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Sample records for break loca experiment

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. 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.

  15. 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

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. 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)

  1. 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

  2. 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

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

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-06-01

    Full Text Available ABSTRAK ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA. Kecelakaan yang diakibatkan oleh kehilangan pendingin (loss of coolant accident / LOCA dari sistem reaktor merupakan kejadian dasar desain yang tetap diantisipasi dalam desain reaktor daya yang mengadopsi teknologi Generasi II hingga IV. LOCA ukuran kecil (small break LOCA memiliki dampak yang lebih signifikan terhadap keselamatan dibandingkan LOCA ukuran besar (large break LOCA seperti terlihat pada kejadian Three-Mile Island (TMI. Fokus makalah adalah pada analisis small break LOCA pada reaktor daya maju Generasi III+ yaitu AP1000 dengan mensimulasikan tiga kejadian pemicu yaitu membukanya katup Automatic Depressurization System (ADS secara tak disengaja, putusnya salah satu pipa Direct Vessel Injection (DVI secara double-ended, dan putusnya pipa lengan dingin dengan diameter bocoran 10 inci. Metode yang digunakan adalah simulasi kejadian pada model AP1000 yang dikembangkan secara mandiri menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Dampak yang ingin dilihat adalah kondisi teras selama terjadinya small break LOCA yang terdiri dari pembentukan mixture level dan transien temperatur kelongsong bahan bakar. Hasil simulasi menunjukkan bahwa mixture level untuk semua kejadian small break LOCA berada di atas tinggi teras aktif yang menunjukkan tidak terjadinya core uncovery. Adanya mixture level berpengaruh pada transien temperatur kelongsong yang menurun dan menunjukkan pendinginan bahan bakar yang efektif. Hasil di atas juga identik dengan hasil perhitungan program lain yaitu NOTRUMP. Keefektifan pendinginan teras juga disebabkan oleh berfungsinya injeksi pendingin melalui fitur keselamatan pasif yang menjadi ciri reaktor daya AP1000. Secara keseluruhan, hasil analisis menunjukkan model AP1000 yang telah dikembangkan dengan RELAP5 dapat digunakan untuk keperluan analisis kecelakaan dasar desain pada reaktor daya maju AP1000. Kata kunci: analisis

  4. 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

    viewpoints and experiences of the largest utility. The presentation summarized the experi ence data for known degradation mechanisms. The EDF presentation also covered leak before-break concept which is one of the integral elements (along with the in-service inspections and leak detection) in redefining the LB LOCA. The SKI presentation discussed the experience with the degradation specific in-service inspection programs. The presentation also covered a number of efforts that are currently underway to develop risk-informed ISI program. It should be recognized that any risk informed approach to in-service inspection programs must be based on an extensive experience, which at present will probably not cover every possible degradation mechanism. The Framatome presentation described the standardize criteria, methods, and procedures for assuring the design and operational adequacy of reactor coolant pressure boundary leak detection systems used in plants. The presentation also covered a new system based on the measurements of local humidity. 3. What are possible new definitions for the LB-LOCA? What are their implications for current and future reactors? Two papers were presented which highlighted different approaches to incorporating a change in the LB-LOCA definition into a plant. The first paper, 'Slovak approach during the gradual upgrading of Bohunice V 1', by Mr Kliment, described a programme of back fits to an operating plant, to increase the DBA LB-LOCA from 32 mm equivalent to 200 mm. The second paper, 'Westinghouse Owner's Group Large Break Loca Redefinition Program', presented by Mr Bastien, described the simplifications of design and operation that would be achieved if the DBA LB-LOCA were (significantly) reduced in size

  5. 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.

  6. 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.

  7. 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)

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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)

  14. 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)

  15. 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)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  17. 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

  18. 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.)

  19. 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)

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

    International Nuclear Information System (INIS)

    1998-01-01

    hot leg region of a PWR which may occur in the reflood phase during a large cold leg break LOCA with cold leg ECC injection. Boundary Conditions: There was no ECC injection. Steam and saturated water were injected into the core simulator. The cold leg break valve was open. The hot leg break valve and all pump simulators were partially open to establish the desired loop flow resistances. Significant Findings: During an initial time period - about 40 s - the water that was injected into the core simulator was distributed within the upper plenum and the adjacent loops. After these regions were 'saturated' with water, the water was carried by the steam into the steam generator simulator inlet plenum and into the cyclones of these simulators where it was separated. The amount of water separated by the cyclones in the steam generator simulators was in good agreement with the amount of water injected into the core simulator after the initial time period of about 40 s. After termination of the steam injection into the core simulator, the water flow back into the core region was in quantitative agreement with the amount of water that was distributed during the initial time periods. During the entire test phase there was no significant pool formation above the tie plate. 3 - Experimental limitations or shortcomings: The system operating pressure was limited to 20 bar. So for some experiments pressure scaling was necessary

  1. 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.

  2. 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.

  3. 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 ...

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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

  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. 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

  11. 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.

  12. 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)

  13. Results of the mock-up experiment on partial LOCA

    International Nuclear Information System (INIS)

    Dreier, J.; Winkler, H.

    1985-01-01

    A mockup experiment has been performed to verify the heat transfer model for a partial loss of coolant accident in the swimming pool reactor SAPHIR. Three coolant channels with the same dimensions as in a SAPHIR fuel element were simulated using four electrically heated plates. For a water level such that the heated plates are partially submerged, plate temperatures remain below 160 deg. C for plate powers of up to 650 W. For water levels low enough to just block the channels, plate temperatures of 400 deg. C are reached for plate powers as low as 60 W. Details of the experiment and further results are discussed. (author)

  14. 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)

  15. 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)

  16. 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

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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)

  3. 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)

  4. 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.

  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. 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

  7. 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

  8. 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)

  9. 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)

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

    International Nuclear Information System (INIS)

    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)

  11. 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)

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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)

  17. 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

  18. Implications of dynamical symmetry breaking for high energy experiments

    International Nuclear Information System (INIS)

    Ali, A.

    1981-06-01

    A scenario of dynamical symmetry breaking as an alternative to the canonical Higgs mechanism with elementary spin-O fields is described, and its implications for high energy experiments contrasted with those of the canonical theory. The potential role of e + e - annihilation physics in unravelling the nature of spontaneous symmetry breaking is emphasized. (orig.)

  19. 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)

  20. 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)

  1. 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)

  2. 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

  3. 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)

  4. 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)

  5. Paediatric SpRs' experiences of breaking bad news.

    Science.gov (United States)

    Horwitz, N; Ellis, J

    2007-09-01

    To ascertain the level of support and training available to paediatric specialist registrars (SpRs) in breaking bad news and their self-reported confidence in this task. A questionnaire-based survey. Paediatric SpRs working in North Thames region. Specialist registrars (n = 206) were sent a questionnaire relating to the level of support and training available to them in breaking bad news and their attitudes to this task. A repeat questionnaire was sent out 2 weeks later. The response rate was 54.9%. The study sample included 78 females and 34 males. The median year of qualification was 1995 [interquartile range (IQR) 1993-1997] and the median year of Calman training was Year 3 (IQR 2-4). Only 15.9% of participants had guidelines where they worked and 91.2% had received training in breaking bad news. Median self-perceived confidence in breaking bad news was rated as 4 out of 5. Only 21.2% of all respondents had both disclosed a diagnosis of Down syndrome and received feedback on their performance from their seniors. Few SpRs were able to adhere to all evidence-based recommendations for breaking bad news. Most SpRs had received training in breaking bad news and self-reported confidence in this skill was high, although their hands-on experience was limited. Recent research shows, however, that parental dissatisfaction with the way in which bad news is broken remains high. The potential discrepancy between self-reported confidence and actual competence casts doubt on the value of self-evaluation.

  6. 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

  7. 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

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

    International Nuclear Information System (INIS)

    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)

  9. 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

  10. 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)

  11. 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

  12. Simulation of breaking waves using the high-order spectral method with laboratory experiments: wave-breaking energy dissipation

    Science.gov (United States)

    Seiffert, Betsy R.; Ducrozet, Guillaume

    2018-01-01

    We examine the implementation of a wave-breaking mechanism into a nonlinear potential flow solver. The success of the mechanism will be studied by implementing it into the numerical model HOS-NWT, which is a computationally efficient, open source code that solves for the free surface in a numerical wave tank using the high-order spectral (HOS) method. Once the breaking mechanism is validated, it can be implemented into other nonlinear potential flow models. To solve for wave-breaking, first a wave-breaking onset parameter is identified, and then a method for computing wave-breaking associated energy loss is determined. Wave-breaking onset is calculated using a breaking criteria introduced by Barthelemy et al. (J Fluid Mech https://arxiv.org/pdf/1508.06002.pdf, submitted) and validated with the experiments of Saket et al. (J Fluid Mech 811:642-658, 2017). Wave-breaking energy dissipation is calculated by adding a viscous diffusion term computed using an eddy viscosity parameter introduced by Tian et al. (Phys Fluids 20(6): 066,604, 2008, Phys Fluids 24(3), 2012), which is estimated based on the pre-breaking wave geometry. A set of two-dimensional experiments is conducted to validate the implemented wave breaking mechanism at a large scale. Breaking waves are generated by using traditional methods of evolution of focused waves and modulational instability, as well as irregular breaking waves with a range of primary frequencies, providing a wide range of breaking conditions to validate the solver. Furthermore, adjustments are made to the method of application and coefficient of the viscous diffusion term with negligible difference, supporting the robustness of the eddy viscosity parameter. The model is able to accurately predict surface elevation and corresponding frequency/amplitude spectrum, as well as energy dissipation when compared with the experimental measurements. This suggests the model is capable of calculating wave-breaking onset and energy dissipation

  13. 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

  14. Simulation of breaking waves using the high-order spectral method with laboratory experiments: Wave-breaking onset

    Science.gov (United States)

    Seiffert, Betsy R.; Ducrozet, Guillaume; Bonnefoy, Félicien

    2017-11-01

    This study investigates a wave-breaking onset criteria to be implemented in the non-linear potential flow solver HOS-NWT. The model is a computationally efficient, open source code, which solves for the free surface in a numerical wave tank using the High-Order Spectral (HOS) method. The goal of this study is to determine the best method to identify the onset of random single and multiple breaking waves over a large domain at the exact time they occur. To identify breaking waves, a breaking onset criteria based on the ratio of local energy flux velocity to the local crest velocity, introduced by Barthelemy et al. (2017) is selected. The breaking parameter is uniquely applied in the numerical model in that calculations of the breaking onset criteria ratio are not made only at the location of the wave crest, but at every point in the domain and at every time step. This allows the model to calculate the onset of a breaking wave the moment it happens, and without knowing anything about the wave a priori. The application of the breaking criteria at every point in the domain and at every time step requires the phase velocity to be calculated instantaneously everywhere in the domain and at every time step. This is achieved by calculating the instantaneous phase velocity using the Hilbert transform and dispersion relation. A comparison between more traditional crest-tracking techniques shows the calculation of phase velocity using Hilbert transform at the location of the breaking wave crest provides a good approximation of crest velocity. The ability of the selected wave breaking criteria to predict single and multiple breaking events in two dimensions is validated by a series of large-scale experiments. Breaking waves are generated by energy focusing and modulational instability methods, with a wide range of primary frequencies. Steep irregular waves which lead to breaking waves, and irregular waves with an energy focusing wave superimposed are also generated. This set of

  15. 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

  16. 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.

  17. 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

  18. Promoting Resilience: Breaking the Intergenerational Cycle of Adverse Childhood Experiences.

    Science.gov (United States)

    Woods-Jaeger, Briana A; Cho, Bridget; Sexton, Chris C; Slagel, Lauren; Goggin, Kathy

    2018-02-01

    Adverse childhood experiences (ACEs), including trauma exposure, parent mental health problems, and family dysfunction, put children at risk for disrupted brain development and increased risk for later health problems and mortality. These negative effects may be prevented by resilience promoting environments that include protective caregiving relationships. We sought to understand (1) parents' experiences of ACEs, (2) the perceived impact on parenting, (3) protective factors that buffer ACEs potential negative impact, and (4) supports and services that can reduce the number and severity of ACEs and promote resilience among children exposed to early adversity. We conducted in-depth qualitative interviews with 11 low-income, urban parents of young children who had experienced ACEs. Interviews were analyzed for emergent themes and shared with parents from the community to ensure relevance and proper interpretation. Themes from these interviews describe the potential intergenerational cycle of ACEs and key factors that can break that cycle, including parent aspirations to make children's lives better and parent nurturance and support. Parents' suggestions for intervention are also presented. Our findings illuminate protective factors and family strengths that are important to build upon when developing and implementing interventions to promote resilience among parents and children exposed to early adversity. This study benefits from highly ecologically valid data obtained from low-socioeconomic status, racial/ethnic minority parents through one-on-one in-depth interviews and interpreted with the aid of community stakeholders through a community-based participatory research approach.

  19. 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

  20. 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

  1. Results of recent LOFT experiments

    International Nuclear Information System (INIS)

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

    1982-01-01

    Five experiments were performed in the Loss-of-Fluid Test (LOFT) facility during the past year. The experiments conducted spanned a wide range of potential accident scenarios, including large and small break loss-of-coolant accidents (LOCAs), control rod withdrawal accidents, uncontrolled boron dilution, and anticipated transients without scram (ATWS). This summary describes these experiments and presents results available from the experiments and experiment prediction calculations. A brief overview is given for the remaining experiment planned in the LOFT Program

  2. Lithium-lead/water interaction. Large break experiments

    International Nuclear Information System (INIS)

    Savatteri, C.; Gemelli, A.

    1991-01-01

    One current concept in fusion blanket module design is to utilize water as coolant and liquid lithium-lead as breeding/neutron-multiplier material. Considering the possibility of certain off-normal events, it is possible that water leakage into the liquid metal may occur due to a tube rupture. The lithium-lead/water contact can lead to a thermal and chemical reaction which should provoke an intolerable pressure increase in the blanket module. For realistic simulation of such in-blanket events, the Blanket Safety Test (BLAST) facility has been built. It simulates the transient event by injecting subcooled water under high pressure into a stagnant pool of about 500 kg liquid Pb-17Li. Eight fully instrumented large break tests were carried out under different conditions. The aim of the experiments is to study the chemical and thermal process and particularly: The pressurization history of the reaction vessel, the formation and deposition of the reaction products, the identification and propagation of the reaction zones and the temperature transient in the liquid metal. In this paper the results of all tests performed are presented and discussed. (orig.)

  3. Leak Before Break concept based on Framatome experience

    International Nuclear Information System (INIS)

    Bhandari, S.; Franco, Ch.

    2000-01-01

    For the last 20 years, Framatome has been engaged in the development of the Leak- Before-Break concept at several levels: in the framework of its own R and D studies, in the framework of the three-party co-operative studies between CEA, EDF and Framatome, at an european level in the joint studies with Siemens, and -at an international level in the framework of several research programmes. For the present 900 MWe French reactors, the pipe integrity has been investigated using the LBB as a defense-in depth concept. As to the future, the application of LBB concept is foreseen on the Primary Circuit of the French/German European Pressurized Reactor (EPR) right at the stage of basic design. In order to be able to do that, a number of major events have significantly contributed to provide Framatome with deepened knowledge on LBB methodology and applications. This paper aims at establishing a progress statement on Framatome experience concerning these applications. It is divided in five main parts dealing with: the technical experience gained by Framatome from its participation to International Research Programs such as the Degraded Piping Program Phases (DP3) and the International Piping Integrity Research Group Program (IPIRG), the application of LBB for the primary piping of the Russian VVER NPP's of the type 440/230 in the framework of the TACIS Program, the LBB demonstration implemented by Framatome on the primary circuit of two Belgian plants, the LBB application for the primary circuit of 900 MWe French reactors and the work being done in the framework of the EPR project mentioned above. (author)

  4. [Breaking bad news in oncology: the Belgian experience].

    Science.gov (United States)

    Delevallez, F; Lienard, A; Gibon, A-S; Razavi, D

    2014-10-01

    Breaking bad news is a complex and frequent clinical task for physicians working in oncology. It can have a negative impact on patients and their relatives who are often present during breaking bad news consultations. Many factors influence how the delivery of bad news will be experienced especially the communication skills used by physicians. A three-phase process (post-delivery phase, delivery phase, pre-delivery phase) has been developed to help physician to handle this task more effectively. Communication skills and specific breaking bad news training programs are both necessary and effective. A recent study conducted in Belgium has shown their impact on the time allocated to each of the three phases of this process, on the communication skills used, on the inclusion of the relative in the consultation and on physicians' physiological arousal. These results underscore the importance of promoting intensive communication skills and breaking bad news training programs for health care professionals.

  5. 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

  6. Review of LOFT [Loss-of-Fluid Test] large break experiments

    International Nuclear Information System (INIS)

    Modro, S.M.; Aksan, S.N.; Berta, V.T.; Wahba, A.B.

    1989-10-01

    Six non-nuclear and five nuclear large break loss-of-coolant experiments were performed in the Loss-of-Fluid Test (LOFT) PWR facility at the Idaho National Engineering Laboratory. These experiments provided a large amount of data necessary for evaluation and refinement of reactor system computer codes and had major impact on the understanding of large break loss-of-coolant accidents. An overview of these nuclear large break experiments performed under NRC and OECD LOFT programs is given and the major research results are presented. 55 refs., 89 figs., 5 tabs

  7. 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)

  8. Experiments and computation of onshore breaking solitary waves

    DEFF Research Database (Denmark)

    Jensen, A.; Mayer, Stefan; Pedersen, G.K.

    2005-01-01

    This is a combined experimental and computational study of solitary waves that break on-shore. Velocities and accelerations are measured by a two-camera PIV technique and compared to theoretical values from an Euler model with a VOF method for the free surface. In particular, the dynamics of a so......-called collapsing breaker is scrutinized and the closure between the breaker and the beach is found to be akin to slamming. To the knowledge of the authors, no velocity measurements for this kind of breaker have been previously reported....

  9. 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

  10. 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

  11. Symmetry breaking on density in escaping ants: experiment and alarm pheromone model.

    Directory of Open Access Journals (Sweden)

    Geng Li

    Full Text Available The symmetry breaking observed in nature is fascinating. This symmetry breaking is observed in both human crowds and ant colonies. In such cases, when escaping from a closed space with two symmetrically located exits, one exit is used more often than the other. Group size and density have been reported as having no significant impact on symmetry breaking, and the alignment rule has been used to model symmetry breaking. Density usually plays important roles in collective behavior. However, density is not well-studied in symmetry breaking, which forms the major basis of this paper. The experiment described in this paper on an ant colony displays an increase then decrease of symmetry breaking versus ant density. This result suggests that a Vicsek-like model with an alignment rule may not be the correct model for escaping ants. Based on biological facts that ants use pheromones to communicate, rather than seeing how other individuals move, we propose a simple yet effective alarm pheromone model. The model results agree well with the experimental outcomes. As a measure, this paper redefines symmetry breaking as the collective asymmetry by deducing the random fluctuations. This research indicates that ants deposit and respond to the alarm pheromone, and the accumulation of this biased information sharing leads to symmetry breaking, which suggests true fundamental rules of collective escape behavior in ants.

  12. 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

  13. 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.

  14. Scaling, experiment, and code assessment on an integral testing facility

    International Nuclear Information System (INIS)

    Yang, J.; Choi, S.W.; Lim, J.; Lee, D.Y.; Rassame, S.; Hibiki, T.; Ishii, M.

    2011-01-01

    A series of integral tests simulating different types of Loss-Of-Coolant Accidents (LOCAs) for new Boiling Water Reactor (BWR) design were conducted on an integral test facility (Purdue University Multi-Dimensional Integral Test Assembly, PUMA) facility. The PUMA facility was built with a scaling methodology addressing both the conservation principles and constitutive laws. A systemic study about the safety evaluation of the advanced passively safe BWR design has been performed with the collaboration of experiments on the scaled-down test facility and RELAP5/Mod3.3 code simulation. Various types of LOCA tests were performed, such as Main Steam Line Break (MSLB), Bottom Drain Line Break (BDLB), Gravity-Driven Line Break (GDLB), and Feed Water Line Break (FWLB). (author)

  15. Large Scale Experiments on the Interaction of a Caisson Breakwater with Breaking Waves

    DEFF Research Database (Denmark)

    Stagonas, Dimitris; Marzeddu, Andrea; Buccino, Mariano

    2014-01-01

    Tests looking at the interaction of a caisson breakwater with steep, breaking waves are outlined here. 4 different wave generation methodologies were employed allowing for experiments with regular, irregular, focused and tailored made waves. The emphasis, however, is given in tests with focused...... waves, which resulted in impulsive conditions at the face of the caisson. Amongst our objectives was to look at the mechanisms occurring when a wave breaks at the structure and to investigate the validity of tactile pressure sensors. As such, for all experiments, pressure, force and surface elevation...

