time for sensitivity induction during step I in the PY-breaking process agreed with the three-parameter Gompertz model. Step II (PY-break) did not agree with the thermal time concept. Q10 values for the rate of sensitivity induction and PY-break were between 2·0 and 3·5 and between 0·02 and 0·1, respectively. The force required to separate the water gap palisade layer from the sub-palisade layer was significantly reduced after sensitivity induction. Step I and step II in PY-breaking of G. carolinianum are controlled by chemical and physical processes, respectively. This study indicates the feasibility of applying the developed thermal time model to predict or manipulate sensitivity induction in seeds with two-step PY-breaking processes. The model is the first and most detailed one yet developed for sensitivity induction in PY-break.

  2. 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.

  3. A validation of ATR LOCA thermal-hydraulic code with a statistical approach

    International Nuclear Information System (INIS)

    Mochizuki, Hiroyasu

    2000-01-01

    When cladding temperatures are measured for a blowdown experiment, cladding temperatures at the same elevation in the fuel bundle have usually some differences due to eccentricity of the fuel bundle and other reasons such as biased two-phase flow. In the present paper, manufacturing tolerances and uncertainties of thermal-hydraulics are incorporated into a LOCA code that is applied with the statistical method. The present method was validated with the results of different blowdown experiments conducted using the 6 MW blowdown facility simulating the Advanced Thermal Reactor (ATR). In the present statistical method, the code was modified to run fast in order to calculate the blowdown thermal-hydraulics a lot of times with the code using different sets of input data. These input data for sizes and empirical correlations are prepared by the effective Monte-Carlo method based on the distribution functions deduced by the measured manufacturing errors and the uncertainties of thermal hydraulics. The calculated curves express uncertainties due to the different input deck. The uncertainty band and tendency of the cladding temperature were dependent on the beak sizes in the experiment. The measured results were traced by the present method. (author)

  4. 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.

  5. Evaluation of the pressure difference across the core during PWR-LOCA reflood phase

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Murao, Yoshio

    1979-03-01

    The flooding rate of the core influences largely cooling of the core during the reflood phase of a PWR-LOCA. Since the void fraction of two-phase flow in the core is important determining the flooding rate, it is essential to examine this void fraction. The void fraction in the core during the reflood phase obtained by experiment was compared with those predicted by the correlations respectively of Akagawa, Nicklin, Zuber, Yeh, Griffice, Behringer and Jhonson. Only Yeh's correlation was found to be usable for the purpose. The pressure difference of the core during the reflood phase was calculated by reflood analyzing code REFLA-1D using Yeh's correlation. Following are the results: (1) During the steady-state period after quenching of the heaters, the prediction agrees within +-15% with the experiment. (2) During the transient period when the quench front is advancing, the prediction is not in agreement with the experiment, the difference being about +-40%. Influence of the advancing quench front upon the void fraction in the core must further be studied. (author)

  6. New roles for astrocytes: the nightlife of an 'astrocyte'. La vida loca!

    Science.gov (United States)

    Horner, Philip J; Palmer, Theo D

    2003-11-01

    Like a newly popular nightspot, the biology of adult stem cells has emerged from obscurity to become one of the most lively new disciplines of the decade. The neurosciences have not escaped this trendy pastime and, from amid the noise and excitement, the astrocyte emerges as a beguiling companion to the adult neural stem cell. A once receding partner to neurons and oligodendrocytes, the astrocyte even takes on an alter ego of the stem cell itself (S. Goldman, this issue of TINS). Putting ego aside, the 'astrocyte' is also (and perhaps more importantly) an integral component of neural progenitor hotspots, where the craziness or 'la vida loca' of the nightlife might not be so wild when compared with our traditional understanding of the astrocyte. Here, astrocytes contribute to the instructive confluence of location, atmosphere and cellular neighbors that define the daily 'vida local' or everyday local life of an adult stem cell. This review discusses astrocytes as influential components in the local stem cell niche.

  7. Analytical methods of leakage rate estimation from a containment under a LOCA

    International Nuclear Information System (INIS)

    Chun, M.H.

    1981-01-01

    Three most outstanding maximum flow rate formulas are identified from many existing models. Outlines of the three limiting mass flow rate models are given along with computational procedures to estimate approximate amount of fission products released from a containment to environment for a given characteristic hole size for containment-isolation failure and containment pressure and temperature under a loss of coolant accident. Sample calculations are performed using the critical ideal gas flow rate model and the Moody's graphs for the maximum two-phase flow rates, and the results are compared with the values obtained from then mass leakage rate formula of CONTEMPT-LT code for converging nozzle and sonic flow. It is shown that the critical ideal gas flow rate formula gives almost comparable results as one can obtain from the Moody's model. It is also found that a more conservative approach to estimate leakage rate from a containment under a LOCA is to use the maximum ideal gas flow rate equation rather than the mass leakage rate formula of CONTEMPT-LT. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. 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.

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

    International Nuclear Information System (INIS)

    Luo Bangqi

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  13. Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

    2003-01-01

    A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

  14. Breaking News as Radicalisation

    DEFF Research Database (Denmark)

    Hartley, Jannie Møller

    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?......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...

  15. 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.”...

  16. Corruption analysis and life span forecast research on sylphon bellows of nuclear-powered steam system of ship

    International Nuclear Information System (INIS)

    Song Chao; Chen Lisheng; Song Meicun; Wang Wei

    2012-01-01

    The fracture of the corrugated pipe has a dad effect to the operation of reactor which can cause the small-break LOCA. The corrosion is the key reason of the fracture. On the base of the analysis on corruption reason of the sylphon bellows and combine the characteristic of the limited sample point of it, the Grey theory was used in the assessment of corrosion life span of corrugated pipe in nuclear steam system. Through applying the GM (1, 1) model in inferring the discipline of corrosion quantity and combining traditional statistical method, the corrosion life of steam pipe was evaluated. It indicates that the model is precise, simple and the result is reliable. (authors)

  17. Global vibrations in the wetwell condensation process caused by LOCA in BWR plants

    International Nuclear Information System (INIS)

    Bjoerndahl, O.; Andersson, Magnus

    1998-12-01

    During the last years a substantial part of third part review work related to dynamical loadings has been review of loading specifications dealing with vibrations in containment building related to so called LOCA-events in Swedish BWR plants. Compared to other loading categories characterised as global vibrations these secondary effects of LOCA-events are complex to analyse. One experience from the review work at SAQ up to now is that it is not fully clear what prediction methods and what model idealisations are the most adequate for structural integrity verification on mechanical systems as pressure vessels and piping under such loading conditions. At SAQ Teknik a project work has been carried out to investigate the general status of the methodology used today in Sweden and a work to in the long term develop simplified prediction models and methods for the loading categories condensation oscillations (CO) and chugging (CH). The work was initially concentrated on a study of the background of the methodology which was developed for these type of loading in American BWR-containments of the Mark-II design. The methodology was developed by General Electric, GE, in cooperation with the Mark-II plant owners. The methodology used in Sweden to predict vibrations in BWR containments of this design is with some minor modifications very close to technique developed by GE. The methodology developed by GE is the only accepted by USNRC for the Mark-II design and could be found as reference in Standard Review Plan 6.2.1.1.C, Rev 6 - August 1984. Based on identical physical assumptions about the dynamic behaviour of the building structure and the water in the suppression pool, mathematical models are derived in this report for predictions of secondary structure response spectra for loading conditions as global vibrations during CO and CH. Based on parameters identified by so called one pipe experiments responses my be predicted. By use of these derived mathematical models as a

  18. 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

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

    International Nuclear Information System (INIS)

    Stephenson, W.; Dutton, L.M.C.; Handy, B.J.; Smedley, C.

    1992-01-01

    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

  20. PELE-IC, 2-D Eulerian Incompressible Hydrodynamic and Bubble Dynamic after LWR LOCA

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.

    1981-01-01

    1 - Description of problem or function: PELE-IC is a two-dimensional semi-implicit Eulerian hydrodynamics program for the solution of incompressible flow coupled to flexible structures. The code was developed to calculate fluid-structure interactions and bubble dynamics of a pressure-suppression system following a loss-of- coolant accident (LOCA). The fluid, structure, and coupling algorithms have been verified by calculation of benchmark problems and air and steam blowdown experiments. The code is written for both plane and cylindrical coordinates. The coupling algorithm is general enough to handle a wide variety of structural shapes. The concepts of void fractions and interface orientation are used to track the movement of free surfaces, allowing great versatility in following fluid-gas interfaces both for bubble definition and water surface motion without the use of marker particles. 2 - Method of solution: The solution strategy is to first solve the Navier-Stokes equations explicitly using values from the previous time-step. Since these values do not necessarily satisfy the continuity equation, the pressure field is iterated upon until the incompressibility condition for each computational cell is satisfied within prescribed limits. The structural motion is computed by a finite element code from the applied pressure at the fluid-structure interface. The shell structure algorithm uses conventional thin-shell theory with transverse shear. The finite-element spatial discretization employs piecewise-linear interpolation functions and one-point quadrature applied to conical frustra. The Newmark implicit time integration method is used as a one-step module. The fluid code then uses the structure's position and velocity as boundary conditions. The fluid pressure field and the structure's response are corrected iteratively until the normal velocities of fluid and structure are equal. The effects of steam condensation and oscillatory chugging on structures are

  1. Mixing of radiolytic hydrogen generated within a containment compartment following a LOCA

    International Nuclear Information System (INIS)

    Willcutt, G.J.E. Jr.; Gido, R.G.

    1978-07-01

    The objective of this work was to determine hydrogen concentration variations with position and time in a closed containment compartment with radiolytic hydrogen generation in the water on the compartment floor following a Loss-of-Coolant-Accident (LOCA). One application is to determine the potential difference between the compartment maximum hydrogen concentration and a hydrogen detector reading, due to the detector location. Three possible mechanisms for hydrogen transport in the compartment were investigated: (1) molecular diffusion, (2) possible bubble formation and motion, and (3) natural convection flows. A base case cubic compartment with 6.55-m (21.5-ft) height was analyzed. Parameter studies were used to determine the sensitivity of results to compartment size, hydrogen generation rates, diffusion coefficients, and the temperature difference between the floor and the ceiling and walls of the compartment. Diffusion modeling indicates that if no other mixing mechanism is present for the base case, the maximum hydrogen volume percent (vol percent) concentration difference between the compartment floor and ceiling will be 4.8 percent. It will be 24.5 days before the maximum concentration difference is less than 0.5 percent. Bubbles do not appear to be a potential source of hydrogen pocketing in a containment compartment. Compartment natural convection circulation rates for a 2.8 K (5 0 F) temperature difference between the floor and the ceiling and walls are estimated to be at least the equivalent of 1 compartment volume per hour and probably in the range of 4 to 9 compartment volumes per hour. Related natural convection studies indicate there will be turbulent mixing in the compartment for a 2.8 K (5 0 F) temperature difference between the floor and the ceiling and walls

  2. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    International Nuclear Information System (INIS)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong

    2006-01-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  3. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    Science.gov (United States)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  4. 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)

    Lelong, Franck

    2010-01-01

    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)

  5. Wave-breaking and generic singularities of nonlinear hyperbolic equations

    International Nuclear Information System (INIS)

    Pomeau, Yves; Le Berre, Martine; Guyenne, Philippe; Grilli, Stephan

    2008-01-01

    Wave-breaking is studied analytically first and the results are compared with accurate numerical simulations of 3D wave-breaking. We focus on the time dependence of various quantities becoming singular at the onset of breaking. The power laws derived from general arguments and the singular behaviour of solutions of nonlinear hyperbolic differential equations are in excellent agreement with the numerical results. This shows the power of the analysis by methods using generic concepts of nonlinear science. (open problem)

  6. Testing to evaluate synergistic effects from LOCA environments. Test IX. Simultaneous mode; cables, splice assemblies, and electrical insulation samples

    Energy Technology Data Exchange (ETDEWEB)

    Thome, F.V.

    1978-04-01

    This test was conducted to complement Test VIII which was a sequential test of cables, cable splices, and insulation samples. In this test, the generic LOCA environments (radiation, temperature, pressure, chemical spray) were simulated and simultaneously applied to the test items. There were no failures of any assemblies and all were able to function at rated current and voltage throughout the entire test. An additional parameter, dissipation factor, was monitored in this test and when used in conjunction with capacitance, provided a better indication of insulation degradation.