  16. 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

  17. Computer simulations of a 1/5-scale experiment of a Mark I boiler water reactor pressure-suppression system under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1978-01-01

    The CHAMP computer code was employed to simulate a plane-geometry cross section of a Mark I boiling water reactor toroidal pressure suppression system air discharge experiment under hypothetical loss-of-coolant accident conditions. The experiments were performed at the Lawrence Livermore Laboratory on a 1 / 5 -scale model of the Peach Bottom Nuclear Power Plant

  18. RELAP5/MOD2 post-test calculation of the OECD LOFT experiment LP-SB-2

    International Nuclear Information System (INIS)

    Perez, J.; Mendizabal, R.

    1992-04-01

    This document presents the analysis of the OECD LOFT LP-SB-2 Experiment performed by the Consejo de Seguridad Nuclear of Spain working group making use of RELAP5/MOD2 in the frame of the Spanish LOFT Project. LB-SB-2 experiment studies the effect of a delayed pump trip in a small break LOCA scenario with a 3-inch equivalent diameter break in the hot leg of a commercial PWR

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

    International Nuclear Information System (INIS)

    Glaeser, H.G.

    2004-01-01

    The first formulation of the USA Code of Federal Regulations (CFR) 10CFR50 with applicable sections specific to NPP licensing requirements was released 1976. Over a decade later 10CFR 50.46 allowed the use of BE codes instead of conservative code models but uncertainties have to be identified and quantified. Guidelines were released that described interpretations developed over the intervening years that are applicable. Other countries established similar conservative procedures and acceptance criteria. Because conservative methods were used to calculate the peak values of key parameters, such as peak clad temperature (PCT), it was always acknowledged that a large margin, between the 'conservative' calculated value and the 'true' value, existed. Beside USA, regulation in other countries, like Germany, for example, allowed that the state of science and technology is applied in licensing. I.e. the increase of experimental evidence and progress in code development during time could be used. There was no requirement to apply a pure evaluation methodology with licensed assumptions and frozen codes. The thermal-hydraulic system codes became more and more best-estimate codes based on comprehensive validation. This development was and is possible because the rules and guidelines provide the necessary latitude to consider further development of safety technology. Best estimate codes are allowed to be used in licensing in combination with conservative initial and boundary conditions. However, uncertainty quantification is not required. Since some of the initial and boundary conditions are more conservative compared with those internationally used (e.g. 106% reactor power instead 102%, a single failure plus a non-availability due to preventive maintenance is assumed, etc.) it is claimed that the uncertainties of code models are covered. Since many utilities apply for power increase, calculation results come closer to some licensing criteria. The situation in German licensing 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

  20. The experience of breaks in psychotherapy with children who suffered massive early trauma

    DEFF Research Database (Denmark)

    Grünbaum, Liselotte

    of the child`s and the therapist`s experience. The applied method was a qualitative case study applying inductive-deductive principles in line with Interpretative Phenomenological Analysis. The case material included the therapist`s session notes, lifespan reports from other informants, follow-up interviews...... by a comparable investigation of summer vacation breaks in student-therapist`s supervised, concluded therapies. Findings highlighted 1) links between breaks and the eruption of hostile parental and sibling figures in mind; 2) links between the emotional quality of sibling and parental figures; 3) differences...

  1. 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

  2. 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)

  3. 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.

  4. 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)

  5. 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)

  6. 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)

  7. 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)

  8. Gas transfer under breaking waves: experiments and an improved vorticity-based model

    Directory of Open Access Journals (Sweden)

    V. K. Tsoukala

    2008-07-01

    Full Text Available In the present paper a modified vorticity-based model for gas transfer under breaking waves in the absence of significant wind forcing is presented. A theoretically valid and practically applicable mathematical expression is suggested for the assessment of the oxygen transfer coefficient in the area of wave-breaking. The proposed model is based on the theory of surface renewal that expresses the oxygen transfer coefficient as a function of both the wave vorticity and the Reynolds wave number for breaking waves. Experimental data were collected in wave flumes of various scales: a small-scale experiments were carried out using both a sloping beach and a rubble-mound breakwater in the wave flume of the Laboratory of Harbor Works, NTUA, Greece; b large-scale experiments were carried out with a sloping beach in the wind-wave flume of Delft Hydraulics, the Netherlands, and with a three-layer rubble mound breakwater in the Schneideberg Wave Flume of the Franzius Institute, University of Hannover, Germany. The experimental data acquired from both the small- and large-scale experiments were in good agreement with the proposed model. Although the apparent transfer coefficients from the large-scale experiments were lower than those determined from the small-scale experiments, the actual oxygen transfer coefficients, as calculated using a discretized form of the transport equation, are in the same order of magnitude for both the small- and large-scale experiments. The validity of the proposed model is compared to experimental results from other researchers. Although the results are encouraging, additional research is needed, to incorporate the influence of bubble mediated gas exchange, before these results are used for an environmental friendly design of harbor works, or for projects involving waste disposal at sea.

  9. 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.

  10. Experiment data report for LOFT large-break loss-of-coolant experiment L2-5

    International Nuclear Information System (INIS)

    Bayless, P.D.; Divine, J.M.

    1982-08-01

    Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)] commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected

  11. LOCA verification and data bank

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.; Cox, N.D.; Atwood, C.L.; Madden, S.C.; Condie, K.G.

    1979-01-01

    The purpose of this task was to derive local conditions heat transfer parameters and their uncertainties using computer codes and experimentally derived boundary conditions. To accomplish this objective, Semiscale S-02-9 blowdown experiment was used along with the INVERT (an inverse heat conduction code) and RELAP4 (a thermal hydraulic code) codes as the analytical tools. The uncertainties calculated for the local conditions were limited to those introduced by inaccuracies in the experimentally measured boundary conditions. The propagation of the measurement uncertainties through the codes was investigated by varying the code input using statistical methods and a response surface technique

  12. 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

  13. 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.

  14. 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

  15. Constraining Anomaly Mediated Supersymmetry Breaking Framework via Ongoing Muon g-2 Experiment at Brookhaven

    CERN Document Server

    Chattopadhyay, U; Roy, S; PH; Chattopadhyay, Utpal; Ghosh, Dilip Kumar; Roy, Sourov

    2000-01-01

    The ongoing high precision E821 Brookhaven National Laboratory experiment on muon g-2 is promising to probe a theory involving supersymmetry. We have studied the constraints on minimal Anomaly Mediated Supersymmetry Breaking (AMSB) model using the current data of muon g-2 from Brookhaven. A scenario of seeing no deviation from the Standard Model is also considered, within $2\\sigma$ limit of the combined error from the Standard Model result and the Brookhaven predicted uncertainty level. The resulting constraint is found to be complementary to what one obtains from $b \\to s+ \\gamma$ bounds within the AMSB scenario, since only a definite sign of $\\mu$ is effectively probed via $b \\to s+ \\gamma$. A few relevant generic features of the model are also described for disallowed regions of the parameter space.

  16. 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.

  17. 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)

  18. 'Breaking Good News': Neurologists' experiences of discussing SUDEP with patients in Scotland.

    Science.gov (United States)

    Nisbet, Tom; Turbull, Sue; Mulhern, Sharon; Razvi, Saif

    2017-05-01

    Since the findings of a Fatal Accident Inquiry (FAI) in 2010, clinicians working in Scotland have been advised to discuss the risk of Sudden Unexpected Death in Epilepsy (SUDEP) with patients immediately or soon after a diagnosis of epilepsy is made. A thematic analysis was used to describe the experiences discussing SUDEP of 10 clinicians (six Consultant Neurologists and four Neurology Registrars) working in Scotland. Contrary to previous research, clinicians appear to be routinely discussing SUDEP in a standardized fashion with newly diagnosed patients and the FAI appears to have instigated this change in practice. Clinicians are ambivalent about the practice and whether this is a Breaking Bad News (BBN) experience. Clinicians appear to anticipate that patients will be anxious or distressed discussing SUDEP, despite their experiences that patients do not react this way. There are further concerns that the pressure to discuss SUDEP, as a result of the FAI, hinders effective communication of the SUDEP message. Implications for guideline development are discussed. Copyright © 2017 Elsevier Inc. All rights reserved.

  19. 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)

  20. 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

  1. 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.

  2. Breaking-Cas-interactive design of guide RNAs for CRISPR-Cas experiments for ENSEMBL genomes.

    Science.gov (United States)

    Oliveros, Juan C; Franch, Mònica; Tabas-Madrid, Daniel; San-León, David; Montoliu, Lluis; Cubas, Pilar; Pazos, Florencio

    2016-07-08

    The CRISPR/Cas technology is enabling targeted genome editing in multiple organisms with unprecedented accuracy and specificity by using RNA-guided nucleases. A critical point when planning a CRISPR/Cas experiment is the design of the guide RNA (gRNA), which directs the nuclease and associated machinery to the desired genomic location. This gRNA has to fulfil the requirements of the nuclease and lack homology with other genome sites that could lead to off-target effects. Here we introduce the Breaking-Cas system for the design of gRNAs for CRISPR/Cas experiments, including those based in the Cas9 nuclease as well as others recently introduced. The server has unique features not available in other tools, including the possibility of using all eukaryotic genomes available in ENSEMBL (currently around 700), placing variable PAM sequences at 5' or 3' and setting the guide RNA length and the scores per nucleotides. It can be freely accessed at: http://bioinfogp.cnb.csic.es/tools/breakingcas, and the code is available upon request. © The Author(s) 2016. Published by Oxford University Press on behalf of Nucleic Acids Research.

  3. Breaking-Cas—interactive design of guide RNAs for CRISPR-Cas experiments for ENSEMBL genomes

    Science.gov (United States)

    Oliveros, Juan C.; Franch, Mònica; Tabas-Madrid, Daniel; San-León, David; Montoliu, Lluis; Cubas, Pilar; Pazos, Florencio

    2016-01-01

    The CRISPR/Cas technology is enabling targeted genome editing in multiple organisms with unprecedented accuracy and specificity by using RNA-guided nucleases. A critical point when planning a CRISPR/Cas experiment is the design of the guide RNA (gRNA), which directs the nuclease and associated machinery to the desired genomic location. This gRNA has to fulfil the requirements of the nuclease and lack homology with other genome sites that could lead to off-target effects. Here we introduce the Breaking-Cas system for the design of gRNAs for CRISPR/Cas experiments, including those based in the Cas9 nuclease as well as others recently introduced. The server has unique features not available in other tools, including the possibility of using all eukaryotic genomes available in ENSEMBL (currently around 700), placing variable PAM sequences at 5′ or 3′ and setting the guide RNA length and the scores per nucleotides. It can be freely accessed at: http://bioinfogp.cnb.csic.es/tools/breakingcas, and the code is available upon request. PMID:27166368

  4. 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.

  5. 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

  6. 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)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  8. 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.

  9. Leak before break detection-annulus gas monitoring system evolution and operating experience at KGS

    International Nuclear Information System (INIS)

    Jain, D.D.; Sanathkumar, V.V.; Ramamurthy, K.; Nageswara Rao, G.

    2002-01-01

    Full text: Pressurised heavy water reactors (PHWR) at RAPS 1 and 2 and MAPS have provision for detection of pressure tube leak by indirect method. The reactor vessel (calandria) is housed in calandria vault (C/V) filled with air and C/V moisture element indicates the water leak from calandria tube or pressure tube. Further, detection of leak is a cumbersome process. From NAPS onwards, calandria is housed in C/V filled with water, annulus between calandria tube and pressure tube is filled with CO 2 and annulus gas monitoring system (AGMS) is provided by design for detection of any pressure tube leak. The design was improved and AGMS for Kaiga 1 and 2 and RAPS 3 and 4 is having re-circulation mode of operation. The design provides for monitoring dew point of annulus gas (CO 2 ) for indicating the leak and later to identify the pressure tube/calandria tube having leak. The paper deals with operating experience of AGMS at Kaiga generating station (KGS). During the commissioning and initial power operation at KGS, problems were encountered in re-circulation mode. These problems were high radiation field near AGMS piping, high temperature on blower body, blower bearing failure and system leaks. Design modifications were carried out for effective performance of the system for detecting leak before break

  10. A preliminary result of the second np charge symmetry breaking experiment at TRIUMF

    International Nuclear Information System (INIS)

    Zhao, J.; Abegg, R.; Berdoz, A.R.; Birchall, J.; Campbell, J.R.; Davis, C.A.; Delheij, P.P.J.; Gan, L.; Green, P.W.; Greeniaus, L.G.; Healey, D.C.; Helmer, R.; Kolb, N.; Korkmaz, E.a.; Lee, L.; Levy, C.D.P.; Li, J.; Miller, C.A.; Opper, A.K.; Page, S.A.; Postma, H.; Ramsay, W.D.; Soukup, J.; Stinson, G.M.; van Oers, W.T.H.; Zelenski, A.N.

    1995-01-01

    TRIUMF experiment 369, a measurement of charge symmetry breaking in np elastic scattering at 350 MeV, has completed data taking. Scattering asymmetries were measured with a polarized (unpolarized) neutron beam incident on an unpolarized (polarized) frozen spin target. Coincident scattered neutrons and recoil protons were detected by a mirror symmetric detection system in the center of mass angle range from 50 degree--90 degree. A preliminary result for the difference of the zero-crossing angles, where analyzing powers cross zero, is Δθ cm =0.48 degree ±0.08 degree(stat.)±0.08 degree(syst.) based on fits over the angle range 55.8 degree ≤θ cm ≤85.4 degree. The difference of the analyzing powers ΔA≡A n -A p , where the subscripts denote polarized nucleons, was deduced with dA/dθ cm =-1.47x10 -2 deg -1 from phase shift analyses to be [71±12(stat.)±12(syst.)]x10 -4 . It is expected that when all data are analyzed a statistical accuracy of about 7x10 -4 in ΔA will be achieved. copyright 1995 American Institute of Physics

  11. 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.)

  12. A preliminary result of the second np charge symmetry breaking experiment at TRIUMF

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, J. [University of Manitoba, Winnipeg, Manitoba, R3T 2N2 (Canada); Abegg, R. [TRIUMF, 4004 Wesbrook Mall, Vancouver, British Columbia, V6T 2A3 (Canada); Berdoz, A.R.; Birchall, J.; Campbell, J.R. [University of Manitoba, Winnipeg, Manitoba, R3T 2N2 (Canada); Davis, C.A.; Delheij, P.P.J. [TRIUMF, 4004 Wesbrook Mall, Vancouver, British Columbia, V6T 2A3 (Canada); Gan, L. [University of Manitoba, Winnipeg, Manitoba, R3T 2N2 (Canada); Green, P.W.; Greeniaus, L.G. [University of Alberta, Edmonton, Alberta, T6G 2N5 (Canada); Healey, D.C.; Helmer, R. [TRIUMF, 4004 Wesbrook Mall, Vancouver, British Columbia, V6T 2A3 (Canada); Kolb, N.; Korkmaz, E.a. [University of Alberta, Edmonton, Alberta, T6G 2N5 (Canada); Lee, L. [University of Manitoba, Winnipeg, Manitoba, R3T 2N2 (Canada); Levy, C.D.P. [TRIUMF, 4004 Wesbrook Mall, Vancouver, British Columbia, V6T 2A3 (Canada); Li, J. [University of Alberta, Edmonton, Alberta, T6G 2N5 (Canada); Miller, C.A. [TRIUMF, 4004 Wesbrook Mall, Vancouver, British Columbia, V6T 2A3 (Canada); Opper, A.K. [University of Alberta, Edmonton, Alberta, T6G 2N5 (Canada); Page, S.A. [University of Manitoba, Winnipeg, Manitoba, R3T 2N2 (Canada); Postma, H. [TRIUMF, 4004 Wesbrook Mall, Vancouver, British Columbia, V6T 2A3 (Canada); Ramsay, W.D. [University of Manitoba, Winnipeg, Manitoba, R3T 2N2 (Canada); Soukup, J.; Stinson, G.M. [University of Alberta, Edmonton, Alberta, T6G 2N5 (Canada); van Oers, W.T.H. [University of Manitoba, Winnipeg, Manitoba, R3T 2N2 (Canada); Zelenski, A.N. [TRIUMF, 4004 Wesbrook Mall, Vancouver, British Columbia, V6T 2A3 (Canada)

    1995-07-10

    TRIUMF experiment 369, a measurement of charge symmetry breaking in np elastic scattering at 350 MeV, has completed data taking. Scattering asymmetries were measured with a polarized (unpolarized) neutron beam incident on an unpolarized (polarized) frozen spin target. Coincident scattered neutrons and recoil protons were detected by a mirror symmetric detection system in the center of mass angle range from 50{degree}--90{degree}. A preliminary result for the difference of the zero-crossing angles, where analyzing powers cross zero, is {Delta}{theta}{sub cm}=0.48{degree}{plus_minus}0.08{degree}({ital stat}.){plus_minus}0.08{degree}({ital syst}.) based on fits over the angle range 55.8{degree}{le}{theta}{sub cm}{le}85.4{degree}. The difference of the analyzing powers {Delta}{ital A}{equivalent_to}{ital A}{sub {ital n}}{minus}{ital A}{sub {ital p}}, where the subscripts denote polarized nucleons, was deduced with {ital dA}/{ital d}{theta}{sub cm}={minus}1.47{times}10{sup {minus}2} deg{sup {minus}1} from phase shift analyses to be [71{plus_minus}12({ital stat}.){plus_minus}12({ital syst}.)]{times}10{sup {minus}4}. It is expected that when all data are analyzed a statistical accuracy of about 7{times}10{sup {minus}4} in {Delta}{ital A} will be achieved. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}.

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. 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

  19. Hele-Shaw beach creation by breaking waves: a mathematics-inspired experiment

    NARCIS (Netherlands)

    Thornton, Anthony Richard; van der Horn, Avraham/Bram; van der Horn, Avraham J.; Gagarina, Elena; Zweers, Wout; van der Meer, Roger M.; Bokhove, Onno

    2014-01-01

    Fundamentals of nonlinear wave-particle interactions are studied experimentally in a Hele-Shaw configuration with wave breaking and a dynamic bed. To design this configuration, we determine, mathematically, the gap width which allows inertial flows to survive the viscous damping due to the side

  20. Breaking down IT silos: a "connected" way to improve customer experience and the bottom line.

    Science.gov (United States)

    Hallowell, Bruce; Turisco, Frances

    2009-03-01

    Hospitals can provide customer service like Amazon.com without purchasing new technology. Making technology interactive requires sharing patient data across applications and enhancing existing IT with decision support. Breaking down technology silos between hospital and outpatient care provider systems significantly improves efficiency, lowers costs, and speeds care delivery.

  1. 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

  2. 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)

  3. 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

  4. 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

  5. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mendler, O J; Takeuchi, K; Young, M Y

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

  6. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    International Nuclear Information System (INIS)

    Mendler, O.J.; Takeuchi, K.; Young, M.Y.

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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

  12. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  14. 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

  15. 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.

  16. Can cognitive activities during breaks in repetitive manual work accelerate recovery from fatigue? A controlled experiment.

    Directory of Open Access Journals (Sweden)

    Svend Erik Mathiassen

    Full Text Available Neurophysiologic theory and some empirical evidence suggest that fatigue caused by physical work may be more effectively recovered during "diverting" periods of cognitive activity than during passive rest; a phenomenon of great interest in working life. We investigated the extent to which development and recovery of fatigue during repeated bouts of an occupationally relevant reaching task was influenced by the difficulty of a cognitive activity between these bouts. Eighteen male volunteers performed three experimental sessions, consisting of six 7-min bouts of reaching alternating with 3 minutes of a memory test differing in difficulty between sessions. Throughout each session, recordings were made of upper trapezius muscle activity using electromyography (EMG, heart rate and heart rate variability (HRV using electrocardiography, arterial blood pressure, and perceived fatigue (Borg CR10 scale and SOFI. A test battery before, immediately after and 1 hour after the work period included measurements of maximal shoulder elevation strength (MVC, pressure pain threshold (PPT over the trapezius muscles, and a submaximal isometric contraction. As expected, perceived fatigue and EMG amplitude increased during the physical work bouts. Recovery did occur between the bouts, but fatigue accumulated throughout the work period. Neither EMG changes nor recovery of perceived fatigue during breaks were influenced by cognitive task difficulty, while heart rate and HRV recovered the most during breaks with the most difficult task. Recovery of perceived fatigue after the 1 hour work period was also most pronounced for the most difficult cognitive condition, while MVC and PPT showed ambiguous patterns, and EMG recovered similarly after all three cognitive protocols. Thus, we could confirm that cognitive tasks between bouts of fatiguing physical work can, indeed, accelerate recovery of some factors associated with fatigue, even if benefits may be moderate and some

  17. Can cognitive activities during breaks in repetitive manual work accelerate recovery from fatigue? A controlled experiment.