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

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

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

  8. A nuclear power unit with a Babcock type steam generating system-analysis of the break-down in the Three Mile Island power plant

    International Nuclear Information System (INIS)

    Werner, A.

    1980-01-01

    Installations of the primary and the secondary circuits and basic automatic control and protection systems for a nuclear power unit with Babcock type vertical, once-through steam generator are described. On this background the course of the break-down in the Three Mile Island power plant at Harrisburg is presented and analysed. (author)

  9. Spontaneous breaking of supersymmetry

    Energy Technology Data Exchange (ETDEWEB)

    Zumino, B.

    1981-12-01

    There has been recently a revival of interest in supersymmetric gauge theories, stimulated by the hope that supersymmetry might help in clarifying some of the questions which remain unanswered in the so called Grand Unified Theories and in particular the gauge hierarchy problem. In a Grand Unified Theory one has two widely different mass scales: the unification mass M approx. = 10/sup 15/GeV at which the unification group (e.g. SU(5)) breaks down to SU(3) x SU(2) x U(1) and the mass ..mu.. approx. = 100 GeV at which SU(2) x U(1) is broken down to the U(1) of electromagnetism. There is at present no theoretical understanding of the extreme smallness of the ratio ..mu../M of these two numbers. This is the gauge hierarchy problem. This lecture attempts to review the various mechanisms for spontaneous supersymmetry breaking in gauge theories. Most of the discussions are concerned with the tree approximation, but what is presently known about radiative correction is also reviewed.

  10. 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.

  11. Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

    International Nuclear Information System (INIS)

    Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

    1981-07-01

    The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times

  12. 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 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...

  13. 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)

  14. Spatially Resolved Analysis of the Interstellar Medium in the Cosmic Eye, a Lensed Lyman Break Galaxy at z=3.074

    Science.gov (United States)

    Ball, Catherine; Riechers, Dominik A.; Pavesi, Riccardo

    2018-01-01

    The [CII]/[NII] ratio combines the [CII] line, a tracer of photodissociation and HII regions emerging from the neutral and ionized phases of the interstellar medium (ISM), with [NII] emission, which only originates from the ionized ISM. In this, the [CII]/[NII] ratio can be used to separate the fractions of [CII] emission emerging from the different phases of the ISM. We present Atacama Large sub-Millimeter Array (ALMA) observations of the Cosmic Eye, a gravitationally lensed Lyman Break Galaxy (LBG). As an LBG, the Cosmic Eye represents a "normal" star forming galaxy in the z>2 universe. LBGs were host to the bulk of star formation during the peak epoch of star formation. Diagnosing star formation in these galaxies provides insight into the evolution of “normal” galaxies in a cosmic sense. The high magnification (30x) allows us to resolve the [CII] 158μm and the [NII] 205μm lines in detail, allowing for a position-resolved analysis of their ratio. We find variations of the line ratio across the galaxy, suggesting the galaxy’s internal structure affects this ratio. We consider the Cosmic Eye in the context of both higher redshift LBGs and local luminous and ultraluminous infrared galaxies, finding that the Cosmic Eye’s line ratio is similar to those of both higher- and lower- redshift galaxies. The Cosmic Eye’s global [CII]/[NII] ratio sits between two previous measurements of z>5 LBGs at low resolution, suggesting that the ratio may correlate more significantly with LFIR than with redshift in this epoch. Furthermore, the Cosmic Eye’s [CII]/[NII] ratio is similar to those of the nearby LIRG/ULIRGs, though we expect local [CII]/[NII] values to be lower due to their different metallicities and dust content. High-resolution studies like this one probe the evolution of [CII]/[NII] over cosmic time by examining the evolution of the ISM’s structure. With a better understanding of the [CII]/[NII] line ratio, we can more effectively use it as a probe of the

  15. Performance evaluation of a new signal processing system design to improve CANDU SDS1 trip response during large break LOCA events

    International Nuclear Information System (INIS)

    Xia, Lingzhi; Gabbar, Hossam A.; Isham, Manir U.; Ponomarev, Vladimir

    2016-01-01

    Performance of a recently developed signal processing system for CANDU (Canada Deuterium Uraniu) reactor shutdown system 1 (SDS1) is evaluated in this paper. The evaluation is carried out in MATLAB/Simulink software environment as well as with an existing power measurement and signal processing system. The new signal processing algorithm is obtained based on the synthesis of several first order low pass filters with different delayed time constants. Throughout this paper, a special attention has been paid to compare the new signal processing system with the existing one. The dynamic behavior of the new signal processing system in the practical large loss of coolant accidents (LLOCA) events has also been examined. Simulation results show that during the LLOCA event, the reactor trip time, as well as the peak power, is decreased remarkably. Through the simulation studies, it has convincingly demonstrated that the new signal processing system has significant advantages over the existing system in terms of the improved trip response and accommodation of the spurious trip immunity. This advantage will significantly enhance the safety margin, or will bring economical benefits to nuclear power plants. (author)

  16. THE PREDICTION OF pH BY GIBBS FREE ENERGY MINIMIZATION IN THE SUMP SOLUTION UNDER LOCA CONDITION OF PWR

    Directory of Open Access Journals (Sweden)

    HYOUNGJU YOON

    2013-02-01

    Full Text Available It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3&4 and UCN 1&2. As results, pH of the sump solution for the SKN 3&4 was between 7.02 and 7.45, and for the UCN 1&2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl, HNO3, and Cs are very low.

  17. 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)

    Chadwick, Chris; Jahouel, Xavier; Swain, Adam

    2014-01-01

    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

  18. 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....

  19. Should I Give the Exam before or after the Break?

    Science.gov (United States)

    O'Connor, Kevin J.

    2014-01-01

    This study was designed to help faculty make decisions about when to administer an exam in relation to an in-semester break. Students in multiple sections of an undergraduate educational psychology class were assigned to take an exam either before or after a scheduled 5-day break (Thursday-Monday). A multiple regression analysis revealed the break…

  20. Global analysis of double-strand break processing reveals in vivo properties of the helicase-nuclease complex AddAB.

    Science.gov (United States)

    Badrinarayanan, Anjana; Le, Tung B K; Spille, Jan-Hendrik; Cisse, Ibrahim I; Laub, Michael T

    2017-05-01

    In bacteria, double-strand break (DSB) repair via homologous recombination is thought to be initiated through the bi-directional degradation and resection of DNA ends by a helicase-nuclease complex such as AddAB. The activity of AddAB has been well-studied in vitro, with translocation speeds between 400-2000 bp/s on linear DNA suggesting that a large section of DNA around a break site is processed for repair. However, the translocation rate and activity of AddAB in vivo is not known, and how AddAB is regulated to prevent excessive DNA degradation around a break site is unclear. To examine the functions and mechanistic regulation of AddAB inside bacterial cells, we developed a next-generation sequencing-based approach to assay DNA processing after a site-specific DSB was introduced on the chromosome of Caulobacter crescentus. Using this assay we determined the in vivo rates of DSB processing by AddAB and found that putative chi sites attenuate processing in a RecA-dependent manner. This RecA-mediated regulation of AddAB prevents the excessive loss of DNA around a break site, limiting the effects of DSB processing on transcription. In sum, our results, taken together with prior studies, support a mechanism for regulating AddAB that couples two key events of DSB repair-the attenuation of DNA-end processing and the initiation of homology search by RecA-thereby helping to ensure that genomic integrity is maintained during DSB repair.

  1. Analysis on Containment Response Following a LBLOCA of APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [KINS, Daejeon (Korea, Republic of)

    2016-05-15

    The predictions are in good agreements with the final safety analysis report, which implies the containment integrity is maintained during or after an accident like loss of coolant accident. In this study, the CONTEMPT-LT/028 was used to calculate the pressure and temperature, and in the follow-up study, CONTAIN 2.0 will be used for the pressure and temperature predictions in APR1400 reactors. Shin-Hanul Units 1 and 2 may possess different characteristics of peak pressure and temperature in containment following a large break loss-of-coolant-accident. To assess the important performance independently and to compare with prediction results presented in the final safety analysis report (FSAR) of Shin-Hanul Units 1 and 2 might be helpful to regulatory review for identifying validity of the FSAR. The end of blowdown (EOB) time during a LOCA could largely affect the peak pressure and temperature in the containment. This paper provides CONTEMPT-LT/028 prediction of the peak pressure and temperature of Shin-Hanul Units 1 and 2 following a large break loss-of-coolant-accident and compares with licensee's prediction results.

  2. Breaking soliton equations and negative-order breaking soliton ...

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics; Volume 87; Issue 5. Breaking soliton ... We use the simplified Hirota's method to obtain multiple soliton solutions for each developed breaking soliton equation. We also develop ... WAZWAZ1. Department of Mathematics, Saint Xavier University, Chicago, IL 60655, USA ...

  3. Breaking soliton equations and negative-order breaking soliton ...

    Indian Academy of Sciences (India)

    We develop breaking soliton equations and negative-order breaking soliton equations of typical and higher orders. The recursion operator of the KdV equation is used to derive these models.We establish the distinctdispersion relation for each equation. We use the simplified Hirota's method to obtain multiple soliton ...

  4. Breaking soliton equations and negative-order breaking soliton ...

    Indian Academy of Sciences (India)

    2016-10-06

    Oct 6, 2016 ... Abstract. We develop breaking soliton equations and negative-order breaking soliton equations of typical and higher orders. The recursion operator of the KdV equation is used to derive these models. We establish the distinct dispersion relation for each equation. We use the simplified Hirota's method to ...

  5. Simulation of a large-break loss of coolant accident in the high performance light water reactor

    International Nuclear Information System (INIS)

    Kurki, Joona; Haenninen, Markku

    2010-01-01

    valves, all the water flowing out of the pressure vessel runs through the break orifice into the containment. Low head safety injection, an active safety component, starts to inject cold water at the reactor inlet as the pressure has decreased sufficiently. The presented analysis suggests that the reactor core of the HPLWR can be kept sufficiently cooled-down in the case of a large break LOCA in the main steam line using the specified safety systems. Some uncertainty to the simulation results is caused by the constitutive equations used at supercritical pressures, but they have very limited effect on the overall results in a simulation case, where the pressure drops to subcritical conditions at a very early stage. Also the current model with only three thermal hydraulic core channels is sufficient for this kind of simulation, where the reactor is brought to decay heat very early on to the simulation. Proper modeling of flows in each fuel cluster would be needed for more elaborate analyses, and for example analyses of reactivity initiated accidents. (author)

  6. Development of a system code for transient analysis in a HTGR

    International Nuclear Information System (INIS)

    Lee, Tae Beom

    2004-02-01

    A GAMMA (GAs Multi-component Multi-dimensional Analysis) code is developed for transient analysis and air ingress analysis in High Temperature Gas-cooled Reactors (HTGR). The PBMR of ESKOM is selected as a reference plant for the High Temperature Gas-cooled Reactor here, which uses a direct helium cycle and pebble fuel. Physical models included in GAMMA are the pebble conduction model, radiation heat transfer model, point kinetics model, decay heat model, and component models for break flow, valve, pump, cooler, power conversion unit model. The temperature distribution and the flow distribution of the PBMR are calculated for initial and accident core in the present study. In the accident analysis, typical design basis accident (DBA), including the load transient accident and depressurization accident into the system are selected and analyzed in detail. The predictions by GAMMA for PBMR at 100% power are compared with those by VSOP and PBR S IM. It turns out that the temperature in the upper region in the third channel predicted by GAMMA is about 62 .deg. C at maximum higher than that by VSOP, but is pretty close to that by PBR S IM. The center temperature of the fuel shows that that predicted by considering swelling effect is higher than that without swelling effect by about 10 .deg. C. The net efficiency of direct system is higher than that of indirect system due to an effect of the circulator power. The transient capability of GAMMA is validated through analytical solution and PBR S IM analyzing the depressurization (Loss Of Coolant Accident, LOCA) and load transient accident. After the LOCA the system pressure decreases dramatically from 8MPa to 0.4MPa within 2 sec. After the PI (Proportional-plus-Integral) controller senses that the power shaft is over the set-point of 3,600 rpm, the bypass valve makes shaft speed back to the set-point

  7. 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

    Energy Technology Data Exchange (ETDEWEB)

    Nocentini, S. [Institut Curie de Biologie, Paris (France)

    1995-11-01

    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 {gamma} 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 {gamma} 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.