    Science.gov (United States)

    Mathiassen, Svend Erik; Hallman, David M; Lyskov, Eugene; Hygge, Staffan

    2014-01-01

    Neurophysiologic theory and some empirical evidence suggest that fatigue caused by physical work may be more effectively recovered during "diverting" periods of cognitive activity than during passive rest; a phenomenon of great interest in working life. We investigated the extent to which development and recovery of fatigue during repeated bouts of an occupationally relevant reaching task was influenced by the difficulty of a cognitive activity between these bouts. Eighteen male volunteers performed three experimental sessions, consisting of six 7-min bouts of reaching alternating with 3 minutes of a memory test differing in difficulty between sessions. Throughout each session, recordings were made of upper trapezius muscle activity using electromyography (EMG), heart rate and heart rate variability (HRV) using electrocardiography, arterial blood pressure, and perceived fatigue (Borg CR10 scale and SOFI). A test battery before, immediately after and 1 hour after the work period included measurements of maximal shoulder elevation strength (MVC), pressure pain threshold (PPT) over the trapezius muscles, and a submaximal isometric contraction. As expected, perceived fatigue and EMG amplitude increased during the physical work bouts. Recovery did occur between the bouts, but fatigue accumulated throughout the work period. Neither EMG changes nor recovery of perceived fatigue during breaks were influenced by cognitive task difficulty, while heart rate and HRV recovered the most during breaks with the most difficult task. Recovery of perceived fatigue after the 1 hour work period was also most pronounced for the most difficult cognitive condition, while MVC and PPT showed ambiguous patterns, and EMG recovered similarly after all three cognitive protocols. Thus, we could confirm that cognitive tasks between bouts of fatiguing physical work can, indeed, accelerate recovery of some factors associated with fatigue, even if benefits may be moderate and some responses may be

  18. Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-04

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Nakamura, Hideo; Saeki, Hiroyuki

    1991-03-01

    Experimental data for the 10% hot leg break test, Run SB-HL-04, conducted on March 29, 1988 at the ROSA-IV Large Scale Test Facility (LSTF), are presented. This test was conducted as part of test series which studied the effect of break orientation on 10% hot leg break transient, and represented a vertical upward break. Other two tests in this test series represented horizontal break and vertical downward break, respectively. The results of these tests were characterized by asymmetric loop responses, flashing in the cold legs as well as upper downcomer, and condensation depressurization in the cold legs following injection of emergency core coolant (ECC) from accumulators. (author)

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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)

  4. 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.

  5. 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.

  6. 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)

  7. Modelling small scale infiltration experiments into bore cores of crystalline rock and break-through curves

    International Nuclear Information System (INIS)

    Hadermann, J.; Jakob, A.

    1987-04-01

    Uranium infiltration experiments for small samples of crystalline rock have been used to model radionuclide transport. The theory, taking into account advection and dispersion in water conducting zones, matrix diffusion out of these, and sorption, contains four independent parameters. It turns out, that the physical variables extracted from those of the best-fit parameters are consistent with values from literature and independent measurements. Moreover, the model results seem to differentiate between various geometries for the water conducting zones. Alpha-autoradiographies corroborate this result. A sensitivity analysis allows for a judgement on parameter dependences. Finally some proposals for further experiments are made. (author)

  8. 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)

  9. 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

  10. 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…

  11. Fast reactor sodium systems operation experience and 'leak-before-break' criterion

    International Nuclear Information System (INIS)

    Ivanenko, V.N.; Zybin, V.A.

    1996-01-01

    In the paper sodium leakage detection systems used at fast reactors are described. Requirements on their main characteristics (sensitivity, response lime) are formulated. Results of tests are presented on studying the parameters of sodium leak detection systems including experiments on the measurement of size distribution of aerosol particles that have passed through sodium systems thermal insulation after leak initiation. Comparison of these data with dispersion of particles formed at free burning is carried out. Experience of real leaks that occurred at fast reactor sodium systems is analyzed. It has been shown that initiation and development of real leaks do not always follow the theoretical scheme. A substantial role of human factor for sodium systems reliability relative to sodium leaks is stressed. (author)

  12. The Mice Drawer System (MDS) experiment and the space endurance record-breaking mice.

    Science.gov (United States)

    Cancedda, Ranieri; Liu, Yi; Ruggiu, Alessandra; Tavella, Sara; Biticchi, Roberta; Santucci, Daniela; Schwartz, Silvia; Ciparelli, Paolo; Falcetti, Giancarlo; Tenconi, Chiara; Cotronei, Vittorio; Pignataro, Salvatore

    2012-01-01

    The Italian Space Agency, in line with its scientific strategies and the National Utilization Plan for the International Space Station (ISS), contracted Thales Alenia Space Italia to design and build a spaceflight payload for rodent research on ISS: the Mice Drawer System (MDS). The payload, to be integrated inside the Space Shuttle middeck during transportation and inside the Express Rack in the ISS during experiment execution, was designed to function autonomously for more than 3 months and to involve crew only for maintenance activities. In its first mission, three wild type (Wt) and three transgenic male mice over-expressing pleiotrophin under the control of a bone-specific promoter (PTN-Tg) were housed in the MDS. At the time of launch, animals were 2-months old. MDS reached the ISS on board of Shuttle Discovery Flight 17A/STS-128 on August 28(th), 2009. MDS returned to Earth on November 27(th), 2009 with Shuttle Atlantis Flight ULF3/STS-129 after 91 days, performing the longest permanence of mice in space. Unfortunately, during the MDS mission, one PTN-Tg and two Wt mice died due to health status or payload-related reasons. The remaining mice showed a normal behavior throughout the experiment and appeared in excellent health conditions at landing. During the experiment, the mice health conditions and their water and food consumption were daily checked. Upon landing mice were sacrificed, blood parameters measured and tissues dissected for subsequent analysis. To obtain as much information as possible on microgravity-induced tissue modifications, we organized a Tissue Sharing Program: 20 research groups from 6 countries participated. In order to distinguish between possible effects of the MDS housing conditions and effects due to the near-zero gravity environment, a ground replica of the flight experiment was performed at the University of Genova. Control tissues were collected also from mice maintained on Earth in standard vivarium cages.

  13. The Mice Drawer System (MDS experiment and the space endurance record-breaking mice.

    Directory of Open Access Journals (Sweden)

    Ranieri Cancedda

    Full Text Available The Italian Space Agency, in line with its scientific strategies and the National Utilization Plan for the International Space Station (ISS, contracted Thales Alenia Space Italia to design and build a spaceflight payload for rodent research on ISS: the Mice Drawer System (MDS. The payload, to be integrated inside the Space Shuttle middeck during transportation and inside the Express Rack in the ISS during experiment execution, was designed to function autonomously for more than 3 months and to involve crew only for maintenance activities. In its first mission, three wild type (Wt and three transgenic male mice over-expressing pleiotrophin under the control of a bone-specific promoter (PTN-Tg were housed in the MDS. At the time of launch, animals were 2-months old. MDS reached the ISS on board of Shuttle Discovery Flight 17A/STS-128 on August 28(th, 2009. MDS returned to Earth on November 27(th, 2009 with Shuttle Atlantis Flight ULF3/STS-129 after 91 days, performing the longest permanence of mice in space. Unfortunately, during the MDS mission, one PTN-Tg and two Wt mice died due to health status or payload-related reasons. The remaining mice showed a normal behavior throughout the experiment and appeared in excellent health conditions at landing. During the experiment, the mice health conditions and their water and food consumption were daily checked. Upon landing mice were sacrificed, blood parameters measured and tissues dissected for subsequent analysis. To obtain as much information as possible on microgravity-induced tissue modifications, we organized a Tissue Sharing Program: 20 research groups from 6 countries participated. In order to distinguish between possible effects of the MDS housing conditions and effects due to the near-zero gravity environment, a ground replica of the flight experiment was performed at the University of Genova. Control tissues were collected also from mice maintained on Earth in standard vivarium cages.

  14. Breaking Good News’: Neurologists' experiences of discussing SUDEP with patients in Scotland

    OpenAIRE

    Nisbet, Tom; Turnbull, Susan; Mulhern, Sharon; Razvi, Saif

    2017-01-01

    Since the findings of a Fatal Accident Inquiry (FAI) in 2010, clinicians working in Scotland have been advised to discuss the risk of Sudden Unexpected Death in Epilepsy (SUDEP) with patients immediately or soon after a diagnosis of epilepsy is made. A thematic analysis was used to describe the experiences discussing SUDEP of 10 clinicians (six Consultant Neurologists and four Neurology Registrars) working in Scotland. Contrary to previous research, clinicians appear to be routinely discussin...

  15. Study of the breaking of the CP symmetry in the BABAR experiment

    International Nuclear Information System (INIS)

    Ganjour, S.

    2007-09-01

    This report summarizes my scientific activities from 1995 to 2007. During this period of time, my research work was related to the particle physics experiment BABAR. The BABAR experiment has been running since 1999 at the PEP-II e + e - asymmetric B-factory located at SLAC. This experiment searches for CP violation in the system of B mesons and tests the Standard Model through the measurements of the angles and the sides of the Unitarity Triangle. My research work is divided in five main topics: study of the BABAR magnet system and measurement of the magnetic field in the central tracking volume; project of the particle identification system based on aerogel counters for the forward region of the detector; conception of the magnetic shield and measurements of the fringe field in the region of photomultipliers of the DIRC (Detector of Internally Reflected Cherenkov light) system, the principal particle identification system of BABAR; development of the partial reconstruction technique of B mesons and study of the B 0 → D s * + D *- decays; measurement of CP violation in the B 0 → D *± π ± decays and constraint on the Unitary Triangle parameter sin(2β + γ) using these decays. (author)

  16. 'Don't blame the middle man': an exploratory qualitative study to explore the experiences of translators breaking bad news.

    Science.gov (United States)

    Prentice, Joanna; Nelson, Annmarie; Baillie, Jessica; Osborn, Hannah; Noble, Simon

    2014-07-01

    Healthcare professionals find breaking bad news difficult and upsetting. Increasing cultural diversity has led to a greater number of patients whose first language differs to that of the healthcare provider, with more patients requiring a translator to facilitate communication. Hospitals often ask non-clinical translators to facilitate breaking bad news. We sought to explore the experiences of translators within a specialist oncology centre. Following ethical and governance approvals, semi-structured interviews were undertaken with five translators recruited from the specialist oncology centre. Interviews were audiotaped and transcribed verbatim. The data were analysed thematically, with major themes and subthemes identified. Outpatient setting of a regional cancer centre. Translators serving a regional cancer centre. Qualitative data identified through thematic analysis. Major themes included the significant emotional impact of translating distressing information, the challenges of accurately conveying information in a culturally congruent format and the need for formal briefing, debriefing and support. Subthemes included feeling guilty for divulging distressing news, being the focus of patients' distress or anger, and feeling in conflict with the patient or family and issues surrounding confidentiality. Translators also felt a strong sense of advocacy for the patients and found encounters with death and dying emotionally challenging. The increasing use of translators in the care of patients with advanced cancer is increasingly resulting in lay people being subject to similar emotional pressures faced by clinical staff, yet without the necessary formal training or support mechanisms that are recommended for clinicians. This exploratory study highlights the training and support needs of non-clinical staff as identifying a unique set of communication challenges faced by translators. © The Royal Society of Medicine.

  17. Breaking bad news--parents' experience of learning that their child has leukaemia.

    LENUS (Irish Health Repository)

    Oshea, J

    2012-02-03

    This study aimed to seek parents\\' experiences of how they learned their child had leukaemia and therefore identify ways of improving this process. To achieve this task a questionnaire was designed to ask parents about specific elements of the initial interview and give them opportunity to add their thoughts and feelings on the subject. All children with a diagnosis of leukaemia over an eighteen-year period were identified and parents of those children still alive were invited to partake in the study. 49 out of 50 families agreed to participate of which 35 (72%) returned completed questionnaires. The majority 29 (83%) expressed overall satisfaction. Their replies confirmed some findings of previous studies, and also offered some new insights. Examples of new findings or expansion on previous findings include observations on the presence of young children at the initial interview; the importance of the language used in conveying the diagnosis and prognostic information, and a preference for actuarial terms when discussing prognosis. Telling parents their child has leukaemia is a challenging and important task. The experience of parents gives us valuable insights into our own communication skills and highlights areas of possible improvement in this difficult area.

  18. 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.

  19. RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bayless, P.D.

    2003-01-17

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

  20. RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    International Nuclear Information System (INIS)

    Bayless, P.D.

    2003-01-01

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior

  1. RELAP-5/MOD 3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    Energy Technology Data Exchange (ETDEWEB)

    Paul D. Bayless

    2003-01-01

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

  2. LOCA verification and data bank. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Varacalle, Jr., D. J.; Cox, N. D.; Atwood, C. L.; Madden, S. C.; Condie, K. G.

    1979-01-01

    The purpose of this task was to derive local conditions heat transfer parameters and their uncertainties using computer codes and experimentally derived boundary conditions. To accomplish this objective, Semiscale S-02-9 blowdown experiment was used along with the INVERT (an inverse heat conduction code) and RELAP4 (a thermal hydraulic code) codes as the analytical tools. The uncertainties calculated for the local conditions were limited to those introduced by inaccuracies in the experimentally measured boundary conditions. The propagation of the measurement uncertainties through the codes was investigated by varying the code input using statistical methods and a response surface technique.

  3. Breaking Bad News: A Survey of Radiology Residents' Experiences Communicating Results to Patients.

    Science.gov (United States)

    Narayan, Anand; Dromi, Sergio; Meeks, Adam; Gomez, Erin; Lee, Bonmyong

    The practice of radiology often includes routine communication of diagnostic test results directly to patients in breast imaging and interventional radiology. There is increasing interest in expanding direct communication throughout radiology. Though these conversations can substantially affect patient well-being, there is limited evidence indicating that radiology residents are specifically taught methods to effectively convey imaging results to patients. Our purpose is to evaluate resident experience communicating imaging results to patients. An IRB-approved study with a total of 11 pilot-tested questions was used. Surveyed programs included radiology residents (PGY2-PGY5) at 2 urban residency programs. Online surveys were administered using SurveyMonkey and e-mailed to residents at both programs (starting November 20, 2015, completed March 31, 2016). Demographics were obtained with survey proportions compared using logistic regression (P < 0.05, statistically significant). A total of 73 residents responded (93.6% response rate) with similar response rates at each institution (P = 0.689). Most were male (71.2%) with 17.8% planning to go into breast imaging (21.9%, interventional radiology (IR)). Furthermore, 83.6% described no training in communicating radiology results to patients; 91.8% of residents communicated results with patients (87.7% diagnostic imaging tests and 57.5% biopsies). Residents most commonly communicated results in person (75.3%) followed by phone (64.4%), and 79.4% agreed or strongly agreed that additional training relaying results would be helpful. A large majority of radiology residents have communicated test results to patients, yet few have received training in how to communicate these results. A large majority of residents expressed interest in obtaining additional communication training. Additional research is required to determine ideal methods to educate residents on communicating test results. Copyright © 2018 Elsevier Inc. All

  4. 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)

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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

  11. RELAP5/Mod3.3 and MARS3.0a Modeling of a Siphon Break Experiment

    International Nuclear Information System (INIS)

    Park, Su Ki; Kim, Heon Il; Park, Cheol; Yoon, Ju Hyeon

    2011-01-01

    Pool water plays a very important role as a final heat sink for most pool-type research reactors following postulated events. Therefore, one of design criteria for the reactors is that the water level of reactor pool must not decrease below a predefined elevation even against the most severe accident due to ruptures of coolant boundary of connecting systems to the reactor pool. In order to accomplish the design criterion, all the connecting systems are usually arranged to be above the elevation of reactor core. However, some research reactors with a downward flow in the reactor core have a primary cooling system located below the elevation of reactor core because of meeting an available net positive suction head of pumps in the system. These reactors have a provision consisting of pipes penetrating a reactor pool wall at a higher elevation than that of reactor core and siphon break devices to meet the design criterion. A series of experiments was carried out to figure out thermal hydraulic characteristics during siphon is blocked and establish design requirements for siphon breaker. The experimental study provided a lot of data and observations to the process of siphon break, but it does not provide a sufficient theoretical analysis and present practical design requirements applicable to industry. The experimental range is not also sufficient to cover operating conditions of siphon breakers for research reactors. A series of numerical simulations on the experimental data has been tried by using thermal hydraulic system analysis codes, RELAP5/Mod3.3 and MARS3.0a. This paper includes a part of the numerical simulations. First output from this study shows an importance of an adequate use of thermal hydraulic models in the codes and a big different prediction between the two codes especially in relation to the use of choked flow option. From this study, it seems that RELAP5/Mod3.3 has some problems on the control of a choked flow option-flag or the prediction of a

  12. 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

  13. Status of the INL gas reactor test system experiment facility

    International Nuclear Information System (INIS)

    Marshall, Theron; Bennet, Brion; Tschaggeny, Charles; Reyes, Jose; Groome, John

    2007-01-01

    The Gas Reactor Test System (GRTS) is an experiment facility for examining the thermal hydraulic performance of the Generation IV, Very High Temperature Reactor (VHTR) during a Large-Break Loss of Coolant Accident (LB-LOCA). The LB-LOCA is defined as the double guillotine break of the VHTR coaxial inlet and outlet cross duct. Two system safety codes, MELCOR and RELAP5-3D were used to calculate core temperatures and flow rates during the LB-LOCA transient. Computational fluid dynamics modeling of the transient produced flow vectors and gas species distribution. The most important phenomenon during the transient is the lock-exchange process, which suppresses the onset of natural circulation until considerable molecular diffusion has occurred. The GRTS was designed based upon a hierarchical two tier scaling analysis whose primary objective was replicating the lock-exchange and natural circulation characteristics of the VHTR. The GRTS uses a scaled graphite core to represent the VHTR's graphite core. An in-depth scaling analysis was performed for the GRTS in order to ensure that it accurately simulated the VHTR thermal responses. RELAP5-3D thermal analyses, ProEngineer stress analyses, and combined FLUENT-STARCD CFD analyses have provided a system design that fulfills the GRTS mission statement. This paper discusses the design analyses and their implications on the GRTS capabilities. A discussion is also presented on the preliminary instrumentation plan. The GRTS will provide an extensive temperature map of the VHTR core outlet plenum and its core support, oxygen transport rates during the lock-exchange phenomenon, and thermal conduction rates from the core to the vessel. As a result of the GRTS using helium coolant at 950 C, the resulting experiment data is expected to considerably extend the U.S. database for high-temperature gas reactor operations. Finally, the discussion will present conclusions from the GRTS manufacturing and quality control processes that may

  14. Writing experience: does ethnography convey a crisis of representation, or an ontological break with the everyday world?

    Science.gov (United States)

    Ho, Wing-Chung

    2008-11-01

    This paper is premised on the "ontological break" as coined by Alfred Schutz that disconnects two realms: the "world of consociates" where social reality is directly experienced face-to-face in the vivid present, and the "world of contemporaries" where the other is interpreted in terms of "types." It is argued that this break is a suggestive vehicle for conducting a meta-exposition of major claims which problematize the traditional authority of ethnography. In the light of the break, the postmodernist attempts to attain or retain the here-and-now understanding of subjective meaning, or "voice" in ethnographies are but epistemological impossibilities. It is concluded that the postmodernist privileging of a "naive ethnography" which emphasizes "experiential," "interpretive," "dialogical," and "polyphonic" processes is neither able to deliver on its promise at the methodic level, nor amendable to Schutz's ontological break at the theoretical level.

  15. 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)

  16. 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)

  17. Breaking bad news in clinical setting - health professionals' experience and perceived competence in Southwestern Nigeria: a cross sectional study.

    Science.gov (United States)

    Adebayo, Philip Babatunde; Abayomi, Olukayode; Johnson, Peter O; Oloyede, Taofeeq; Oyelekan, Abimbola A A

    2013-01-01

    Communication skills are vital in clinical settings because the manner in which bad news is delivered could be a huge determinant of responses to such news; as well as compliance with beneficial treatment option. Information on training, institutional guidelines and protocols for breaking bad news (BBN) is scarce in Nigeria. We assessed the training, experience and perceived competence of BBN among medical personnel in southwestern Nigeria. The study was a cross-sectional descriptive study conducted out among doctors and nurses in two healthcare institutions in southwestern Nigeria using an anonymous questionnaire (adapted from the survey by Horwitz et al.), which focused on the respondents training, awareness of protocols in BBN; and perceived competence (using a Five-Point Likert Scale) in five clinical scenarios. We equally asked the respondents about an instance of BBN they have recently witnessed. A total of 113 of 130 selected (response rate 86.9%) respondents were studied. Eight (7.1%) of the respondents knew of the guidelines on BBN in the hospital in which they work. Twenty-three (20.3%) respondents claimed knowledge of a protocol. The median perceived competence rating was 4 out of 5 in all the clinical scenarios. Twenty-five (22.1%) respondents have had a formal training in BBN and they generally had significant higher perceived competence rating (P = 0.003-0.021). There is poor support from fellow workers during instances of BBN. It appears that the large proportion of the respondents in this study were unconsciously incompetent in BBN in view of the low level of training and little or no knowledge of well known protocols for BBN even though self-rated competence is high. Continuous medical education in communication skills among health personnel in Nigeria is advocated.

  18. 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

  19. 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...

  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. 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. Preliminary analysis of ROSAIII experiment, (2)

    International Nuclear Information System (INIS)

    Kitaguchi, Hidemi; Suzuki, Mitsuhiro; Sobajima, Makoto; Adachi, Hiromichi; Shiba, Masayoshi.