  8. Core break-off mechanism

    Science.gov (United States)

    Myrick, Thomas M. (Inventor)

    2003-01-01

    A mechanism for breaking off and retaining a core sample of a drill drilled into a ground substrate has an outer drill tube and an inner core break-off tube sleeved inside the drill tube. The break-off tube breaks off and retains the core sample by a varying geometric relationship of inner and outer diameters with the drill tube. The inside diameter (ID) of the drill tube is offset by a given amount with respect to its outer diameter (OD). Similarly, the outside diameter (OD) of the break-off tube is offset by the same amount with respect to its inner diameter (ID). When the break-off tube and drill tube are in one rotational alignment, the two offsets cancel each other such that the drill can operate the two tubes together in alignment with the drill axis. When the tubes are rotated 180 degrees to another positional alignment, the two offsets add together causing the core sample in the break-off tube to be displaced from the drill axis and applying shear forces to break off the core sample.

  9. supersymmetry breaking with extra dimensions

    Indian Academy of Sciences (India)

    This talk reviews some aspects of supersymmetry breaking in the presence of extra dimensions. The first part is a general introduction, recalling the motivations for supersymmetry and extra dimensions, as well as some unsolved problems of four-dimensional models of supersymmetry breaking. The central part is a more ...

  10. Inflation from supersymmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Antoniadis, I. [UMR CNRS 7589 Sorbonne Universites, UPMC Paris 6, LPTHE, Paris (France); University of Bern, Albert Einstein Center, Institute for Theoretical Physics, Bern (Switzerland); Chatrabhuti, A.; Isono, H.; Knoops, R. [Chulalongkorn University, Department of Physics, Faculty of Science, Pathumwan, Bangkok (Thailand)

    2017-11-15

    We explore the possibility that inflation is driven by supersymmetry breaking with the superpartner of the goldstino (sgoldstino) playing the role of the inflaton. Moreover, we impose an R-symmetry that allows one to satisfy easily the slow-roll conditions, avoiding the so-called η-problem, and leads to two different classes of small-field inflation models; they are characterised by an inflationary plateau around the maximum of the scalar potential, where R-symmetry is either restored or spontaneously broken, with the inflaton rolling down to a minimum describing the present phase of our Universe. To avoid the Goldstone boson and be left with a single (real) scalar field (the inflaton), R-symmetry is gauged with the corresponding gauge boson becoming massive. This framework generalises a model studied recently by the present authors, with the inflaton identified by the string dilaton and R-symmetry together with supersymmetry restored at weak coupling, at infinity of the dilaton potential. The presence of the D-term allows a tuning of the vacuum energy at the minimum. The proposed models agree with cosmological observations and predict a tensor-to-scalar ratio of primordial perturbations 10{sup -9}

  11. Dynamic breaking of a single gold bond

    DEFF Research Database (Denmark)

    Pobelov, Ilya V.; Lauritzen, Kasper Primdal; Yoshida, Koji

    2017-01-01

    of a single Au-Au bond and show that the breaking force is dependent on the loading rate. We probe the temperature and structural dependencies of breaking and suggest that the paradox can be explained by fast breaking of atomic wires and slow breaking of point contacts giving very similar breaking forces....

  12. Fragmentation in DNA double-strand breaks

    International Nuclear Information System (INIS)

    Wei Zhiyong; Suzhou Univ., Suzhou; Zhang Lihui; Li Ming; Fan Wo; Xu Yujie

    2005-01-01

    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≅E 0 + E 1 l + E 2 l 2 , the distribution functions are obtained as exp(αl + βl 2 ). There are two components, the one proportional to exp(βl 2 ), 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)

  13. 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)

    Cervi, Ricardo Ghantous; Pinotti, Elvio Brasil; Esperancini, Maura Seiko Tsutsui; Bueno, Osmar de Carvalho

    2010-01-01

    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)

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

    International Nuclear Information System (INIS)

    Nagler, A.; Gilat, J.; Hirshfeld, H.

    1991-01-01

    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 power 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.)

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

    International Nuclear Information System (INIS)

    Nagler, A.; Gilat, J.; Hirshfeld, H.

    1991-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1976-09-01

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

  17. Remote Sensing Characteristics of Wave Breaking Rollers

    Science.gov (United States)

    Haller, M. C.; Catalan, P.

    2006-12-01

    The wave roller has a primary influence on the balances of mass and momentum in the surf zone (e.g. Svendsen, 1984; Dally and Brown, 1995; Ruessink et al., 2001). In addition, the roller area and its angle of inclination on the wave front are important quantities governing the dissipation rates in breaking waves (e.g Madsen et al., 1997). Yet, there have been very few measurements published of individual breaking wave roller geometries in shallow water. A number of investigators have focused on observations of the initial jet-like motion at the onset of breaking before the establishment of the wave roller (e.g. Basco, 1985; Jansen, 1986), while Govender et al. (2002) provide observations of wave roller vertical cross-sections and angles of inclination for a pair of laboratory wave conditions. Nonetheless, presently very little is known about the growth, evolution, and decay of this aerated region of white water as it propagates through the surf zone; mostly due to the inherent difficulties in making the relevant observations. The present work is focused on analyzing observations of the time and space scales of individual shallow water breaking wave rollers as derived from remote sensing systems. Using a high-resolution video system in a large-scale laboratory facility, we have obtained detailed measurements of the growth and evolution of the wave breaking roller. In addition, by synchronizing the remote video with in-situ wave gages, we are able to directly relate the video intensity signal to the underlying wave shape. Results indicate that the horizontal length scale of breaking wave rollers differs significantly from the previous observations of Duncan (1981), which has been a traditional basis for roller model parameterizations. The overall approach to the video analysis is new in the sense that we concentrate on individual breaking waves, as opposed to the more commonly used time-exposure technique. In addition, a new parameter of interest, denoted Imax, is

  18. NPP Krsko Containment Response Following Main Steam Line Break

    International Nuclear Information System (INIS)

    Spalj, S.; Grgic, D.; Cavlina, N.

    2000-01-01

    This paper presents the calculation of thermohydraulic environmental parameters (pressure and temperature) inside containment of Krsko NPP after postulated Main Steam Line Break (MSLB) accident. This analysis was done as a part of the ambient parameters specification in the frame of the NPP Krsko Equipment Qualification (EQ) project. The RELAP5/mod2 computer code was used for the determination of MSLB mass and energy release and computer code GOTHIC was used to calculate pressure and temperature profiles inside NPP Krsko containment. The analysis was performed for spectrum of break sizes to account for possible steam superheating during accidents with smaller break sizes. (author)

  19. ANALISA BREAK EVENT POINT (BEP TERHADAP LABA PERUSAHAAN

    Directory of Open Access Journals (Sweden)

    Muhammad Yusuf

    2015-09-01

    Full Text Available Break event point or the break-even point can be defined as a situation where the operating company does not make a profit and not a loss. The goal is to provide the knowledge to increase knowledge about the break event point (the point of principal and its relationship with the company profit and to know how the results of the. Analysis break event point is very important for the leadership of the company to determine the production rate how much the cost will be equal to the amount of sales or in other words to determine the break event point we will determine the relationship between sales, production, selling price, cost, loss or profit, making it easier for leaders to take discretion.DOI: 10.15408/ess.v4i1.1955 

  20. Smiles count but minutes matter: responses to classroom exercise breaks.

    Science.gov (United States)

    Howie, Erin K; Newman-Norlund, Roger D; Pate, Russell R

    2014-09-01

    To determine the subjective responses of teachers and students to classroom exercise breaks, and how responses varied by duration. This mixed-methods experimental study included focus groups with teachers (N = 8) and 4(th)- and 5(th)-grade students (N = 96). Students participated in 5-, 10-, and 20-minute exercise breaks and 10 minutes of sedentary activity. In an additional exploratory analysis, video-tapes of each condition were coded and compared for positive affect. Students and teachers discussed multiple benefits, but teachers discussed barriers to implementing regular breaks of 5-minutes or more. Students exhibited higher positive affect during each exercise condition. Classroom exercise breaks are an enjoyable way to increase physical activity, but additional support may be needed to encourage teachers to implement breaks of 5 minutes or longer.

  1. 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...

  2. Metastable supersymmetry breaking without scales

    Energy Technology Data Exchange (ETDEWEB)

    Bruemmer, Felix

    2010-11-15

    We construct new examples of models of metastable D=4 N=1 supersymmetry breaking in which all scales are generated dynamically. Our models rely on Seiberg duality and on the ISS mechanism of supersymmetry breaking in massive SQCD. Some of the electric quark superfields arise as composites of a strongly coupled gauge sector. This allows us to start with a simple cubic superpotential and an asymptotically free gauge group in the ultraviolet, and end up with an infrared effective theory which breaks supersymmetry dynamically in a metastable state. (orig.)

  3. 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

    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. 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. 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. 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 patient's ability to break tablets should

  4. 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)

    Hoffmann, W.; Kryk, H.

    2012-09-01

    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 NUKON R (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

  5. Receiver function and magnetotelluric analysis to understand the first stage of a continental lithospheric break-up : case of the North Tanzanian Rift

    Science.gov (United States)

    Plasman, M.; Tiberi, C.; Tarits, P.; Hautot, S.; Gautier, S.; Ebinger, C. J.; Mulibo, G. D.; Khalfan, M.

    2015-12-01

    First stage of continental break-up, though intensively studied, is yet poorly understood. This is partly because actual rifting areas are either too mature (more than 10 My) or not easily accessible (thick sediment cover or under water). The North Tanzania part of the East African Rift is the place of a lithosphere's early break-up (less than 5My) in response to a combination of regional pulling forces and mantle upwelling. Deformation there results from complex interactions between magmatic intrusions, faulting, asthenospheric dynamics and far field stresses. CoLiBrEA (ANR) and CRAFTI (NSF) are two multidisciplinary projects which collaboratively focus on this area to understand the interactions between faults and magma, the role of inherited structures and rheological heterogeneities of the lithosphere. For that purpose, we deployed 38 broadband seismic stations in the Natron and Ngorongoro areas from January 2013 to December 2014 and carried out a 120 km East-West magnetotelluric (MT) profile to image the crustal and mantle structures. The 3D resistivity model, obtained from the inversion of the MT data along the profile, shows an highly heterogeneous crust with three-dimensional structures over a more homogeneous upper mantle. The first inversion result from the receiver function (RF) by the Zhu and Kanamori's inversion method show a thick crust (~35 km) with important variations (maximum 15km) especially in the Ngorongoro area, and an average Vp/Vs ratio of 1.75. We then completed this study by combining the MT data and the RF at the 11 sites of the EW profile. For each site, we built a 1D velocity model (Vs and VpVs) obtained by combining the Sambridge forward solution with a non linear descent research algorithm and constrained by the resistivity structure. The inversion shows an heterogeneous crust obviously dominated by the Moho interface at different depths, with low velocity layers mainly corresponding to low resistivity features.

  6. 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)

  7. Breaking beer bottles with cavitation

    Science.gov (United States)

    Jung, Sunny; Fontana, Jake; Palffy-Muhoray, Peter; Shelley, Michael

    2009-03-01

    Hitting the top of a beer bottle, nearly full of water, with an open hand can cause the bottle to break, with the bottom separating from upper section. We have studied this phenomenon using a high-speed camera, and observed the formation, coalescence and collapse of bubbles. The breaking of glass is due to cavitation, typically occurring near the bottom edge. We make numerical estimates of the relevant physical parameters, and compare these with experimental observations.

  8. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  9. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH Univ. of Applied Sciences, Deggendorf (Germany)

    2014-07-01

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation programme was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment with integrated pressure suppression system. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The main target was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. (orig.)