    1978-02-01

    Loss-of-coolant accident (LOCA) experiments to be performed in ROSAIII has been examined with computer code RELAP-4J concerning the experimental conditions. From the results (1) to (3), the needs (4) to (6) are there. (1) Initial enthalpy distribution is important for simulation of break flow of an actual BWR. (2) The simulations of lower plenumn flashing and pressure transient in pressure vessel are good except when power is lacking. (3) The simulation of the cladding temperature transient is difficult because of lack of physical properties. (4) The initial pressure distribution in the facility for different core flow rates up to 72 lb/sec must be attained to analyze accurately. (5) Reverse core flow detectors and reverse jet pump flow detectors are necessary to compare flow pattern of recirculation loops between calculation and experiment. (6) Further information is necessary on physical properties of the fuels. (auth.)

  4. 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)

  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. More Than Just a Break from Treatment: How Substance Use Disorder Patients Experience the Stable Environment in Horse-Assisted Therapy.

    Science.gov (United States)

    Kern-Godal, Ann; Brenna, Ida Halvorsen; Arnevik, Espen Ajo; Ravndal, Edle

    2016-01-01

    Inclusion of horse-assisted therapy (HAT) in substance use disorder (SUD) treatment is rarely reported. Our previous studies show improved treatment retention and the importance of the patient-horse relationship. This qualitative study used thematic analysis, within a social constructionist framework, to explore how eight patients experienced contextual aspects of HAT's contribution to their SUD treatment. Participants described HAT as a "break from usual treatment". However, four interrelated aspects of this experience, namely "change of focus", "activity", "identity", and "motivation," suggest HAT is more than just a break from usual SUD treatment. The stable environment is portrayed as a context where participants could construct a positive self: one which is useful, responsible, and accepted; more fundamentally, a different self from the "patient/self" receiving treatment for a problem. The implications extend well beyond animal-assisted or other adjunct therapies. Their relevance to broader SUD policy and treatment practices warrants further study.

  7. 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.

  8. Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-02

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Hirata, Kazuo; Gotou, Hiroki

    1990-03-01

    Experimental data for the 10% hot leg break test, Run SB-HL-02, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on June 30, 1987, are presented. This test assumed total failure of both high pressure injection (HPI) and auxiliary feedwater (AFW) systems. The test results were characterized by asymmetric loop responses, flashing in the cold legs and upper downcomer, as well as condensation depressurization in the cold legs following injection of emergency core coolant (ECC) from accumulators. (author)

  9. Study of proton-deuteron break-up reaction in exclusive experiment at 1 GeV

    International Nuclear Information System (INIS)

    Aleshin, N.P.; Belostotskij, S.L.; Dotsenko, Yu.V.

    1987-07-01

    The exclusive proton-deuteron break-up reaction pD yields ppn was studied at 1 GeV. Differential cross sections and polarizations of the final protons were measured in the range of neutron-spectator momenta 0 3 3 <0.2 GeV/c, respectively. The data obtained are well described within the framework of impulse approximation with the Paris wave function of the deuteron. (author)

  10. 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...

  11. 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 ...

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

    International Nuclear Information System (INIS)

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

    1985-11-01

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

  13. 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 ...

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  7. 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)

  8. 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.

  9. Rotational symmetry breaking in the topological superconductor SrxBi2Se3 probed by upper-critical field experiments.

    Science.gov (United States)

    Pan, Y; Nikitin, A M; Araizi, G K; Huang, Y K; Matsushita, Y; Naka, T; de Visser, A

    2016-06-28

    Recently it was demonstrated that Sr intercalation provides a new route to induce superconductivity in the topological insulator Bi2Se3. Topological superconductors are predicted to be unconventional with an odd-parity pairing symmetry. An adequate probe to test for unconventional superconductivity is the upper critical field, Bc2. For a standard BCS layered superconductor Bc2 shows an anisotropy when the magnetic field is applied parallel and perpendicular to the layers, but is isotropic when the field is rotated in the plane of the layers. Here we report measurements of the upper critical field of superconducting SrxBi2Se3 crystals (Tc = 3.0 K). Surprisingly, field-angle dependent magnetotransport measurements reveal a large anisotropy of Bc2 when the magnet field is rotated in the basal plane. The large two-fold anisotropy, while six-fold is anticipated, cannot be explained with the Ginzburg-Landau anisotropic effective mass model or flux flow induced by the Lorentz force. The rotational symmetry breaking of Bc2 indicates unconventional superconductivity with odd-parity spin-triplet Cooper pairs (Δ4-pairing) recently proposed for rhombohedral topological superconductors, or might have a structural nature, such as self-organized stripe ordering of Sr atoms.

  10. 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

  11. 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)

  12. 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).

  13. 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

  14. More than Just a Break from Treatment: How Substance Use Disorder Patients Experience the Stable Environment in Horse-Assisted Therapy

    Directory of Open Access Journals (Sweden)

    Ann Kern-Godal

    2016-01-01

    Full Text Available Inclusion of horse-assisted therapy (HAT in substance use disorder (SUD treatment is rarely reported. Our previous studies show improved treatment retention and the importance of the patient–horse relationship. This qualitative study used thematic analysis, within a social constructionist framework, to explore how eight patients experienced contextual aspects of HAT's contribution to their SUD treatment. Participants described HAT as a “break from usual treatment”. However, four interrelated aspects of this experience, namely “change of focus”, “activity”, “identity”, and “motivation,” suggest HAT is more than just a break from usual SUD treatment. The stable environment is portrayed as a context where participants could construct a positive self: one which is useful, responsible, and accepted; more fundamentally, a different self from the “patient/self” receiving treatment for a problem. The implications extend well beyond animal-assisted or other adjunct therapies. Their relevance to broader SUD policy and treatment practices warrants further study.

  15. 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.

  16. Dynamic breaking of a single gold bond

    DEFF Research Database (Denmark)

    Pobelov, Ilya V.; Lauritzen, Kasper Primdal; Yoshida, Koji

    2017-01-01

    . Conversely, if the force is loaded rapidly it is more likely that the maximum breaking force is measured. Paradoxically, no clear differences in breaking force were observed in experiments on gold nanowires, despite being conducted under very different conditions. Here we explore the breaking behaviour...

  17. 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

  18. 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.

  19. 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.

  20. 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)

  1. Examining the Experiences of a Short Break Scheme amongst Adolescents with Disabilities (Service Users) and Their Parents

    Science.gov (United States)

    Spruin, Elizabeth; Abbott, Nicola; Holt, Nicole

    2018-01-01

    Globally, families who care for a child or adolescent with disabilities have been found to experience high levels of maternal ill health, stress, depression and family breakdown. In extreme cases, children and adolescents may have to move away from their family to a permanent residential placement. A potentially more appropriate and cost-effective…

  2. "Give Me a Break--English Is Not My First Language!": Experiences of Linguistically Diverse Student Teachers

    Science.gov (United States)

    Eros, John

    2016-01-01

    Today's K-12 music educators interact regularly with students from culturally diverse communities and backgrounds. Although research exists on culturally diverse students, there is comparatively little research on music teachers who do, themselves, represent diverse cultures. The purpose of this study was to investigate the experiences of three…

  3. 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

  4. Validation of leak before break cases applied to reactor pressure vessels by large scale experiments and the R6 fracture assessment method

    International Nuclear Information System (INIS)

    Sharples, J.K.; Sherry, A.H.; Stewart, G.

    1996-01-01

    Leak-before-break (LBB) arguments are often relied upon when making plant life extension safety cases for primary pressure vessels in some UK Magnox stations. The calculation of critical crack length of through-wall defects is a central feature to LBB arguments. Within the UK, such calculations are performed by the R6 fracture assessment method. In order to underwrite the integrity of specific Magnox reactor vessels in terms of LBB, a series of large scale fracture experiments were carried out. These were undertaken on wide plate specimens containing a through-wall crack situated in appropriate weld material. Applying R6 methodology to the experiments in the same way that it was applied to the plant vessels enabled margins on the critical crack length calculations to be determined. Data obtained from the experiments, together with calculations based on the higher order (Option 3) R6 failure assessment diagram, also enabled structural J-resistance fracture curves to be obtained. These curves were compared with those obtained from conventional small-scale specimens of the same weld material

  5. Experience in KINS on Best Estimate Calculation with Uncertainty Evaluation

    International Nuclear Information System (INIS)

    Bang, Young Seok; Huh, Byung-Gil; Cheong, Ae-ju; Woo, Sweng-Woong

    2013-01-01

    In the present paper, experience of Korea Institute of Nuclear Safety (KINS) on Best Estimate (BE) calculation and uncertainty evaluation of large break loss-of-coolant accident (LB LOCA) of Korean Pressurized Water Reactor (PWR) with various type of Emergency Core Cooling System (ECCS) design is addressed. Specifically, the current status of BE code, BE calculations and uncertainty parameters and related approaches are discussed. And the specific problem such as how to recover the difficulty in treating the uncertainty related to the phenomena specific to ECCS design (e.g., upper plenum injection phenomena) is discussed. Based on the results and discussion, it is found that the present KINS-REM has been successfully developed and applied to the regulatory auditing calculations. Need of further study includes the improvement of reflood model of MARS code, uncertainty of blowdown quenching, and reconsideration of the unconcerned model and fuel conductivity degradation with burnup. (authors)

  6. 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

  7. Breaking Boundaries

    DEFF Research Database (Denmark)

    . As a fundamental human experience, liminality transmits cultural practices, codes, rituals, and meanings in-between aggregate structures and uncertain outcomes. As a methodological tool it is well placed to overcome disciplinary boundaries, which often direct attention to specific structures or sectors of society....... Its capacity to provide explanatory accounts of seemingly unstructured situations provides an opportunity to link experience-based and culture-oriented approaches not only to contemporary problems but also to undertake comparisons across historical periods. From a perspective of liminality...

  8. Review of the Safety Concern Related to CANDU Moderator Temperature Distribution and Status of KAERI Moderator Circulation Test (MCT) Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Bo W.; Kim, Hyoung T. [Severe Accident and PHWR Safety Research Division, Daejeon (Korea, Republic of); Kim, Tongbeum [University of the Witwatersrand, Johannesburg (South Africa); Im, Sunghyuk [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep threshold temperature and no further deformation is expected. Consequently, a sufficient condition to ensure fuel channel integrity following a large LOCA, is the avoidance of sustained calandria tubes dryout. If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The subcooling requirements are determined from a set of experiments known as fuel channel contact experiments. The difference between available subcooling and required subcooling is called subcooling margins. The moderator flow circulation patterns are complicated slow flows that significantly vary from buoyancy dominated to inertia dominated patterns. Accurate predictions of flow patterns are essential for accurate calculation of moderator temperature distributions and the related moderator subcooling. Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep

  9. Microdosimetrical calculations of the rate of repairable DNA - double strand breaks based on a model for the interpretation of experiments with different doses and radiation qualities

    International Nuclear Information System (INIS)

    Rosemann, M.; Regel, K.

    1990-01-01

    When comparing various DNA injuries induced by radiation double breaks were shown to play peculiar role in subsequent cell changes such as inactivation, aberrations, mutations and transformations. However it was proved that significant part of radiation-induced double breaks could be repaied within cell. 3 refs

  10. 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

  11. 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.)

  12. 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.

  13. 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)

  14. 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)

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

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. Study of the breaking of the CP symmetry in the BABAR experiment; Etude de la violation de la symetrie CP dans l'experience BABAR

    Energy Technology Data Exchange (ETDEWEB)

    Ganjour, S

    2007-09-15

    This report summarizes my scientific activities from 1995 to 2007. During this period of time, my research work was related to the particle physics experiment BABAR. The BABAR experiment has been running since 1999 at the PEP-II e{sup +}e{sup -} asymmetric B-factory located at SLAC. This experiment searches for CP violation in the system of B mesons and tests the Standard Model through the measurements of the angles and the sides of the Unitarity Triangle. My research work is divided in five main topics: study of the BABAR magnet system and measurement of the magnetic field in the central tracking volume; project of the particle identification system based on aerogel counters for the forward region of the detector; conception of the magnetic shield and measurements of the fringe field in the region of photomultipliers of the DIRC (Detector of Internally Reflected Cherenkov light) system, the principal particle identification system of BABAR; development of the partial reconstruction technique of B mesons and study of the B{sup 0} {yields} D{sub s}{sup *} + D{sup *-} decays; measurement of CP violation in the B{sup 0} {yields} D{sup *{+-}}{pi}{sup {+-}} decays and constraint on the Unitary Triangle parameter sin(2{beta} + {gamma}) using these decays. (author)

  2. 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).

  3. Aluminum break-point contacts

    NARCIS (Netherlands)

    Heinemann, Martina; Groot, R.A. de

    1997-01-01

    Ab initio molecular dynamics is used to study the contribution of a single Al atom to an aluminum breakpoint contact during the final stages of breaking and the initial stages of the formation of such a contact. A hysteresis effect is found in excellent agreement with experiment and the form of the

  4. Sediment transport under breaking waves

    DEFF Research Database (Denmark)

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

    2000-01-01

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

  5. LOCA simulation in the NRU reactor: materials test-1

    International Nuclear Information System (INIS)

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607 0 F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions

  6. 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.

  7. CFD and FEM modeling of PPOOLEX experiments

    Energy Technology Data Exchange (ETDEWEB)

    Paettikangas, T.; Niemi, J.; Timperi, A. (VTT Technical Research Centre of Finland (Finland))

    2011-01-15

    Large-break LOCA experiment performed with the PPOOLEX experimental facility is analysed with CFD calculations. Simulation of the first 100 seconds of the experiment is performed by using the Euler-Euler two-phase model of FLUENT 6.3. In wall condensation, the condensing water forms a film layer on the wall surface, which is modelled by mass transfer from the gas phase to the liquid water phase in the near-wall grid cell. The direct-contact condensation in the wetwell is modelled with simple correlations. The wall condensation and direct-contact condensation models are implemented with user-defined functions in FLUENT. Fluid-Structure Interaction (FSI) calculations of the PPOOLEX experiments and of a realistic BWR containment are also presented. Two-way coupled FSI calculations of the experiments have been numerically unstable with explicit coupling. A linear perturbation method is therefore used for preventing the numerical instability. The method is first validated against numerical data and against the PPOOLEX experiments. Preliminary FSI calculations are then performed for a realistic BWR containment by modeling a sector of the containment and one blowdown pipe. For the BWR containment, one- and two-way coupled calculations as well as calculations with LPM are carried out. (Author)

  8. 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.

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. Experimental study of head loss and filtration for LOCA debris

    International Nuclear Information System (INIS)

    Rao, D.V.; Souto, F.J.

    1996-02-01

    A series of controlled experiments were conducted to obtain head loss and filtration characteristics of debris beds formed of NUKON trademark fibrous fragments, and obtain data to validate the semi-theoretical head loss model developed in NUREG/CR-6224. A thermally insulated closed-loop test set-up was used to conduct experiments using beds formed of fibers only and fibers intermixed with particulate debris. A total of three particulate mixes were used to simulate the particulate debris. The head loss data were obtained for theoretical fiber bed thicknesses of 0.125 inches to 4.0 inches; approach velocities of 0.15 to 1.5 ft/s; temperatures of 75 F and 125 F; and sludge-to-fiber nominal concentration ratios of 0 to 60. Concentration measurements obtained during the first flushing cycle were used to estimate the filtration efficiencies of the debris beds. For test conditions where the beds are fairly uniform, the head loss data were predictable within an acceptable accuracy range by the semi-theoretical model. The model was equally applicable for both pure fiber beds and the mixed beds. Typically the model over-predicted the head losses for very thin beds and for thin beds at high sludge-to-fiber mass ratios. This is attributable to the non-uniformity of such debris beds. In this range the correlation can be interpreted to provide upper bound estimates of head loss. This is pertinent for loss of coolant accidents in boiling water reactors

  14. Experimental study of head loss and filtration for LOCA debris

    Energy Technology Data Exchange (ETDEWEB)

    Rao, D.V.; Souto, F.J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

    1996-02-01

    A series of controlled experiments were conducted to obtain head loss and filtration characteristics of debris beds formed of NUKON{trademark} fibrous fragments, and obtain data to validate the semi-theoretical head loss model developed in NUREG/CR-6224. A thermally insulated closed-loop test set-up was used to conduct experiments using beds formed of fibers only and fibers intermixed with particulate debris. A total of three particulate mixes were used to simulate the particulate debris. The head loss data were obtained for theoretical fiber bed thicknesses of 0.125 inches to 4.0 inches; approach velocities of 0.15 to 1.5 ft/s; temperatures of 75 F and 125 F; and sludge-to-fiber nominal concentration ratios of 0 to 60. Concentration measurements obtained during the first flushing cycle were used to estimate the filtration efficiencies of the debris beds. For test conditions where the beds are fairly uniform, the head loss data were predictable within an acceptable accuracy range by the semi-theoretical model. The model was equally applicable for both pure fiber beds and the mixed beds. Typically the model over-predicted the head losses for very thin beds and for thin beds at high sludge-to-fiber mass ratios. This is attributable to the non-uniformity of such debris beds. In this range the correlation can be interpreted to provide upper bound estimates of head loss. This is pertinent for loss of coolant accidents in boiling water reactors.

  15. 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.

  16. 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.

  17. '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.

  18. 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.

  19. 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)

  20. 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.

  1. 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

  2. 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.

  3. Experiment search of the electroweak symmetry breaking in the H → γγ channel and of a solution of the hierarchy problem in the Atlas experiment: participation to the tests of the electronics of the electromagnetic calorimeter

    International Nuclear Information System (INIS)

    Escalier, M.

    2005-04-01

    This thesis deals with the understanding of the spontaneous electroweak symmetry breaking mechanism in the ATLAS experiment at LHC collider, by studying two complementary topics: the search for the Higgs boson in the H → γγ channel, and a search for extra dimensions in the gluon sector. Tests of the electronic of the electromagnetic calorimeter allowed us to validate various cards that were under the responsibility of the LPNHE. Using full simulation data of the detector allowed us to precisely compute mass resolution of the di-photon system. Due to recent theoretical improvements, signal and background have been studied at the next order of the perturbative development, which increases cross-sections. With regards to the jet background, a study has been done using discriminating variables in order to obtain, for a 80 % photons efficiency, a rejection factor of 7000. The discovery potential benefits from this change of cross-sections and increases by 50 % in comparison with the same analysis done at the leading order. In addition to this, a new analysis using a maximum likelihood method allowed us to increase by 40 % the discovery potential in comparison with our classical analysis. In conclusion, the Higgs boson of 120 GeV/c 2 can be now discovered in this channel with an integrated luminosity of 10 fb -1 . Furthermore, the consistency of the problem of the Higgs boson mass can be solved by introducing extra dimensions in which gluons can propagate. We have shown that it was possible to discover extra-dimensions up to a compactification scale of 15 TeV. (author)

  4. 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.)

  5. 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

  6. 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)

  7. 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

  8. In-pile experiments on fuel rod behaviour during a LOCA

    International Nuclear Information System (INIS)

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

    1981-07-01

    This report describes the results of the Test Series G2/3 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program ist the burnup, ranging from 2500 to 35000 MWd/t. The results of test series G2/3 (35000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  9. 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

  10. 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.

  11. 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)

  12. 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,...

  13. 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.

  14. 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

  15. 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.

  16. 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.

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

    International Nuclear Information System (INIS)

    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

  18. 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....

  19. 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,…

  20. 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.

  1. 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.

  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. 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.

  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. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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)

  10. 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

  11. 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...

  12. 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.

  13. 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.”...

  14. 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

  15. Experiment data of ROSA-III integral test RUN 913

    International Nuclear Information System (INIS)

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

    1989-09-01

    This report presents the experimental data of RUN 913, a 15 % split break test at the recirculation pump suction line. The ROSA-III test facility is a volumetrically scaled (1/424) model for the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV closure and the ECCS actuation are tripped by the liquid level in the upper downcomer as well as the BWR. The test was conducted successfully and important data base was obtained to assess the predictability of the LOCA analysis code. (author)

  16. Introduction to symmetry breaking and spin

    International Nuclear Information System (INIS)

    Ng, J.N.

    1992-05-01

    These lectures form an elementary introduction to the physics of symmetry breaking and the role polarization experiments play in the study of gauge symmetry breaking. Included here is an introduction to testing the electroweak sector of the standard model to one-loop and the use of oblique corrections as a probe of new physics. The second part of the lectures consists of an introduction to multiple Higgs models as sources of spontaneous CP violation. A brief discussion of using spin measurements in meson decays to study these sources of CP violation is also included. (author)

  17. The experimental investigation of supersymmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Peskin, M.E.

    1996-04-01

    If Nature is supersymmetric at the weak interaction scale, what can we hope to learn from experiments on supersymmetric particles? The most mysterious aspect of phenomenological supersymmetry is the mechanism of spontaneous supersymmetry breaking. This mechanism ties the observable pattern of supersymmetric particle masses to aspects of the underlying unified theory at very small distance scales. In this article, I will discuss a systematic experimental program to determine the mechanism of supersymmetry breaking. Both pp and e{sup +}e{sup -} colliders of the next generation play an essential role.

  18. The experimental investigation of supersymmetry breaking

    International Nuclear Information System (INIS)

    Peskin, M.E.