  10. Breaking Bad News in Oncology: A Metasynthesis.

    Science.gov (United States)

    Bousquet, Guilhem; Orri, Massimiliano; Winterman, Sabine; Brugière, Charlotte; Verneuil, Laurence; Revah-Levy, Anne

    2015-08-01

    The delivery of bad news by oncologists to their patients is a key moment in the physician-patient relationship. We performed a systematic review of qualitative studies (a metasynthesis) that focused on the experiences and points of view of oncologists about breaking bad news to patients. We searched international publications to identify relevant qualitative research exploring oncologists' perspectives about this topic. Thematic analysis, which compensates for the potential lack of generalizability of the primary studies by their conjoint interpretation, was used to identify key themes and synthesize them. NVivo qualitative analysis software was used. We identified 40 articles (> 600 oncologists) from 12 countries and assessed their quality as good according to the Critical Appraisal Skills Programme (CASP). Two main themes emerged: the patient-oncologist encounter during the breaking of bad news, comprising essential aspects of the communication, including the process of dealing with emotions; and external factors shaping the patient-oncologist encounter, composed of factors that influence the announcement beyond the physician-patient relationship: the family, systemic and institutional factors, and cultural factors. Breaking bad news is a balancing act that requires oncologists to adapt continually to different factors: their individual relationships with the patient, the patient's family, the institutional and systemic environment, and the cultural milieu. Extending the development of the ability to personalize and adapt therapeutic treatment to this realm of communications would be a major step forward from the stereotyped way that oncologists are currently trained in communication skills. © 2015 by American Society of Clinical Oncology.

  11. 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...

  12. Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk Methods Integration and Evaluation Program (RMIEP)

    International Nuclear Information System (INIS)

    Payne, A.C. Jr.; Eide, S.A.; LaChance, J.C.; Whitehead, D.W.

    1992-10-01

    This volume presents the results of the initiating event and accident sequence delineation analyses of the LaSalle Unit II nuclear power plant performed as part of the Level III PRA being performed by Sandia National Laboratories for the Nuclear Regulatory Commission. The initiating event identification included a thorough review of extant data and a detailed plant specific search for special initiators. For the LaSalle analysis, the following initiating events were defined: eight general transients, ten special initiators, four LOCAs inside containment, one LOCA outside containment, and two interfacing LOCAs. Three accident sequence event trees were constructed: LOCA, transient, and ATWS. These trees were general in nature so that a tree represented all initiators of a particular type (i.e., the LOCA tree was constructed for evaluating small, medium, and large LOCAs simultaneously). The effects of the specific initiators on the systems and the different success criteria were handled by including the initiating events directly in the system fault trees. The accident sequence event trees were extended to include the evaluation of containment vulnerable sequences. These internal event accident sequence event trees were also used for the evaluation of the seismic, fire, and flood analyses

  13. Missed retinal breaks in rhegmatogenous retinal detachment

    Directory of Open Access Journals (Sweden)

    Brijesh Takkar

    2016-12-01

    Full Text Available AIM: To evaluate the causes and associations of missed retinal breaks (MRBs and posterior vitreous detachment (PVD in patients with rhegmatogenous retinal detachment (RRD. METHODS: Case sheets of patients undergoing vitreo retinal surgery for RRD at a tertiary eye care centre were evaluated retrospectively. Out of the 378 records screened, 253 were included for analysis of MRBs and 191 patients were included for analysis of PVD, depending on the inclusion criteria. Features of RRD and retinal breaks noted on examination were compared to the status of MRBs and PVD detected during surgery for possible associations. RESULTS: Overall, 27% patients had MRBs. Retinal holes were commonly missed in patients with lattice degeneration while missed retinal tears were associated with presence of complete PVD. Patients operated for cataract surgery were significantly associated with MRBs (P=0.033 with the odds of missing a retinal break being 1.91 as compared to patients with natural lens. Advanced proliferative vitreo retinopathy (PVR and retinal bullae were the most common reasons for missing a retinal break during examination. PVD was present in 52% of the cases and was wrongly assessed in 16%. Retinal bullae, pseudophakia/aphakia, myopia, and horse shoe retinal tears were strongly associated with presence of PVD. Traumatic RRDs were rarely associated with PVD. CONCLUSION: Pseudophakic patients, and patients with retinal bullae or advanced PVR should be carefully screened for MRBs. Though Weiss ring is a good indicator of PVD, it may still be over diagnosed in some cases. PVD is associated with retinal bullae and pseudophakia, and inversely with traumatic RRD.

  14. Development and application of an entrainment model for the PWR U-tube steam generators for main steam line break analysis

    International Nuclear Information System (INIS)

    Song, Dong-Soo; Park, Young-Chan

    2004-01-01

    The purpose of this paper is to present the analyses that were performed to develop and use an entrainment model for pressurized water reactor U-tube steam generators (SG) for main steam line break (MSLB) analyses. The entrainment model was developed using the RETRAN-3D computer program, and the model was benchmarked against experimental data of moisture carryover during a simulated MSLB accident. The application methodology was also developed to incorporate into the MSLB mass and energy release calculations for Kori Unit 1. This methodology utilizes LOFTRAN and RETRAN-3D codes in an iterative sequence of cases in which the LOFTRAN nuclear steam supply system model provides boundary conditions for the RETRAN-3D broken loop steam generator model, and the RETRAN-3D model provides the entrainment data that is input back into the LOFTRAN model. FORTRAN programs were developed to facilitate the sequencing of these iterative calculations. As a result of applying the entrainment model to Kori Unit 1, the temperature calculated inside Containment during MSLB accident using the CONTEMP-LT computer program decreased by about 25degC. Consequently this entrainment model provides a significant benefit by decreasing the temperature envelop for environment qualification as well as decreasing the peak Containment pressure. (author)

  15. On breaks of the Indian monsoon

    Indian Academy of Sciences (India)

    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 ...

  16. Report of Subcommittee on Investigation of Empirical Formulas Concerning Reactor Safety Analysis

    International Nuclear Information System (INIS)

    1981-01-01

    For the safety evaluation when nuclear power stations are going to be installed, the transient state following the various abnormal events occurring in reactor facilities is analyzed. In the case of light water reactors, the most important safety analysis is related to the loss of coolant accident caused by the break of primary cooling pipes. At the time of the analysis, the safety analysis codes based on various models are used, and many empirical formulas and correlation formulas concerning heat transfer are included in them. In Japan Society of Mechanical Engineers, for the purpose of investigating and collecting the empirical and correlation formulas on heat transferring flow related to reactor safety analysis and contributing to the perfection of the basic data for reactor safety evaluation, the ''Subcommittee on investigation of empirical formulas concerning reactor safety analysis'' was established in February, 1978. This is the report on the results of investigation by the subcommittee, intended for general use. The correlation formulas on heat transfer in the analysis of LOCA in light water reactors are taken up, and classified into eight subjects. This report is significant for those who are going to develop the analysis codes or who engage in the safety analysis. (Kako, I.)

  17. 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

  18. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

    2014-05-15

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

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

    International Nuclear Information System (INIS)

    Gibelli, S.M.O.; Oliveira, L.F.S. de.

    1981-12-01

    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.) [pt

  20. Symmetries and stochastic symmetry breaking in multifractal geophysics: analysis and simulation with the help of the Lévy-Clifford algebra of cascade generators..

    Science.gov (United States)

    Schertzer, D. J. M.; Tchiguirinskaia, I.

    2016-12-01

    Multifractal fields, whose definition is rather independent of their domain dimension, have opened a new approach of geophysics enabling to explore its spatial extension that is of prime importance as underlined by the expression "spatial chaos". However multifractals have been until recently restricted to be scalar valued, i.e. to one-dimensional codomains. This has prevented to deal with the key question of complex component interactions and their non trivial symmetries. We first emphasize that the Lie algebra of stochastic generators of cascade processes enables us to generalize multifractals to arbitrarily large codomains, e.g. flows of vector fields on large dimensional manifolds. In particular, we have recently investigated the neat example of stable Levy generators on Clifford algebra that have a number of seductive properties, e.g. universal statistical and robust algebra properties, both defining the basic symmetries of the corresponding fields (Schertzer and Tchiguirinskaia, 2015). These properties provide a convenient multifractal framework to study both the symmetries of the fields and how they stochastically break the symmetries of the underlying equations due to boundary conditions, large scale rotations and forcings. These developments should help us to answer to challenging questions such as the climatology of (exo-) planets based on first principles (Pierrehumbert, 2013), to fully address the question of the limitations of quasi- geostrophic turbulence (Schertzer et al., 2012) and to explore the peculiar phenomenology of turbulent dynamics of the atmosphere or oceans that is neither two- or three-dimensional. Pierrehumbert, R.T., 2013. Strange news from other stars. Nature Geoscience, 6(2), pp.8183. Schertzer, D. et al., 2012. Quasi-geostrophic turbulence and generalized scale invariance, a theoretical reply. Atmos. Chem. Phys., 12, pp.327336. Schertzer, D. & Tchiguirinskaia, I., 2015. Multifractal vector fields and stochastic Clifford algebra

  1. Isospin breaking from diquark clustering

    Science.gov (United States)

    Gibbs, W. R.; Dedonder, Jean-Pierre

    2017-09-01

    Background: Although SU(2) isospin symmetry is generally assumed in the basic theory of the strong interaction, a number of significant violations have been observed in scattering and bound states of nucleons. Many of these violations can be attributed to the electromagnetic interaction but the question of how much of the violation is due to it remains open. Purpose: To establish the connection between diquark clustering in the two-nucleon system and isospin breaking from the Coulomb interaction between the members of diquark pairs. Method: A schematic model based on clustering of quarks in the interior of the confinement region of the two-nucleon system is introduced and evaluated. In this model the Coulomb interaction is the source of all isospin breaking. It draws on a picture of the quark density based on the diquark-quark model of hadron structure which has been investigated by a number of groups. Results: The model produces three isospin breaking potentials connecting the unbroken value of the low-energy scattering amplitude to those of the p p , n n , and n p singlet channels. A simple test of the potentials in the three-nucleon energy difference problem yields results in agreement with the known binding energy difference. Conclusion: The illustrative model suggests that the breaking seen in the low-energy nucleon-nucleon (NN) interaction may be understood in terms of the Coulomb force between members of diquark clusters. It allows the prediction of the charge symmetry breaking interaction and the n n scattering length from the well measured n p singlet scattering length. Values of the n n scattering length around -18 fm are favored. Since the model is based on the quark picture, it can be easily extended, in the SU(3) limit, to calculate isospin breaking in the strange sector in the corresponding channels. A natural consequence of isospin breaking from diquark clustering is that the breaking in the strange sector, as measured by the separation energy

  2. 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

  3. 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.

  4. Geochemical and Hydrologic Controls of Copper-Rich Surface Waters in the Yerba Loca-Mapocho System

    Science.gov (United States)

    Pasten, P.; Montecinos, M.; Coquery, M.; Pizarro, G. E.; Abarca, M. I.; Arce, G. J.

    2015-12-01

    Andean watersheds in Northern and Central Chile are naturally enriched with metals, many of them associated to sulfide mineralizations related to copper mining districts. The natural and anthropogenic influx of toxic metals into drinking water sources pose a sustainability challenge for cities that need to provide safe water with the smallest footprint. This work presents our study of the transformations of copper in the Yerba Loca-Mapocho system. Our sampling campaign started from the headwaters at La Paloma Glacier and continues to the inlet of the San Enrique drinking water treatment plant, a system feeding municipalities in the Eastern area of Santiago, Chile. Depending on the season, total copper concentrations go as high as 22 mg/L for the upper sections, which become diluted to TXRF (total reflection X ray fluorescence) and XRD (X-ray diffraction). Major elements detected in the precipitates were Al (200 g/kg), S (60 g/kg), and Cu (6 g/kg). Likely solid phases include hydrous amorphous phases of aluminum hydroxides and sulfates, and copper hydroxides/carbonates. Efforts are undergoing to find the optimal mixing ratios between the acidic stream and more alkaline streams to maximize attenuation of dissolved copper. The results of this research could be used for enhancing in-stream natural attenuation of copper and reducing treatment needs at the drinking water facility. Acknowledgements to Fondecyt 1130936 and Conicyt Fondap 15110020

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