    1996-04-01

    If Nature is supersymmetric at the weak interaction scale, what can we hope to learn from experiments on supersymmetric particles? The most mysterious aspect of phenomenological supersymmetry is the mechanism of spontaneous supersymmetry breaking. This mechanism ties the observable pattern of supersymmetric particle masses to aspects of the underlying unified theory at very small distance scales. In this article, I will discuss a systematic experimental program to determine the mechanism of supersymmetry breaking. Both pp and e + e - colliders of the next generation play an essential role

  19. Breaking the cycle/mending the hoop: adverse childhood experiences among incarcerated American Indian/Alaska Native women in New Mexico.

    Science.gov (United States)

    De Ravello, Lori; Abeita, Jessica; Brown, Pam

    2008-03-01

    Incarcerated American Indian/Alaska Native (AI/AN) women have multiple physical, social, and emotional concerns, many of which may stem from adverse childhood experiences (ACE). We interviewed 36 AI/AN women incarcerated in the New Mexico prison system to determine the relationship between ACE and adult outcomes. ACE assessment included physical neglect, dysfunctional family (e.g., household members who abused substances, were mentally ill or suicidal, or who were incarcerated), violence witnessed in the home, physical abuse, and sexual abuse. The most prevalent ACE was dysfunctional family (75%), followed by witnessing violence (72%), sexual abuse (53%), physical abuse (42%), and physical neglect (22%). ACE scores were positively associated with arrests for violent offenses, lifetime suicide attempt(s), and intimate partner violence.

  20. 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

  1. Breaking waves on a dynamic Hele-Shaw beach

    NARCIS (Netherlands)

    Bokhove, Onno; van der Horn, Bram; van der Horn, Avraham/Bram; van der Meer, Roger M.; Zweers, Wout; Thornton, Anthony Richard

    We report the formation of quasi-steady beaches and dunes via breaking waves in our tabletop ‘Hele-Shaw’ beach experiment. Breaking waves are generated by a wave maker, and zeolite particles act as sand. The tank is narrow, just over one-particle diameter wide, creating a quasi-2D set-up. Classical

  2. Electroweak symmetry breaking and Higgs physics: basic concepts

    International Nuclear Information System (INIS)

    Gomez-Bock, M; Mondragon, M; Muehlleitner, M; Noriega-Papaqui, R; Pedraza, I; Spira, M; Zerwas, P M

    2005-01-01

    We present an introduction to the basic concepts of electroweak symmetry breaking and Higgs physics within the Standard Model and its sypersymmetric extensions. A brief overview will also be given on alternative mechanisms of symmetry breaking. In addition to the theoretical basis, the present experimental status of Higgs physics and implications for future experiments at the LHC and e + e - linear colliders are discussed

  3. 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

  4. 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

  5. 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)

  6. Break the "wall" and become creative: Enacting embodied metaphors in virtual reality.

    Science.gov (United States)

    Wang, Xinyue; Lu, Kelong; Runco, Mark A; Hao, Ning

    2018-03-19

    This study investigated whether the experience of "breaking the walls", the embodiment of the metaphor "breaking the rules", could enhance creative performance. The virtual reality technology was used to simulate the scenario where participants could "break the walls" while walking in a corridor. Participants were asked to solve the creativity-demanding problems (ie., alternative uses tasks, AUT) in either the "break" condition in which they had to break the walls to move forward in VR, or the "no-break" condition where no barrier walls would appear. Results showed higher AUT originality and AUT fluency in the "break" condition than in the "no-break" condition. Moreover, the effects of "breaking the walls" on AUT originality were fully mediated by cognitive flexibility and persistence. These findings may indicate that enacting metaphors such as "breaking the rules" contribute to creative performance. The enhanced cognitive flexibility and persistence may account for the benefits. Copyright © 2018 Elsevier Inc. All rights reserved.

  7. Bogotá experience in government: device and educational strategies to expressions, scenarios and breaks on a proposal for citizen culture

    Directory of Open Access Journals (Sweden)

    Gloria Clemencia VALENCIA GONZÁLEZ

    2012-05-01

    Full Text Available The Bogotá currently known, rebuilt from the ashes of 9 April got used for decades to live labors, fear and mistrust, but particularly in exclusion. With the arrival of Antanas Mockus to Mayor broke various paradigms: he and his cabinet not came from the traditional political caste, came from the Academy and claimed other relationships with citizenship. In addition, became the city learning space because their Government project drew on the political but evidenced in the pedagogical and found its deployment in the communicative. Six priorities proposed for government experience are articulated around one: civic culture. Culture that should be disseminated, learned and implementation from train in town, then so is generate sense of belonging, would facilitate urban coexistence and would lead to the respect of the common heritage and recognition of the rights and duties of citizens. In the Mayor Mockus’s «great classroom» citizenship would of apathy to the habitancia. Far pedagogy and communication, as not nominated devices, managed to go beyond the labelling and the indication and approached the civic culture to political socialization? This reflection article links to research «higher education and training citizen, case Bogotá» and consists of five sections.

  8. 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)

  9. 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)

  10. 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)

  11. 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

    In recent years the demands on 'fuel duties' have increased, including transient regimes, higher burnups and longer fuel cycles. To satisfy these demands, fuel vendors have developed and introduced new cladding and fuel material designs to provide sufficient margins for safe operation of the fuel components. National and international experimental programmes have been launched, and models have been developed or adapted to take into account the changed conditions. These developments enable water cooled reactors, which contribute about 95% of the nuclear power in the world today, to operate safely under all operating conditions; moreover, even under severe transient or accident conditions, such as reactivity initiated accidents (RIAs) or loss of coolant accidents (LOCAs), the behaviour of the fuel can be adequately predicted and the consequences of such events can be safely contained. In 2010 the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) recommended that a technical meeting on ''Fuel Behaviour and Modelling under Severe Transient and LOCA Conditions'' be held in Japan. The accident at the Fukushima Daiichi nuclear power plant in March 2011 highlighted the need to address this subject, and despite the difficult situation in Japan at the time, the recommended plan was confirmed, and the Japan Atomic Energy Agency (JAEA) hosted the technical meeting in Mito, Ibaraki Prefecture, Japan, from 18 to 21 October 2011. This meeting was the eighth in a series of IAEA meetings, which reflects Member States' continuing interest in the above issues. The previous meetings were held in 1980 (jointly with OECD Nuclear Energy Agency, Helsinki, Finland), 1983 (Riso, Denmark), 1986 (Vienna, Austria), 1988 (Preston, United Kingdom), 1992 (Pembroke, Canada), 1995 (Dimitrovgrad, Russian Federation) and 2001 (Halden, Norway). The purpose of the technical meeting was to provide a forum for international experts to review the current situation and the state of

  12. 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.

  13. 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.

  14. 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

  15. Swedish experience with RELAP5/MOD2 assessment

    International Nuclear Information System (INIS)

    Sandervag, O.

    1987-01-01

    The Swedish assessment of RELAP5/MOD2 is a part of the International Code Assessment program which is organized by the US NRC. The major part of the experimental data used for assessment is of Swedish origin. The data encompass critical flow and level swell data from the Marviken facility. A part of the agreed assessment matrix has been completed. Comparison with BWR integral test data shows that the major phenomena which control the core cooling during intermediate and large break LOCA are qualitatively reproduced by RELAP5. Assessment against separate and integral experiments shows that the dominant uncertainty in prediction of clad temperatures is due to a poor calculation of dryout. Predicted post dryout wall temperatures, given the experimental dryout location as input parameter, generally agree well with data. Simulations of level swell following depressurization of the large diameter Marviken vessel showed that RELAP5/MOD2 was able to calculate overall axial void profiles in fair agreement with data. The assessment indicated that increasing the modeling detail could give rise to numerical instabilities. Assessment against large scale critical flow data revealed that the agreement with data was somewhat dependent on upstream fluid conditions and modeling. Low quality two phase flow was, in general, accurately predicted while subcooled liquid flow and saturated steam flow were generally overpredicted if no discharge coefficient was applied

  16. 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.

  17. 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.

  18. 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)

  19. Breaking bad news in cancer patients.

    Science.gov (United States)

    Konstantis, Apostolos; Exiara, Triada

    2015-01-01

    In a regional hospital, many patients are newly diagnosed with cancer. Breaking the bad news in these patients and their relatives is a tough task. Many doctors are not experienced in talking to patients about death or death-related diseases. In recent years, there have been great efforts to change the current situation. The aim of this study was to investigate the experience and education of medical personnel in breaking bad news in a secondary hospital. 59 doctors from General Hospital of Komotini, Greece were included in the study. All the doctors were in clinical specialties that treated cancer patients. A brief questionnaire was developed based on current guidelines such as Baile/SPIKES framework and the ABCDE mnemonic. Residents are involved in delivering bad news less frequently than specialists. Only 21 doctors (35.59%) had specific training on breaking bad news. 20 doctors (33.90%) were aware of the available techniques and protocols on breaking bad news. 47 doctors (79.66%) had a consistent plan for breaking bad news. 57 (96.61%) delivered bad news in a quiet place, 53 (89.83%) ensured no interruptions and enough time, 53 (89.83%) used simple words and 54 (91.53%) checked for understanding and did not rush through the news. 46 doctors (77.97%) allowed relatives to determine patient's knowledge about the disease. There were low rates of specific training in breaking bad news. However, the selected location, the physician's speech and their plan were according to current guidelines.

  20. Breaking bad news in cancer patients

    Directory of Open Access Journals (Sweden)

    Apostolos Konstantis

    2015-01-01

    Full Text Available Objective: In a regional hospital, many patients are newly diagnosed with cancer. Breaking the bad news in these patients and their relatives is a tough task. Many doctors are not experienced in talking to patients about death or death-related diseases. In recent years, there have been great efforts to change the current situation. The aim of this study was to investigate the experience and education of medical personnel in breaking bad news in a secondary hospital. Materials and Methods: 59 doctors from General Hospital of Komotini, Greece were included in the study. All the doctors were in clinical specialties that treated cancer patients. A brief questionnaire was developed based on current guidelines such as Baile/SPIKES framework and the ABCDE mnemonic. Results: Residents are involved in delivering bad news less frequently than specialists. Only 21 doctors (35.59% had specific training on breaking bad news. 20 doctors (33.90% were aware of the available techniques and protocols on breaking bad news. 47 doctors (79.66% had a consistent plan for breaking bad news. 57 (96.61% delivered bad news in a quiet place, 53 (89.83% ensured no interruptions and enough time, 53 (89.83% used simple words and 54 (91.53% checked for understanding and did not rush through the news. 46 doctors (77.97% allowed relatives to determine patient′s knowledge about the disease. Conclusions: There were low rates of specific training in breaking bad news. However, the selected location, the physician′s speech and their plan were according to current guidelines.

  1. 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.

  2. 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

  3. Amino acid chirality breaking by N-phosphorylation

    International Nuclear Information System (INIS)

    Zhao Yufen; Yan Qingjin.

    1995-01-01

    The chirality breaking of amino acid is a focus issue in the origin of life. For chemists, there are some interesting chemical approaches to solve the symmetry breaking problem. Our previous experiments indicated that when amino acids were phosphorylated, there were many bio-mimic reactions happened. In this paper, it was found that there had significant difference between the N-phosphoryl L- and D- amino acids such as serine and threonine. The optical rotation tracing experiments of the racemic N-phosphoamino acids also showed the similar results. The chirality breaking of amino acids by N-phosphorylation was a novel phenomena. (author). 3 refs, 1 fig. Abstract only

  4. 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....

  5. 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)

  6. 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)

  7. 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 ...

  8. 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 ...

  9. 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 ...

  10. 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...

  11. 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...

  12. 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

  13. 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.

  14. 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 ...

  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. 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.

  17. 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}

  18. 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....

  19. Big break for charge symmetry

    CERN Document Server

    Miller, G A

    2003-01-01

    Two new experiments have detected charge-symmetry breaking, the mechanism responsible for protons and neutrons having different masses. Symmetry is a crucial concept in the theories that describe the subatomic world because it has an intimate connection with the laws of conservation. The theory of the strong interaction between quarks - quantum chromodynamics - is approximately invariant under what is called charge symmetry. In other words, if we swap an up quark for a down quark, then the strong interaction will look almost the same. This symmetry is related to the concept of sup i sospin sup , and is not the same as charge conjugation (in which a particle is replaced by its antiparticle). Charge symmetry is broken by the competition between two different effects. The first is the small difference in mass between up and down quarks, which is about 200 times less than the mass of the proton. The second is their different electric charges. The up quark has a charge of +2/3 in units of the proton charge, while ...

  20. 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

  1. 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.

  2. 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.)

  3. 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.)

  4. 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...

  5. Spontaneous Breaking of Spatial and Spin Symmetry in Spinor Condensates

    DEFF Research Database (Denmark)

    Scherer, M.; Lücke, B.; Gebreyesus, G.

    2010-01-01

    Parametric amplification of quantum fluctuations constitutes a fundamental mechanism for spontaneous symmetry breaking. In our experiments, a spinor condensate acts as a parametric amplifier of spin modes, resulting in a twofold spontaneous breaking of spatial and spin symmetry in the amplified...... broken, but phase squeezing prevents spin-symmetry breaking. If, however, nondegenerate spin modes contribute to the amplification, quantum interferences lead to spin-dependent density profiles and hence spontaneously formed patterns in the longitudinal magnetization....... clouds. Our experiments permit a precise analysis of the amplification in specific spatial Bessel-like modes, allowing for the detailed understanding of the double symmetry breaking. On resonances that create vortex-antivortex superpositions, we show that the cylindrical spatial symmetry is spontaneously...

  6. 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)

  7. Installation for rewetting experiments

    International Nuclear Information System (INIS)

    Rezende, H.C.; Ladeira, L.C.D.

    1986-03-01

    A test facility for rewetting experiments (ITR), has been erected at the Thermalhydraulics Laboratory of Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigation of basic phenomena that occur during the reflood phase of a Loss of Coolant Accident (LOCA), in a Pressurized Water Reactor (PWR), utilizing tubular and annular test sections. The mechanical aspects of the facility, its power supply system and its instrumentation are described. The results of the calibration of the instruments and the description of two typical testes performed to verify the operational conditions are presented. A comparison with calculations using a computer code is also presented. (Author) [pt

  8. 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.)

  9. Breaking Bad News in Veterinary Medicine.

    Science.gov (United States)

    Nickels, Bonnie McCracken; Feeley, Thomas Hugh

    2017-06-16

    The patient-provider relationship in the context of veterinary medicine represents a unique opportunity for studying how bad news is communicated to pet owners by conducting structured interviews with veterinarians. A sample of 44 veterinarians' responses was recorded and content-analyzed in an effort to identify themes among providers in their clinical experience of breaking bad news (BBN). Two coders revealed several themes in the data that were organized by three overarching areas: (1) breaking bad news in general, (2) euthanasia, and (3) social support. The findings from interviews indicated the COMFORT model (Villagran, Goldsmith, Wittenberg-Lyles, & Baldwin, 2010) in medical education provided a useful framework to organize the communication of BBN in veterinary medicine. Results were discussed in relation to future research in patient-provider communication and COMFORT's potential value for training students in veterinarian education.

  10. Electroweak symmetry breaking: Higgs/whatever

    International Nuclear Information System (INIS)

    Chanowitz, M.S.

    1989-01-01

    In the first of these two lectures the Higgs mechanism is reviewed in its most general form, which does not necessarily require the existence of Higgs bosons. The general consequences of the hypothesis that electroweak symmetry breaking is due to the Higgs mechanism are deduced just from gauge invariance and unitarity. In the second lecture the general properties are illustrated with three specific models: the Weinberg-Salam model, its minimal supersymmetric extension, and technicolor. The second lecture concludes with a discussion of the experiment signals for strong WW scattering, whose presence or absence will allow us to determine whether the symmetry breaking sector lies above or below 1 TeV. 57 refs

  11. 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)

  12. Breaking car use habits

    DEFF Research Database (Denmark)

    Thøgersen, John; Møller, Berit Thorup

    2008-01-01

    Based on calls for innovative ways of reducing car traffic and research indicating that car driving is often the result of habitual decision-making and choice processes, this paper reports on a field experiment designed to test a tool aimed to entice drivers to skip the habitual choice of the car...... and consider using-or at least trying-public transport instead. About 1,000 car drivers participated in the experiment either as experimental subjects, receiving a free one-month travelcard, or as control subjects. As predicted, the intervention had a significant impact on drivers' use of public transport...... and it also neutralized the impact of car driving habits on mode choice. However, in the longer run (i.e., four months after the experiment) experimental subjects did not use public transport more than control subjects. Hence, it seems that although many car drivers choose travel mode habitually, their final...

  13. 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.

  14. Taking a break

    DEFF Research Database (Denmark)

    Lundgaard Andersen, Linda; Zukas, Miriam

    2011-01-01

    This chapter focuses on the doctoral summer school as a challenging pedagogy for doctoral education, in which the traditional supervisory relationship and the disciplinary curriculum are deconstructed through intensive group processes. We draw on our experiences as pedagogues at the Roskilde Univ...

  15. 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.)

  16. Breaking the waves

    DEFF Research Database (Denmark)

    Neergaard, Helle; Christensen, Dorthe Refslund

    2015-01-01

    Learning is related to the environment created for the learning experience. This learning environment is often highly routinized and involves a certain social structure but in entrepreneurship education such routinization and structure may actually counteract the learning goals. The purpose...... an ‘identity shaping dimension in that they establish order by creating social feelings ensuring unity through emotional and symbolic dimensions’. Many rituals used in the classroom originate from behaviourism but in entrepreneurship education we often replace these with new ones. The paper builds on extensive...... of this paper is therefore to investigate how classroom routines and rituals impact on entrepreneurship education in order to problematize how existing classroom environments may hinder critical learning experiences from taking place. Previously, rituals have predominantly been examined with an emphasis...

  17. Breaking the Waves:

    DEFF Research Database (Denmark)

    Neergaard, Helle; Christensen, Dorthe Refslund

    2017-01-01

    Learning is related to the environment created for the learning experience. This learning environment is often highly routinized and involves a certain social structure but in entrepreneurship education such routinization and structure may actually counteract the learning goals. The purpose...... an ‘identity shaping dimension in that they establish order by creating social feelings ensuring unity through emotional and symbolic dimensions’. Many rituals used in the classroom originate from behaviourism but in entrepreneurship education we often replace these with new ones. The paper builds on extensive...... of this paper is therefore to investigate how classroom routines and rituals impact on entrepreneurship education in order to problematize how existing classroom environments may hinder critical learning experiences from taking place. Previously, rituals have predominantly been examined with an emphasis...

  18. Ideas that break through

    CERN Multimedia

    Antonella Del Rosso

    2013-01-01

    The EU-cofunded project ULICE (Union of Light Ion Centres in Europe) was launched in 2009 in response to the need to share clinical experience in hadron therapy treatment in Europe and knowledge of the associated complex technical aspects. After four successful years of activity the project is now over but the “transnational access” idea will survive thanks to an extension granted by the European Commission.   A treatment room at CNAO, the Italian centre for hadron therapy. CNAO is participating in ULICE’s transnational access initiative. Image: CNAO. Until a few years ago, the landscape of hadron therapy in Europe was advancing in a fragmented way and facilities were being built without a common shared approach. EU-cofunded projects such as ENLIGHT, ULICE, PARTNER, ENVISION and ENTERVISION helped to build a unified platform where the different – private and public – stakeholders were able to share their views and practical experience in the ...

  19. 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.

  20. Phenomenology of GUT-less Supersymmetry Breaking

    CERN Document Server

    Ellis, Jonathan Richard; Sandick, Pearl

    2007-01-01

    We study models in which supersymmetry breaking appears at an intermediate scale, M_{in}, below the GUT scale. We assume that the soft supersymmetry-breaking parameters of the MSSM are universal at M_{in}, and analyze the morphology of the constraints from cosmology and collider experiments on the allowed regions of parameter space as M_{in} is reduced from the GUT scale. We present separate analyses of the (m_{1/2},m_0) planes for tan(beta)=10 and tan(beta)=50, as well as a discussion of non-zero trilinear couplings, A_0. Specific scenarios where the gaugino and scalar masses appear to be universal below the GUT scale have been found in mirage-mediation models, which we also address here. We demand that the lightest neutralino be the LSP, and that the relic neutralino density not conflict with measurements by WMAP and other observations. At moderate values of M_{in}, we find that the allowed regions of the (m_{1/2},m_0) plane are squeezed by the requirements of electroweak symmetry breaking and that the ligh...

  1. 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)

  2. 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)

  3. Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

    International Nuclear Information System (INIS)

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

    1981-06-01

    This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

  4. 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)

  5. 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 ...

  6. Breaking the glass ceiling.

    Science.gov (United States)

    Lazarus, A

    1997-03-01

    The glass ceiling is a form of organizational bias and discrimination that prevents qualified professionals from achieving positions of top governance and leadership. This article examines glass ceiling barriers that keep physicians from the upper reaches of management. While these factors apply mainly to women and minority physicians in academia, and are attributable to sexual harassment and discrimination, physicians as a class are frequently denied executive management positions. Such denial results from inadequate preparation for a career in health care administration. Important issues in the professional development of physician executives include mentoring, training and education, administrative experience, and cultural and personality factors. All of those must be considered when making the transition from medicine to management.

  7. 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

  8. 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.)

  9. 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

  10. 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.

  11. Breaking the Rules

    CERN Multimedia

    2015-01-01

    This week it's the turn of heavy-ion physics to take the spotlight as the Quark Matter 2015 conference takes place in Kobe, Japan. This is the year’s most important conference for the ALICE collaboration, but there have also been many results presented by ATLAS, CMS and LHCb.   ALICE presented a wide range of results elucidating the behaviour of the hot, strongly interacting state of matter produced when conditions mimicking those present in the first instants after the Big Bang are recreated in lead-ion collisions at the LHC. Taken together with the lead-ion studies carried out by the other LHC experiments, these have significantly advanced our understanding of the nascent Universe. Further details can be found here. Next week sees a very different kind of conference with the third edition of TEDxCERN. As with previous editions, this is CERN’s chance to showcase science and the essential role it plays, and must continue to play, in all areas of socie...

  12. 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

  13. 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

  14. 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

  15. Data report for ROSA-IV LSTF gravity-driven safety injection experiment run SB-CL-27

    International Nuclear Information System (INIS)

    Yonomoto, Taisuke; Saitou, Seishi; Kuroda, Takeshi

    1994-03-01

    Experimental data are presented for the passive injection test, Run SB-CL-27, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on September 17, 1992. This experiment simulated thermal-hydraulic behavior of a gravity-driven, passive safety injection system during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The injection system consisted of a gravity-driven injection tank, located above the reactor vessel, with connecting lines. The tank was initially filled with water of room temperature at the same pressure as the pressurizer. The connecting lines to the cold leg and to the vessel downcomer were opened at the test initiation. Then, a natural circulation flow developed in the loop which was formed by these lines and the injection tank. The hot water in the cold leg circulated into the upper part of tank and accumulated there causing a significant thermal stratification. This thermal stratification prevented direct-contact condensation of steam from occurring during the subsequent tank drain-down phase. Therefore, no condensation-induced depressurization of the tank, affecting adversely the injection performance, occurred. (author)

  16. Breaking Rules – Making Bonds

    Indian Academy of Sciences (India)

    IAS Admin

    RESONANCE | January 2016. GENERAL | ARTICLE. Breaking Rules – Making Bonds. A G Samuelson. Boron-containing molecules discovered recently have new types of dative bonds between carbenes and borylenes. At the same time, they show that traditional thumb rules regarding acids and bases are no longer valid.

  17. Oil prices: Breaks and trends

    International Nuclear Information System (INIS)

    Noguera, José

    2013-01-01

    This paper contributes to the literature of the stationarity of financial time series and the literature on oil and macroeconomics in several ways. First, it uses Kejriwal and Perron (2010) sequential procedure to endogenously determine multiple structural changes in real oil prices without facing the circular testing problem between structural changes and stationary assumptions of previous tests. Second, it performs a diagnostic check to detect the significance and magnitude of the potential breaks. Third, it uses the above information to test for the existence of stochastic trends in real oil prices, and fourth, it speculates about possible explanations for the break dates found in order to encourage further work and discussions. The exercise uses monthly data from January 1861 to August 2011. - Highlights: ► The model endogenously determine multiple structural changes in real oil prices. ► The methods used does not face the circular testing problem. ► It also detect the significance and magnitude of the breaks detected. ► It tests for the existence of stochastic trends. ► It explains the reasons for the break dates found

  18. Breaking Bad News to Parents.

    Science.gov (United States)

    Miller, Susan A.

    1996-01-01

    Discusses the difficulty of breaking bad news to parents, whether the news pertains to center policy or a child's behavior. Provides strategies for presenting news and for helping parents to overcome difficult situations, including gathering facts in advance, arranging an appropriate time, and having resource materials available for parents. (MOK)

  19. Inflationary implications of supersymmetry breaking

    NARCIS (Netherlands)

    Borghese, Andrea; Roest, Diederik; Zavala, Ivonne

    2013-01-01

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

  20. Code breaking in the pacific

    CERN Document Server

    Donovan, Peter

    2014-01-01

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

  1. Breaking Rules–Making Bonds

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 21; Issue 1. Breaking Rules - Making Bonds. A G Samuelson. General Article Volume 21 Issue 1 January 2016 pp 43- ... Author Affiliations. A G Samuelson1. Department of Inorganic and Physical Chemistry, Indian Institute of Science, Bengaluru 560 012 ...

  2. Tight bounds for break minimization

    NARCIS (Netherlands)

    Brouwer, Andries E.; Post, Gerhard F.; Woeginger, Gerhard

    We consider round-robin sports tournaments with n teams and n − 1 rounds. We construct an infinite family of opponent schedules for which every home-away assignment induces at least 1/4 n(n−2) breaks. This construction establishes a matching lower bound for a corresponding upper bound from the

  3. supersymmetry breaking with extra dimensions

    Indian Academy of Sciences (India)

    mechanism for (super)symmetry breaking, proposed first by Scherk and Schwarz, where extra dimensions play a crucial role. The last part is devoted to the description of some recent results and of some open problems. Keywords. Supersymmetry; supergravity; extra dimensions. PACS Nos 11.25.Мj; 11.25.Wx; 11.25.

  4. Breaking Carbon Lock-in

    DEFF Research Database (Denmark)

    Driscoll, Patrick Arthur

    2014-01-01

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

  5. Parental Break-Ups and Stress

    DEFF Research Database (Denmark)

    Dissing, Agnete S.; Dich, Nadya; Nybo Andersen, Anne-Marie

    2017-01-01

    Background: Parental break-up is wide spread, and the effects of parental break-up on children’s well-being are known. The evidence regarding child age at break-up and subsequent family arrangements is inconclusive. Aim: to estimate the effects of parental break-up on stress in pre-adolescent chi......Background: Parental break-up is wide spread, and the effects of parental break-up on children’s well-being are known. The evidence regarding child age at break-up and subsequent family arrangements is inconclusive. Aim: to estimate the effects of parental break-up on stress in pre......-adolescent children with a specific focus on age at break-up and post-breakup family arrangements. Methods: We used data from the Danish National Birth Cohort. Participants included 44 509 children followed from birth to age 11. Stress was self-reported by children at age 11, when the children also reported...... on parental break-up and post break-up family arrangements. Results: Twenty-one percent of the children had experienced a parental break-up at age 11, and those who had experienced parental break-up showed a higher risk of stress (OR:1.72, 95%CI:1.55;1.91) regardless of the child’s age at break-up. Children...

  6. A model of intrinsic symmetry breaking

    International Nuclear Information System (INIS)

    Ge, Li; Li, Sheng; George, Thomas F.; Sun, Xin

    2013-01-01

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

  7. Symmetry breaking: The standard model and superstrings

    International Nuclear Information System (INIS)

    Gaillard, M.K.

    1988-01-01

    The outstanding unresolved issue of the highly successful standard model is the origin of electroweak symmetry breaking and of the mechanism that determines its scale, namely the vacuum expectation value (vev)v that is fixed by experiment at the value v = 4m//sub w//sup 2///g 2 = (√2G/sub F/)/sup /minus/1/ ≅ 1/4 TeV. In this talk I will discuss aspects of two approaches to this problem. One approach is straightforward and down to earth: the search for experimental signatures, as discussed previously by Pierre Darriulat. This approach covers the energy scales accessible to future and present laboratory experiments: roughly (10/sup /minus/9/ /minus/ 10 3 )GeV. The second approach involves theoretical speculations, such as technicolor and supersymmetry, that attempt to explain the TeV scale. 23 refs., 5 figs

  8. Symmetry breaking: The standard model and superstrings

    Energy Technology Data Exchange (ETDEWEB)

    Gaillard, M.K.

    1988-08-31

    The outstanding unresolved issue of the highly successful standard model is the origin of electroweak symmetry breaking and of the mechanism that determines its scale, namely the vacuum expectation value (vev)v that is fixed by experiment at the value v = 4m//sub w//sup 2///g/sup 2/ = (..sqrt..2G/sub F/)/sup /minus/1/ approx. = 1/4 TeV. In this talk I will discuss aspects of two approaches to this problem. One approach is straightforward and down to earth: the search for experimental signatures, as discussed previously by Pierre Darriulat. This approach covers the energy scales accessible to future and present laboratory experiments: roughly (10/sup /minus/9/ /minus/ 10/sup 3/)GeV. The second approach involves theoretical speculations, such as technicolor and supersymmetry, that attempt to explain the TeV scale. 23 refs., 5 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bianco, Andrea

    2015-07-23

    An experimental investigation was conducted in hot cells on single fuel rod segments to appraise the behavior of fuel pellets fragmentation during a loss of coolant accident in a light water reactor. In pursuing the conceptual design of the experiment, calculations were performed to study the thermal-hydraulics boundary conditions and the fuel rod behavior during the transient. The experiment's results encompass non-destructive and destructive examinations. In order to describe the resulting fuel fragments size distribution, a semi-empirical correlation was derived from the fractal theory.

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

    International Nuclear Information System (INIS)

    Bianco, Andrea

    2015-01-01

    An experimental investigation was conducted in hot cells on single fuel rod segments to appraise the behavior of fuel pellets fragmentation during a loss of coolant accident in a light water reactor. In pursuing the conceptual design of the experiment, calculations were performed to study the thermal-hydraulics boundary conditions and the fuel rod behavior during the transient. The experiment's results encompass non-destructive and destructive examinations. In order to describe the resulting fuel fragments size distribution, a semi-empirical correlation was derived from the fractal theory.

  11. ISP-27 OECD/NEA/CSNI International Standard Problem n.27. Bethsy experiment 9.1 B. 2. cold leg break without HPSI and with delayed ultimate procedure. Comparison report. Volume 1 + 2

    International Nuclear Information System (INIS)

    1992-11-01

    This report is the final comparison report for ISP-27, a blind problem which is based on the BETHSY test 9.1b performed in december 1989 at the Nuclear Research Center in Grenoble (France). The BETHSY integral test facility is a scaled down model of a 3 loop 900 e MW FRAMATOME PWR; the overall scaling factor applied to every volume, mass flowrate and power level is close to 1/100, while elevations are 1/1 in order to preserve the gravitational heads. The cold leg break is combined with the High Pressure Injection System (HPIS) failure. In that case, the state oriented approach requires operators to start an Ultimate Procedure, which consists in fully opening the Steam Generator (SG) atmospheric dumps as soon as they are informed of the unavailability of the HPIS. The presently studied scenario assumes a delayed application of this procedure, which is started only when the core outlet temperature rises significantly higher than the saturation temperature. The BETHSY Test 9.1b addresses, besides typical problems relevant to Small Break Loss Of Coolant Accidents (SBLOCA) such as critical 2-phase flow, loop seal clearing, heat-transfer during boil-off or accumulator injection, specific aspects related to the fast depressurization (primary to secondary and structural heat transfer), uncovered core behavior when intense condensation takes place in the SG, and primary side refilling by the Low Pressure Injection System (LPIS)

  12. Always Running. La Vida Loca: Gang Days in L.A.

    Science.gov (United States)

    Rodriguez, Luis J.

    This autobiographical narrative describes the early life of Luis J. Rodriguez, a journalist and poet who was immersed in the youth gang culture of Los Angeles (California). Framed by the story of the pull of the gang life for the poet's son, it recounts his experiences from his childhood on the United States-Mexico border through his family's…

  13. 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...

  14. ArtBreak: A Creative Group Counseling Program for Children

    Science.gov (United States)

    Ziff, Katherine; Pierce, Lori; Johanson, Susan; King, Margaret

    2012-01-01

    This article describes the pilot of a school-based creative group-counseling program for children called ArtBreak, a choice-based studio art experience based on the restorative possibilities of art making delineated in the expressive therapies continuum (ETC; Kagin & Lusebrink, 1978). The ETC features a developmental hierarchy in relation to how…

  15. Electroweak Symmetry Breaking without Higgs Bosons at LHC

    CERN Document Server

    Delsart, P A

    2007-01-01

    It is possible that Electroweak Symmetry Breaking does not occur in Nature through the Higgs mechanism. Several alternate scenarios are studied at LHC experiment and this presentation review some of them : Technicolor searches in CMS and and Vector Boson Scattering in the Chiral Lagrangian model or in extra-dimension model in Atlas.

  16. Breaking through the tranfer tunnel

    CERN Document Server

    Laurent Guiraud

    2001-01-01

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

  17. Models of electroweak symmetry breaking

    CERN Document Server

    Pomarol, Alex

    2015-01-01

    This chapter present models of electroweak symmetry breaking arising from strongly interacting sectors, including both Higgsless models and mechanisms involving a composite Higgs. These scenarios have also been investigated in the framework of five-dimensional warped models that, according to the AdS/CFT correspondence, have a four-dimensional holographic interpretation in terms of strongly coupled field theories. We explore the implications of these models at the LHC.

  18. Experimental thermal hydraulic facility for simulating LOCA behaviour of pressurised heavy water power reactor

    International Nuclear Information System (INIS)

    Sahu, M.K.; John, P.K.; Jayaraj, N.; Chatterjee, P.B.; Das, Sandeep; John, Benny; Sharma, A.K.; Prasad, N.; Singhal, M.; Malhotra, P.K.; Haldar, S.C.; Bhambra, H.S.; Chadda, S.K.; Chandra, Umesh

    2006-01-01

    Experimental thermal hydraulic facility being set up adjacent to R and D Centre at Tarapur is a 13 MW full-elevation scaled down facility having the key components of PHT System of Pressurised Heavy Water Reactor (PHWR). The objective of the facility is to study thermal hydraulic behaviour of PHT System of PHWR by simulating various transients and accidental scenarios, to conduct safety related and operational transient studies and validation of various thermal hydraulic computer codes developed for analysis. The design of thermal hydraulic facility is based on the process parameters of a large PHWR with respect to fluid mass flux, transit time, flow velocity, pressure, temperature and enthalpy in PHT System. Experiments would be conducted in the facility to gain an improved understanding of the thermal hydraulic behaviour of large size PHWR during loss of coolant accident scenarios with forced and natural thermo-siphoning circulation modes etc. The data collected from the experiments would be used in validating computer codes developed for safety analysis. The facility is extensively instrumented to measure parameters such as temperature, pressure, flow, level, void-fraction at key locations. This paper gives the design philosophy used for scaling, design of major components of primary and secondary circuit of Experimental Thermal Hydraulic Facility and details of simulated experiments to be carried out. (author)

  19. Supersymmetry breaking at finite temperature

    International Nuclear Information System (INIS)

    Kratzert, K.

    2002-11-01

    The mechanism of supersymmetry breaking at finite temperature is still only partly understood. Though it has been proven that temperature always breaks supersymmetry, the spontaneous nature of this breaking remains unclear, in particular the role of the Goldstone fermion. The aim of this work is to unify two existing approaches to the subject. From a hydrodynamic point of view, it has been argued under very general assumptions that in any supersymmetric quantum field theory at finite temperature there should exist a massless fermionic collective excitation, named phonino because of the analogy to the phonon. In the framework of a self-consistent resummed perturbation theory, it is shown for the example of the Wess-Zumino model that this mode fits very well into the quantum field theoretical framework pursued by earlier works. Interpreted as a bound state of boson and fermion, it contributes to the supersymmetric Ward-Takahashi identities in a way showing that supersymmetry is indeed broken spontaneously with the phonino playing the role of the Goldstone fermion. The second part of the work addresses the case of supersymmetric quantum electrodynamics. It is shown that also here the phonino exists and must be interpreted as the Goldstone mode. This knowledge allows a generalization to a wider class of models. (orig.)

  20. Estimation of heat transfer rates to droplets under the conditions of a LOCA in a PWR in the ballooned zone

    Energy Technology Data Exchange (ETDEWEB)

    Gradeck, Michel; Maillet, Denis [LEMTA Nancy-University CNRS, 2 av de la foret de Haye, BP160, 54504 Vandoeuvre cedex (France); Lelong, Franck [LEMTA Nancy-University CNRS, 2 av de la foret de Haye, BP160, 54504 Vandoeuvre cedex (France)]|[IRSN/DPAM/SEMCA/LEIDC, Cadarache Batiment 700, BP3 - 13 115 Saint Paul lez Durance cedex (France); Seiler, Nathalie [IRSN/DPAM/SEMCA/LEIDC, Cadarache Batiment 700, BP3 - 13 115 Saint Paul lez Durance cedex (France)

    2008-07-01

    Full text of publication follows: During a LOCA (Loss Of Coolant Accident), the critical regions (in terms of safety) of the fuel assemblies could be ballooned. The cooling of theses partially blocked fuel assemblies depends on the coolant flow characteristics in the blockage region. Most models for heat transfer concentrate on cooling of the ballooned walls by vapor convection. Since a two-phase mist flow occurs when reflooding, the possibility of additional cooling by direct liquid droplet impingement on the blockage surfaces must be investigated. As the temperature of the fuel assemblies is higher than the Leidenfrost temperature, the impact regime should be only the bouncing one. Up to now, no model of heat transfer of droplet impacts has been developed for that regime. As the coolability from droplet impacts must be modeled, we realize an experimental study with droplets and wall characteristics (velocity, diameter, temperature) close to the LOCA ones. As the interaction between the droplet and the wall is very short (a few of ms), the estimation of the heat flux during the resident time of the droplet at the wall must be accurately designed. The purpose of this work is to show how such heat flux can be experimentally estimated used an adapted inverse heat conduction model. The final goal of the present collaboration between LEMTA (Laboratory of Applied and Theoretical Energy and Mechanic) and IRSN (Institut of Radioprotection and Nuclear Safety) is to introduce the cooling model within NEPTUNE-CFD code of the NEPTUNE thermal-hydraulic platform, a joint project of CEA, EDF, IRSN and AREVA. (authors) [French] Dans le cas d'un APRP (Accident de Perte de Refrigerant Primaire), les zones critiques de l'assemblage combustible peuvent etre deformees. Le refroidissement de ces zones depend de l'importance du blocage qui affectera l'ecoulement diphasique les traversant. La plupart des modeles de refroidissement de ces zones assechees, a hautes

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  2. Dense Gravity Currents with Breaking Internal Waves

    Science.gov (United States)

    Tanimoto, Yukinobu; Hogg, Charlie; Ouellette, Nicholas; Koseff, Jeffrey

    2017-11-01

    Shoaling and breaking internal waves along a pycnocline may lead to mixing and dilution of dense gravity currents, such as cold river inflows into lakes or brine effluent from desalination plants in near-coastal environments. In order to explore the interaction between gravity currents and breaking interfacial waves a series of laboratory experiments was performed in which a sequence of internal waves impinge upon a shelf-slope gravity current. The waves are generated in a two-layer thin-interface ambient water column under a variety of conditions characterizing both the waves and the gravity currents. The mixing of the gravity current is measured through both intrusive (CTD probe) and nonintrusive (Planar-laser inducted fluorescence) techniques. We will present results over a full range of Froude number (characterizing the waves) and Richardson number (characterizing the gravity current) conditions, and will discuss the mechanisms by which the gravity current is mixed into the ambient environment including the role of turbulence in the process. National Science Foundation.

  3. Non-universal SUSY breaking, hierarchy and squark degeneracty

    International Nuclear Information System (INIS)

    Murayama, Hitoshi.

    1995-01-01

    I discuss non-trivial effects in the soft SUSY breaking terms which appear when one integrates out heavy fields. The effects exist only when the SUSY breaking terms are non-universal. They may spoil (1) the hierarchy between the weak and high-energy scales, or (2) degeneracy among the squark masses even in the presense of a horizontal symmetry. I argue, in the end, that such new effects may be useful in probing physics at high-energy scales from TeV-scale experiments

  4. CNN for breaking text-based CAPTCHA with noise

    Science.gov (United States)

    Liu, Kaixuan; Zhang, Rong; Qing, Ke

    2017-07-01

    A CAPTCHA ("Completely Automated Public Turing test to tell Computers and Human Apart") system is a program that most humans can pass but current computer programs could hardly pass. As the most common type of CAPTCHAs , text-based CAPTCHA has been widely used in different websites to defense network bots. In order to breaking textbased CAPTCHA, in this paper, two trained CNN models are connected for the segmentation and classification of CAPTCHA images. Then base on these two models, we apply sliding window segmentation and voting classification methods realize an end-to-end CAPTCHA breaking system with high success rate. The experiment results show that our method is robust and effective in breaking text-based CAPTCHA with noise.

  5. Effects of intact loop hydraulic resistance of PWR LOCA behavior in scaled experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jacoby, M.S.

    1977-09-01

    The scaling of experiments in the Water Reactor Safety Program has been on a power/volume basis. This scaling philosophy has resulted in good thermal modeling of the core, but, combined with core design considerations, compromises the modeling of the hydraulic resistance in the intact loop. Tests in LOFT, Semiscale MOD-1 and FLECHT-SET have been conducted for the purpose of determining the effect of scaling hydraulic resistance to core area ratio (low hydraulic resistance) or to core power ratio (high hydraulic resistance). The results of these tests, together with computer model results using RELAP4/MOD5, RE-FRAP, and FLOOD4, were studied to determine the effects of the scaling compromise. The review of available information has shown no significant results of varying intact loop hydraulic resistance.

  6. Passive appendages generate drift through symmetry breaking

    Science.gov (United States)

    Lācis, U.; Brosse, N.; Ingremeau, F.; Mazzino, A.; Lundell, F.; Kellay, H.; Bagheri, S.

    2014-10-01

    Plants and animals use plumes, barbs, tails, feathers, hairs and fins to aid locomotion. Many of these appendages are not actively controlled, instead they have to interact passively with the surrounding fluid to generate motion. Here, we use theory, experiments and numerical simulations to show that an object with a protrusion in a separated flow drifts sideways by exploiting a symmetry-breaking instability similar to the instability of an inverted pendulum. Our model explains why the straight position of an appendage in a fluid flow is unstable and how it stabilizes either to the left or right of the incoming flow direction. It is plausible that organisms with appendages in a separated flow use this newly discovered mechanism for locomotion; examples include the drift of plumed seeds without wind and the passive reorientation of motile animals.

  7. Passive appendages aid locomotion through symmetry breaking

    Science.gov (United States)

    Bagheri, Shervin; Lacis, Ugis; Mazzino, Andrea; Kellay, Hamid; Brosse, Nicolas; Lundell, Fredrik; Ingremeau, Francois

    2014-11-01

    Plants and animals use plumes, barbs, tails, feathers, hairs, fins, and other types of appendages to aid locomotion. Despite their enormous variation, passive appendages may contribute to locomotion by exploiting the same physical mechanism. We present a new mechanism that applies to body appendages surrounded by a separated flow, which often develops behind moving bodies larger than a few millimeters. We use theory, experiments, and numerical simulations to show that bodies with protrusions turn and drift by exploiting a symmetry-breaking instability similar to the instability of an inverted pendulum. Our model explains why the straight position of an appendage in flowing fluid is unstable and how it stabilizes either to the left or right of the incoming fluid flow direction. The discovery suggests a new mechanism of locomotion that may be relevant for certain organisms; for example, how plumed seeds may drift without wind and how motile animals may passively reorient themselves.

  8. Small Neutrino Masses from Supersymmetry Breaking

    Energy Technology Data Exchange (ETDEWEB)

    Arkani-Hamed, Nima; Hall, Lawrence; Murayama, Hitoshi; Smith, David; Weiner, Neal

    2000-06-27

    An alternative to the conventional see-saw mechanism is proposed to explain the origin of small neutrino masses in supersymmetric theories. The masses and couplings of the right-handed neutrino field are suppressed by supersymmetry breaking, in a way similar to the suppression of the Higgs doublet mass, $\\mu$. New mechanisms for light Majorana, Dirac and sterile neutrinos arise, depending on the degree of suppression. Superpartner phenomenology is greatly altered by the presence of weak scale right-handed sneutrinos, which may have a coupling to a Higgs boson and a left-handed sneutrino. The sneutrino spectrum and couplings are quite unlike the conventional case - the lightest sneutrino can be the dark matter and predictions are given for event rates at upcoming halo dark matter direct detection experiments. Higgs decays and search strategies are changed. Copious Higgs production at hadron colliders can result from cascade decays of squarks and gluinos.

  9. Inflationary implications of supersymmetry breaking

    Energy Technology Data Exchange (ETDEWEB)

    Borghese, Andrea; Roest, Diederik; Zavala, Ivonne [Centre for Theoretical Physics, University of Groningen, Nijenborgh 4, 9747 AG Groningen (Netherlands)

    2013-07-23

    We discuss a general bound on the possibility to realise inflation in any minimal supergravity with F-terms. The derivation crucially depends on the sGoldstini, the scalar field directions that are singled out by spontaneous supersymmetry breaking. The resulting bound involves both slow-roll parameters and the geometry of the Kähler manifold of the chiral scalars. We analyse the inflationary implications of this bound, and in particular discuss to what extent the requirements of single field and slow-roll can both be met in F-term inflation.

  10. Leaders break ground for INFINITY

    Science.gov (United States)

    2008-01-01

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

  11. The development of a burst criterion for zircaloy fuel cladding under LOCA conditions

    International Nuclear Information System (INIS)

    Neitzel, H.J.; Rossinger, H.E.

    1980-02-01

    A burst criterion model, which assumes that deformation is controlled by steady-state creep, has been developed for a thin-walled cladding, in this case Zircaloy-4, subjected to a differential pressure and high temperature. The creep equation is integrated to obtain a burst time at the singularity of the strain. Once the burst time is known, the burst temperature and burst pressure can be calculated from the known temperature and pressure histories. A further relationship between burst stress and burst temperature is used to calculate the burst strain. Comparison with measured burst data shows good agreement between theory and experiment was found that, if the heating rate is constant, the burst temperature increases with decreasing stress, and that, if the stress level is constant, the burst temperature increases with increasing heating rate. It was also found that anisotropy alters the burst temperature and burst strain, and that test conditions in the α-Zr temperature range have no influence on the burst data. (auth)

  12. The development of a burst criterion for Zircaloy fuel cladding under LOCA conditions

    International Nuclear Information System (INIS)

    Neitzel, H.J.; Rosinger, H.E.

    1980-10-01

    A burst criterion model, which assumes that deformation is controlled by steady-state creep, has been developed for a thin-walled cladding, in this case Zircaloy-4, subjected to a differential pressure and high temperature. The creep equation is integrated to obtain a burst time at the singularity of the strain. Once that urst time is known, the burst temperature and burst pressure can be calculated from the known temperature and pressure histories. A further relationship between burst stress and burst temperature is used to calculate the burst strain. Comparison with measured burst data shows good agreement between theory and experiment. It was found that, if the heating rate is constant, the burst temperature increases with decreasing stress, and that, if the stress level is constant, the burst temperature increases with increasing heating rate. It was also found that anisotropy alters the burst temperature and burst strain, and that thest conditions in the α-Zr temperature range have no influence on the burst data. (orig.) [de

  13. Breaking bad news in prenatal medicine: a literature review.

    Science.gov (United States)

    Luz, Rita; George, Astrid; Spitz, Elisabeth; Vieux, Rachel

    2017-02-01

    The diagnosis of a fetal anomaly in perinatal medicine forces expectant parents and healthcare providers to face the difficult process of breaking bad news. This exploratory literature review was aimed at providing a medical and psychological view of the psychological experience in expectant parents and physicians in the context of prenatal diagnosis of a fetal anomaly. An exploratory search of PubMed and PsycINFO/PsycARTICLES databases performed by an interdisciplinary team composed of a physician and psychologists. Search terms were: prenatal diagnosis AND bad news; prenatal diagnosis AND psychological consequences; prenatal diagnosis AND psychological sequelae; prenatal diagnosis AND fetal abnormality. The processing of selected articles followed a standardised five-step procedure. A total of 860 articles were screened of which 32 were retained for analysis. Four main themes emerged from the explanatory content analysis: (1) parents' subjective experience; (2) physicians' subjective experience; (3) encounters between expectant parents and professionals; and (4) ethical challenges in breaking bad news in prenatal medicine. Expectant parents go through a complex and multidimensional experience when the diagnosis of a fetal anomaly is disclosed. Simultaneously, physicians consider breaking bad news as a very stressful event and are poorly prepared in this regard. A better knowledge of factors underlying psychological adjustment of the parental dyad and on the subjective experience of physicians delivering these diagnoses could enable better adaptation for both patients and professionals.

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

    International Nuclear Information System (INIS)

    Gerardin, J.

    2012-01-01

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

  15. Induction and repair of double- and single-strand DNA breaks in bacteriophage lambda superinfecting Escherichia coli

    International Nuclear Information System (INIS)

    Boye, E.; Krisch, R.E.

    1980-01-01

    Induction and repair of double-and single-strand DNA breaks have been measured after decays of 125 I and 3 H incorporated into the DNA and after external irradiation with 4 MeV electrons. For the decay experiments, cells of wild type Escherichia coli K-12 were superinfected with bacteriophage lambda DNA labelled with 5'-( 125 I)iodo-2'-deoxyuridine or with (methyl- 3 H)thymidine and frozen in liquid nitrogen. Aliquots were thawed at intervals and lysed at neutral pH, and the phage DNA was assayed for double- and single-strand breakage by neutral sucrose gradient centrifugation. The gradients used allowed measurements of both kinds of breaks in the same gradient. Decays of 125 I induced 0.39 single-strand breaks per double-strand break. No repair of either break type could be detected. Each 3 H disintegration caused 0.20 single-strand breaks and very few double-strand breaks. The single-strand breaks were rapidly rejoined after the cells were thawed. For irradiation with 4 MeV electrons, cells of wild type E. coli K-12 were superinfected with phage lambda and suspended in growth medium. Irradiation induced 42 single-strand breaks per double-strand break. The rates of break induction were 6.75 x 10 -14 (double-strand breaks) and 2.82 x 10 -12 (single-strand breaks) per rad and per dalton. The single-strand breaks were rapidly repaired upon incubation whereas the double-strand breaks seemed to remain unrepaired. It is concluded that double-strand breaks in superinfecting bacteriophage lambda DNA are repaired to a very small extent, if at all. (Author)

  16. Radiation Dose Calculation for a Large Break Loss of Coolant Accident for the Dry Process Fuel Core with a Dual Failure

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jong Ho; Kim, Taek Mo; Choi, Hang Bok

    2005-05-15

    The compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel with the existing 713 MWe CANDU (CANDU-6) reactor has been analyzed for a limiting large break loss of coolant accident (LOCA) scenario such as 100% reactor outlet header break accompanied by a dual failure of the containment isolation logic. For the DUPIC fuel, the radiation source term was calculated for a 1/4 of fission products inventory in the fuel gap of the CANDU-6 reactor being steadily operated at the full power. However it was assumed that all the fission products of the DUPIC fuel core are instantaneously released to the containment building at 3 sec after the break, because the transient release model of the fission products has not yet been developed for the DUPIC fuel. The radiation effect was estimated for the personal dose of the critical age and the public dose. The calculations have shown that the personal doses are 231 mSv and 1954 mSv for the whole body and thyroid, respectively, which are blow the limits of 250 mSv and 2500 mSv. In fact, the personal doses of the DUPIC fuel core are higher than those of the natural uranium core, which is due to the assumption that all the fission products are instantaneously released into the containment building. Therefore if a realistic transient model of the fission products release is used, it is expected that the radiation doses of the DUPIC fuel core are much less that those of the natural uranium core. The public doses are 157 person-Sv and 1929 person-Sv for the whole body and thyroid, respectively, which are much less that the design limit of 10000 person-Sv. This study has confirmed that the personal and public doses of the DUPIC fuel core satisfy the design limits for the large break LOCA accompanied by a dual failure of the containment isolation logic.

  17. Lorentz Symmetry Breaking in Quantum Electrodynamics

    OpenAIRE

    Oliveira, D. M.

    2010-01-01

    In this dissertation, we study the implications generated by the Lorentz breaking symmetry in quantum electrodynamics. We analyze fermions interacting with an electromagnetic field in the contexts of quantum mechanics and make radiative corrections. In quantum mechanics, the terms of the Lorentz breaking symmetry were treated as perturbations to the Dirac equation, and their expected values were obtained in a vacuum. In the radiative corrections, the Lorentz breaking symmetry was introduced i...

  18. Rock breaking methods to replace blasting

    Science.gov (United States)

    Zhou, Huisheng; Xie, Xinghua; Feng, Yuqing

    2018-03-01

    The method of breaking rock by blasting has a high efficiency and the cost is relatively low, but the associated vibration, flyrock, production of toxic gases since the 1970’s, the Western developed countries began to study the safety of breaking rock. This paper introduces different methods and their progress to safely break rock. Ideally, safe rock breaking would have little vibration, no fly stone, and no toxic gases, which can be widely used in municipal engineering, road excavation, high-risk mining, quarrying and complex environment.

  19. Disruption of Perceptual Learning by a Brief Practice Break.

    Science.gov (United States)

    Little, David F; Zhang, Yu-Xuan; Wright, Beverly A

    2017-12-04

    Some forms of associative learning require only a single experience to create a lasting memory [1, 2]. In contrast, perceptual learning often requires extensive practice within a day for performance to improve across days [3, 4]. This suggests that the requisite practice for durable perceptual learning is integrated throughout each day. If the total amount of daily practice is the only important variable, then a practice break within a day should not disrupt across-day improvement. To test this idea, we trained human listeners on an auditory frequency-discrimination task over multiple days and compared the performance of those who engaged in a single continuous practice session each day [4] with those who were given a 30-min break halfway through each practice session. Continuous practice yielded significant perceptual learning [4]. In contrast, practice with a rest break led to no improvement, indicating that the integration process had decayed within 30 min. In a separate experiment, a 30-min practice break also disrupted durable learning on a non-native phonetic classification task. These results suggest that practice trials are integrated up to a learning threshold within a transient memory store before they are sent en masse into a memory that lasts across days. Thus, the oft cited benefits of distributed over massed training [5, 6] may arise from different mechanisms depending on whether the breaks occur before or after a learning threshold has been reached. Trial integration could serve as an early gatekeeper to plasticity, helping to ensure that longer-lasting changes are only made when deemed worthwhile. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Experimental study on fundamental phenomena in HTGR small break air-ingress accident

    International Nuclear Information System (INIS)

    Kim, Jae Soon; Hwang, Jin-Seok; Kim, Eung Soo; Kim, Byung Jun; Oh, Chang Ho

    2016-01-01

    Highlights: • Air-ingress phenomena on the small break in a HTGR are experimentally investigated. • Experiment is investigated for various break sizes, angles, and density ratios. • Maximum air-ingress rate is observed at 120° in break angle. • This study reveals that air-ingress in the small break is governed by; buoyancy and flow inertia. • A non-dimensional parameter is newly proposed to determine the air-ingress flow regimes. • Newly proposed parameter is based on buoyancy versus inertia force. - Abstract: This study experimentally investigates fundamental phenomena in the HTGR small break air-ingress accident. Several important parameters including density ratio, break angle, break size, and main flow velocity are considered in the measurement and the analysis. The test-section is made of a circular pipe with small holes drilled around the surface and it is installed in the helium/air flow circulation loop. Oxygen concentrations and flow rates are recorded during the tests with fixed break angles, break sizes, and flow velocities for measurement of the air-ingress rates. According to the experimental results, the higher density difference leads to the higher rates of air-ingress with large sensitivity of the break angles. It is also found that the break angle significantly affects the air-ingress rates, which is gradually increased from 0° to 120° and suddenly decreased to 180°. The minimum air ingress rate is found at 0° and the maximum, at 110°. The air-ingress rate increases with the break size due to the increased flow-exchange area. However, it is not directly proportional to the break area due to the complexity of the phenomena. The increased flow velocity in the channel inside enhances the air-ingress process. However, among all the parameters, the main flow velocity exhibits the lowest effect on this process. In this study, the Froude Number relevant to the small break air-ingress conditions are newly defined considering both heavy

  1. Lie-algebra approach to symmetry breaking

    International Nuclear Information System (INIS)

    Anderson, J.T.

    1981-01-01

    A formal Lie-algebra approach to symmetry breaking is studied in an attempt to reduce the arbitrariness of Lagrangian (Hamiltonian) models which include several free parameters and/or ad hoc symmetry groups. From Lie algebra it is shown that the unbroken Lagrangian vacuum symmetry can be identified from a linear function of integers which are Cartan matrix elements. In broken symmetry if the breaking operators form an algebra then the breaking symmetry (or symmetries) can be identified from linear functions of integers characteristic of the breaking symmetries. The results are applied to the Dirac Hamiltonian of a sum of flavored fermions and colored bosons in the absence of dynamical symmetry breaking. In the partially reduced quadratic Hamiltonian the breaking-operator functions are shown to consist of terms of order g 2 , g, and g 0 in the color coupling constants and identified with strong (boson-boson), medium strong (boson-fermion), and fine-structure (fermion-fermion) interactions. The breaking operators include a boson helicity operator in addition to the familiar fermion helicity and ''spin-orbit'' terms. Within the broken vacuum defined by the conventional formalism, the field divergence yields a gauge which is a linear function of Cartan matrix integers and which specifies the vacuum symmetry. We find that the vacuum symmetry is chiral SU(3) x SU(3) and the axial-vector-current divergence gives a PCAC -like function of the Cartan matrix integers which reduces to PCAC for SU(2) x SU(2) breaking. For the mass spectra of the nonets J/sup P/ = 0 - ,1/2 + ,1 - the integer runs through the sequence 3,0,-1,-2, which indicates that the breaking subgroups are the simple Lie groups. Exact axial-vector-current conservation indicates a breaking sum rule which generates octet enhancement. Finally, the second-order breaking terms are obtained from the second-order spin tensor sum of the completely reduced quartic Hamiltonian

  2. Systematics in break-up fusion reactions at low energies

    International Nuclear Information System (INIS)

    Singh, B.P.

    2017-01-01

    The reaction mechanism of heavy-ion interactions at low projectile energies is still not well understood. Both the fusion as well as break-up fusion of projectile have been observed at these energies. In the break-up fusion, also referred to as the incomplete fusion (ICF), a part of the incident ion fuses with the target nucleus while the remnant moves forward with the same velocity as that of projectile. Several models have been proposed, however, none of them could reproduce the data on ICF process. In order to study the break-up fusion reactions and to study its influence on various entrance channel parameters, several experiments have been carried out using pelletron accelerator facility at the Inter University Accelerator Centre (IUAC), New Delhi. The analysis of data has indicated significant break-up fusion contributions at energies ≈ 4-7 MeV/nucleon. Important systematic for incomplete or breakup fusion reactions have been observed and will be presented. (author)

  3. The difficulties experienced by nurses and healthcare staff involved in the process of breaking bad news.

    Science.gov (United States)

    Warnock, Clare; Buchanan, Jean; Tod, Angela Mary

    2017-07-01

    The aim of this study was to explore the difficulties experienced by nurses and healthcare professionals when engaging in the process of breaking bad news. The challenges faced by staff when breaking bad news have previously been researched in relation to particular settings or participants. This study involved staff from diverse settings and roles to develop broader insights into the range of difficulties experienced in clinical practice. The study used a descriptive survey design involving self-reported written accounts and framework analysis. Data were collected using a structured questionnaire containing a free text section that asked participants to describe a difficult experience they had encountered when involved in the process of breaking bad news. Data were collected from healthcare staff from hospital, community, hospice and care home settings attending training days on breaking bad news between April 2011 and April 2014. Multiple inter-related factors presented challenges to staff engaging in activities associated with breaking bad news. Traditional subjects such as diagnostic and treatment information were described but additional topics were identified such as the impact of illness and care at the end of life. A descriptive framework was developed that summarizes the factors that contribute to creating difficult experiences for staff when breaking bad news. The framework provides insights into the scope of the challenges faced by staff when they engage in the process of breaking bad news. This provides the foundation for developing interventions to support staff that more closely matches their experiences in clinical practice. © 2017 John Wiley & Sons Ltd.

  4. Effects of Rest-Break Intention on Rest-Break Frequency and Work-Related Fatigue.

    Science.gov (United States)

    Blasche, Gerhard; Pasalic, Sanja; Bauböck, Verena-Maria; Haluza, Daniela; Schoberberger, Rudolf

    2017-03-01

    The present paper presents findings from two studies addressing the effects of the employee's intention to have rest breaks on rest-break frequency and the change of well-being during a workday. Rest breaks are effective in avoiding an accumulation of fatigue during work. However, little is known about individual differences in rest-break behavior. In Study 1, the association between rest-break intention and the daily number of rest breaks recorded over 4 consecutive workdays was determined by generalized linear model in a sample of employees ( n = 111, 59% females). In Study 2, professional geriatric nurses ( n = 95 females) who worked over two consecutive 12-hour day shifts recorded well-being (fatigue, distress, effort motivation) at the beginning and the end of their shifts. The effect of rest-break intention on the change of well-being was determined by multilevel modeling. Rest-break intention was positively associated with the frequency of rest breaks (Study 1) and reduced the increase of fatigue and distress over the workday (Study 2). The results indicate that individual differences account for the number of breaks an employee takes and, as a consequence, for variations in the work-related fatigue and distress. Strengthening rest-break intentions may help to increase rest-break behavior to avoid the buildup of fatigue and distress over a workday.

  5. Breaking antidunes: Cyclic behavior due to hysteresis

    DEFF Research Database (Denmark)

    Deigaard, Rolf

    2006-01-01

    The cyclic behavior of breaking antidunes (growth, breaking of surface wave, obliteration) is investigated by use of a numerical model. The model includes the transition between supercritical and transcritical flow. As the antidune grows the flow becomes transcritical and a hydraulic jump is form...

  6. Breaking Bad News - Perceptions of Pediatric Residents.

    Science.gov (United States)

    Geeta, M G; Krishnakumar, P

    2017-08-15

    The present study evaluated the perceptions and practice of 92 final year pediatric residents with regard to breaking bad news. Only 16% of residents had received any training in communication skills. Majority (65%) of the residents were not comfortable while breaking bad news.

  7. Symmetry breaking signaling mechanisms during cell polarization

    NARCIS (Netherlands)

    Bruurs, LJM

    2017-01-01

    Breaking of cellular symmetry in order to establish an apico-basal polarity axis initiates de novo formation of cell polarity. However, symmetry breaking provides a formidable challenge from a signaling perspective, because by definition no spatial cues are present to instruct axis establishment.

  8. Structural Break Tests Robust to Regression Misspecification

    NARCIS (Netherlands)

    Abi Morshed, Alaa; Andreou, E.; Boldea, Otilia

    2016-01-01

    Structural break tests developed in the literature for regression models are sensitive to model misspecification. We show - analytically and through simulations - that the sup Wald test for breaks in the conditional mean and variance of a time series process exhibits severe size distortions when the

  9. Strongly coupled semidirect mediation of supersymmetry breaking

    International Nuclear Information System (INIS)

    Ibe, M.; Izawa, K.-I.; Nakai, Y.

    2009-01-01

    Strongly coupled semidirect gauge mediation models of supersymmetry breaking through massive mediators with standard-model charges are investigated by means of composite degrees of freedom. Sizable mediation is realized to generate the standard-model gaugino masses for a small mediator mass without breaking the standard-model symmetries.

  10. On breaks of the Indian monsoon

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    2000) but little overlap with breaks identified by Webster et al (1998). Further, there are three or four active-break cycles in a season according to Webster et al (1998) which implies a time scale of about 40 days for which Goswami and Mohan ...

  11. Dynamical study of symmetries: breaking and restauration

    International Nuclear Information System (INIS)

    Schuck, P.

    1986-09-01

    First symmetry breaking (spontaneous) is explained and the physical implication discussed for infinite systems. The relation with phase transitions is indicated. Then the specific aspects of symmetry breaking in finite systems is treated and illustrated in detail for the case of translational invariance with the help of an oversimplified but exactly solvable model. The method of projection (restauration of symmetry) is explained for the static case and also applied to the model. Symmetry breaking in the dynamical case and for instance the notion of a soft mode responsible for the symmetry breaking is discussed in the case of superfluidity and another exactly solvable model is introduced. The Goldstone mode is treated in detail. Some remarks on analogies with the breaking of chiral symmetry are made. Some recent developments in the theory of symmetry restauration are briefly outlined [fr

  12. Research progress on dam-break floods

    KAUST Repository

    Wu, Jiansong

    2011-08-01

    Because of the catastrophic effects downstream of dam-break failure, more and more researchers around the world have been working on the study of dam-break flows to accurately forecast the downstream inundation mapping. With the rapid development of computer hardware and computing techniques, numerical study on dam-break flows has been a popular research subject. In the paper, the numerical methodologies used to solve the governing partial differential equations of dam-break flows are classified and summarized, and their characteristics and applications are discussed respectively. Furthermore, the fully-developed mathematical models developed in recent decades are reviewed, and also introduced the authors\\' on-going work. Finally, some possible future developments on modeling the dam-break flows and some solutions are presented and discussed. © 2011 IEEE.

  13. Supersymmetry and electroweak breaking in the interval

    International Nuclear Information System (INIS)

    Diego, David; Gersdorff, Gero von; Quiros, Mariano

    2005-01-01

    Hypermultiplets are considered in the five-dimensional interval where all fields are continuous and the boundary conditions are dynamically obtained from the action principle. The orbifold boundary conditions are obtained as particular cases. We can interpret the Scherk-Schwarz supersymmetry breaking as a misalignment of boundary conditions while a new source of supersymmetry breaking corresponding to a mismatch of different boundary parameters is identified. The latter can be viewed as coming from boundary supersymmetry breaking masses for hyperscalars and the nature of the corresponding supersymmetry breaking parameter is analyzed. For some regions of the parameter space where supersymmetry is broken (either by Scherk-Schwarz boundary conditions or by boundary hyperscalar masses) electroweak symmetry breaking can be triggered at the tree level

  14. 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

  15. To count or not to count: the effect of instructions on expecting a break in timing.

    Science.gov (United States)

    Gaudreault, Rémi; Fortin, Claudette

    2013-04-01

    When a break is expected during a time interval production, longer intervals are produced as the break occurs later during the interval. This effect of break location was interpreted as a result of distraction related to break expectancy in previous studies. In the present study, the influence of target duration and of instructions about chronometric counting strategies on the break location effect was examined. Using a strategy such as chronometric counting enhances the reliability of temporal processing, typically in terms of reduced variability, and could influence how timing is affected by break expectancy, especially when relatively long target durations are used. In two experiments, results show that time productions lengthened with increasing value of break location at various target durations and that variability was greater in the no-counting than in the counting instruction condition. More important, the break location effect was stronger in the no-counting than in the counting instruction condition. We conclude that chronometric counting orients attention toward timing processes, making them less likely to be disrupted by concurrent nontemporal processes.

  16. 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

    A few years ago, within the framework of the CEA/ EDF /Framatome ANP R and D cooperative program, we made the assumption that the burn-up influence on the thermal-mechanical behavior of the cladding tubes under LOCA conditions is strongly linked to the hydrogen uptake due to the in-service oxidation. Thus, since that time, an extensive experimental program has been conducted in CEA labs on as-received and pre-hydrided Zircaloy-4 and M5 TM alloys of Framatome-ANP to get a better insight into the influence of the hydrogen on the thermal-mechanical cladding behavior during the first phase of the LOCA transient (ballooning and rupture) and for post-quenched conditions (residual ductility / toughness). On the one hand, one of the main assumptions here was that the microstructural defects, and the resultant hardening produced under heavy neutron irradiation within the Zr matrix, are annealed early in the first phase of the LOCA transient (i.e. first thermal ramp) and thus, that the main effects of high burn-up should come from the hydrogen uptake. To assess this hypothesis, specific thermal-mechanical tests have been performed on as-received, pre-hydrided and irradiated cladding tubes. This confirmed that the effect of hydrogen uptake dominates over that of irradiation on the thermal-mechanical response of the materials. So, in a first part of the paper, we will summarize the main results obtained here and, from the metallurgical point of view, we will illustrate the strong influence of hydrogen on the decrease of the α-to-β phase transformation temperatures of the zirconium alloys studied. On the other hand, studies have been performed on the post-quench mechanical behavior of as-received and pre-hydrided cladding tubes after single-face oxidation at 1000-1200 degree C and quenching. In parallel with these mechanical tests, in-depth metallurgical investigations have been developed, to be able to quantify the resultant phases thickness (ZrO 2 , α-Zr(O) and ex-β phase

  17. Poethical: Breaking Ground for Reconstruction

    Science.gov (United States)

    Krojer, Jo; Holge-Hazelton, Bibi

    2008-01-01

    Departing from a methodological experiment performed by the authors, this article reflects on and discusses issues of ethics and politics in poetic strategies of "representation". In relation to the experiment the article questions how to conceive the notion of connectedness between empirical time and the reconstruction of it in poststructuralist…

  18. Breaking the prejudice habit: Mechanisms, timecourse, and longevity.

    Science.gov (United States)

    Forscher, Patrick S; Mitamura, Chelsea; Dix, Emily L; Cox, William T L; Devine, Patricia G

    2017-09-01

    The prejudice habit-breaking intervention (Devine et al., 2012) and its offshoots (e.g., Carnes et al., 2012) have shown promise in effecting long-term change in key outcomes related to intergroup bias, including increases in awareness, concern about discrimination, and, in one study, long-term decreases in implicit bias. This intervention is based on the premise that unintentional bias is like a habit that can be broken with sufficient motivation, awareness, and effort. We conducted replication of the original habit-breaking intervention experiment in a sample more than three times the size of the original ( N = 292). We also measured all outcomes every other day for 14 days and measured potential mechanisms for the intervention's effects. Consistent with previous results, the habit-breaking intervention produced a change in concern that endured two weeks post-intervention. These effects were associated with increased sensitivity to the biases of others and an increased tendency to label biases as wrong. Contrasting with the original work, both control and intervention participants decreased in implicit bias, and the effects of the habit-breaking intervention on awareness declined in the second week of the study. In a subsample recruited two years later, intervention participants were more likely than control participants to object on a public online forum to an essay endorsing racial stereotyping. Our results suggest that the habit-breaking intervention produces enduring changes in peoples' knowledge of and beliefs about race-related issues, and we argue that these changes are even more important for promoting long-term behavioral change than are changes in implicit bias.

  19. Improving long term driving comfort by taking breaks - How break activity affects effectiveness.

    Science.gov (United States)

    Sammonds, George M; Mansfield, Neil J; Fray, Mike

    2017-11-01

    During long duration journeys, drivers are encouraged to take regular breaks. The benefits of breaks have been documented for safety; breaks may also be beneficial for comfort. The activity undertaken during a break may influence its effectiveness. Volunteers completed 3 journeys on a driving simulator. Each 130 min journey included a 10 min break after the first hour. During the break volunteers either stayed seated, left the simulator and sat in an adjacent room, or took a walk on a treadmill. The results show a reduction in driver discomfort during the break for all 3 conditions, but the effectiveness of the break was dependent on activity undertaken. Remaining seated in the vehicle provided some improvement in comfort, but more was experienced after leaving the simulator and sitting in an adjacent room. The most effective break occurred when the driver walked for 10 min on a treadmill. The benefits from taking a break continued until the end of the study (after a further hour of driving), such that comfort remained the best after taking a walk and worst for those who remained seated. It is concluded that taking a break and taking a walk is an effective method for relieving driving discomfort. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Breaking-bud pollination: a new pollination process in partially opened flowers by small bees.

    Science.gov (United States)

    Yamaji, Futa; Ohsawa, Takeshi A

    2015-09-01

    Plant-pollinator interactions have usually been researched in flowers that have fully opened. However, some pollinators can visit flowers before full opening and contribute to fruit and seed sets. In this paper, we researched the pollination biology of flowers just starting to open in four field experiments. We observed the insect visitors to Lycoris sanguinea var. sanguinea for 3 years at five sites. These observations revealed that only small bees, Lasioglossum japonicum, often entered through tiny spaces between the tepals of 'breaking buds' (i.e. partially opened flowers) and collected pollen. We hypothesized that they can pollinate this species at the breaking-bud stage, when the stigma is located near the anthers. To measure the pollination effect of small bees at the breaking-bud stage, we bagged several breaking buds after small bees had visited them and examined whether these buds were pollinated. In bagging experiments, 30% of the breaking buds set fruit and seeds. Fruit-set ratios of the breaking buds did not differ significantly from those of the fully opened flowers, which had been visited by several insect species. We also counted the pollen grain numbers on the body of L. japonicum and on the anthers of randomly-selected and manipulated flowers. These experiments revealed that all of the captured bees had some pollen of target plants and that L. japonicum collected most of the pollen grains at the breaking-bud stage. Our results showed that the new pollination process, breaking-bud pollination, happened in breaking buds by L. japonicum, although there is no evidence to reveal that this is the most effective pollination method for L. sanguinea var. sanguinea. In principle, this new pollination process can occur in other flowering plants and our results are a major contribution to studies of plant-pollinator interactions.

  1. Assessment of passive safety system performance under main steam line break accident

    International Nuclear Information System (INIS)

    Lim, J.; Choi, S.W.; Yang, J.; Lee, D.Y.; Rassame, S.; Hibiki, T.; Ishii, M.

    2014-01-01

    Highlights: • An integral test of MSLB was conducted in a small-scale test facility, i.e. PUMA which is scaled to generation III + BWR applications. • GDCS as an ECCS provided adequate supply of water to keep RPV coolant level well above TAF. • PCCS kept containment below design pressure during long-term cooling phase. - Abstract: A generation III + Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features which require no emergency injection pump and no operator action or Alternating Current (AC) power supply. The generation III + BWR’s passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS), and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A Main Steam Line Break (MSLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III + BWR. The main results of PUMA MSLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the minimum water level (1.706 m) was 5% higher than the TAF (1.623 m) and the containment maximum pressure (271 kPa) was 35% lower than the safety limit (414 kPa), respectively

  2. 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)

  3. 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)

  4. Design of experiments and equipment to test the ballooning characteristics of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Forrest, C.F.; Stern, F.; Hart, R.G.

    1992-01-01

    Experiments have been planned and an apparatus has been designed to enable creep testing of end-of-life pressure tube specimens in a LOCA environment. Effects that could be studied include: annealing of irradiation damage during transient heating; effects of hydride blisters on pressure tube ballooning strains; and, effects of uniformly-distributed hydrogen content on pressure tube ballooning strains. The proposed experimental program will consist of separate effects creep tests on pressure tube sections under transient heating conditions

  5. MM98.19 An automatic system for elaboration of chip breaking diagrams

    DEFF Research Database (Denmark)

    Andreasen, Jan Lasson; Chiffre, Leonardo De

    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...

  6. 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.

  7. 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

  8. Particle production from symmetry breaking after inflation

    CERN Document Server

    García-Bellido, J; Garcia-Bellido, Juan; Morales, Ester Ruiz

    2002-01-01

    Recent studies suggest that the process of symmetry breaking after inflation typically occurs very fast, within a single oscillation of the symmetry-breaking field, due to the spinodal growth of its long-wave modes, otherwise known as `tachyonic preheating'. In this letter we show how this sudden transition from the false to the true vacuum can induce a significant production of particles, bosons and fermions, coupled to the symmetry-breaking field. We find that this new mechanism of particle production in the early Universe may have interesting consequences for the origin of dark matter and the generation of the observed baryon asymmetry through leptogenesis.

  9. 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

  10. Breaking object correspondence across saccadic eye movements deteriorates object recognition

    Directory of Open Access Journals (Sweden)

    Christian H. Poth

    2015-12-01

    Full Text Available Visual perception is based on information processing during periods of eye fixations that are interrupted by fast saccadic eye movements. The ability to sample and relate information on task-relevant objects across fixations implies that correspondence between presaccadic and postsaccadic objects is established. Postsaccadic object information usually updates and overwrites information on the corresponding presaccadic object. The presaccadic object representation is then lost. In contrast, the presaccadic object is conserved when object correspondence is broken. This helps transsaccadic memory but it may impose attentional costs on object recognition. Therefore, we investigated how breaking object correspondence across the saccade affects postsaccadic object recognition. In Experiment 1, object correspondence was broken by a brief postsaccadic blank screen. Observers made a saccade to a peripheral object which was displaced during the saccade. This object reappeared either immediately after the saccade or after the blank screen. Within the postsaccadic object, a letter was briefly presented (terminated by a mask. Observers reported displacement direction and letter identity in different blocks. Breaking object correspondence by blanking improved displacement identification but deteriorated postsaccadic letter recognition. In Experiment 2, object correspondence was broken by changing the object’s contrast-polarity. There were no object displacements and observers only reported letter identity. Again, breaking object correspondence deteriorated postsaccadic letter recognition. These findings identify transsaccadic object correspondence as a key determinant of object recognition across the saccade. This is in line with the recent hypothesis that breaking object correspondence results in separate representations of presaccadic and postsaccadic objects which then compete for limited attentional processing resources (Schneider, 2013. Postsaccadic

  11. Influence of processes of nonlinear transformations of waves in the coastal zone on the height of breaking waves

    Science.gov (United States)

    Saprykina, Ya. V.; Kuznetsov, S. Yu.; Divinskii, B. V.

    2017-05-01

    Using data from laboratory, field, and numerical experiments, we investigated regularities in changes in the relative limit height of breaking waves (the breaking index) from peculiarities of nonlinear wave transformations and type of wave breaking. It is shown that the value of the breaking index depends on the relative part of the wave energy in the frequency range of the second nonlinear harmonic. If this part is more than 35%, then the breaking index can be taken as a constant equal to 0.6. These waves are spilling breaking waves, asymmetric on the horizontal axis, and are almost symmetric on the vertical axis. If this part of the energy is less than 35%, then the breaking index increases with increasing energy in the frequency range of the second harmonic. These waves are plunging breaking waves, asymmetric on the vertical axis, and are almost symmetric on the horizontal axis. It is revealed that the breaking index depends on the asymmetry of waves on the vertical axis, determined by the phase shift between the first and second nonlinear harmonic (biphase). It is shown that the relation between the amplitudes of the second and first nonlinear harmonics for an Ursell number less than 1 corresponds to Stokes' second-order wave theory. The empirical dependences of the breaking index on the parameters of nonlinear transformation of waves are proposed.

  12. Unconventional supersymmetry and its breaking

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, Pedro D., E-mail: alvarez@physics.ox.ac.uk [Centro de Estudios Científicos (CECS), Av. Arturo Prat 514, Valdivia (Chile); Universidad Andrés Bello, Av. República 440, Santiago (Chile); Rudolf Peierls Centre for Theoretical Physics, University of Oxford, Oxford (United Kingdom); Pais, Pablo, E-mail: pais@cecs.cl [Centro de Estudios Científicos (CECS), Av. Arturo Prat 514, Valdivia (Chile); Universidad Andrés Bello, Av. República 440, Santiago (Chile); Zanelli, Jorge, E-mail: z@cecs.cl [Centro de Estudios Científicos (CECS), Av. Arturo Prat 514, Valdivia (Chile); Universidad Andrés Bello, Av. República 440, Santiago (Chile)

    2014-07-30

    We present a gauge theory for a superalgebra that includes an internal gauge (G) and local Lorentz (so(1,D−1)) algebras. These two symmetries are connected by fermionic supercharges. The field content of the system includes a (non-)abelian gauge potential A, a spin-1/2 Dirac spinor ψ, the Lorentz connection ω{sup ab}, and the vielbein e{sub μ}{sup a}. The connection one-form A is in the adjoint representation of G, while ψ is in the fundamental. In contrast to standard supersymmetry and supergravity, the metric is not a fundamental field and is in the center of the superalgebra: it is not only invariant under the internal gauge group, G, and under Lorentz transformations, SO(1,D−1), but is also invariant under supersymmetry. The distinctive features of this theory that mark the difference with standard supersymmetries are: i) the number of fermionic and bosonic states is not necessarily the same; ii) there are no superpartners with equal mass; iii) although this supersymmetry originates in a local gauge theory and gravity is included, there is no gravitino; iv) fermions acquire mass from their coupling to the background or from higher order self-couplings, while bosons remain massless. In odd dimensions, the Chern–Simons (CS) form provides an action that is (quasi-)invariant under the entire superalgebra. In even dimensions, the Yang–Mills (YM) form is the only natural option and the symmetry breaks down to G⊗SO(1,D−1). In four dimensions, the construction follows the Townsend–Mac Dowell–Mansouri approach, starting with an osp(4|2)∼usp(2,2|1) connection. Due to the absence of osp(4|2)-invariant traces in four dimensions, the resulting Lagrangian is only invariant under u(1)⊕so(3,1), which includes a Nambu–Jona-Lasinio (NJL) term. In this case, the Lagrangian depends on a single dimensionful parameter that fixes Newton's constant, the cosmological constant and the NJL coupling.

  13. LOFT/LP-FP-1B, Loss of Fluid Test, Fission Product Release 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 seventh OECD LOFT experiment was conducted on 19 December 1984. It was the first of the two experiments to be performed in the LOFT facility with intentional release of fission products. Its objectives were to obtain data on fission product release from the fuel-cladding gap into vapor and reflood water and to collect data on transport of these fission products through and out of the reactor coolant system. The experiment was initiated by a reactor scram with one second delayed opening of the quick-opening blowdown valves. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  14. Tidal Mixing at the Shelf Break

    National Research Council Canada - National Science Library

    Hogg, Nelson; Legg, Sonya

    2005-01-01

    The aim of this project was to study mixing forced by tidal flow over sudden changes in topographic slope such as near the shelf-break, using high-resolution nonhydrostatic numerical simulations employing the MIT gem...

  15. Water Breaking: Understand This Sign of Labor

    Science.gov (United States)

    Healthy Lifestyle Labor and delivery, postpartum care Water breaking worries? Prepare yourself for childbirth by getting the facts about this important sign of labor. By Mayo Clinic Staff If you're ...

  16. Higgsless grand unified theory breaking and trinification

    International Nuclear Information System (INIS)

    Carone, Christopher D.; Conroy, Justin M.

    2004-01-01

    Boundary conditions on an extra dimensional interval can be chosen to break bulk gauge symmetries and to reduce the rank of the gauge group. We consider this mechanism in models with gauge trinification. We determine the boundary conditions necessary to break the trinified gauge group directly down to that of the standard model. Working in an effective theory for the gauge-symmetry-breaking parameters on a boundary, we examine the limit in which the grand-unified theory-breaking-sector is Higgsless and show how one may obtain the low-energy particle content of the minimal supersymmetric standard model. We find that gauge unification is preserved in this scenario, and that the differential gauge coupling running is logarithmic above the scale of compactification. We compare the phenomenology of our model to that of four dimensional 'trinified' theories

  17. Proposed changes in intermediate pipe break criteria

    International Nuclear Information System (INIS)

    Schmitz, R.P.

    1984-01-01

    Bechtel Power Corporation proposed to the US NRC in 1983 that the NRC eliminate from their criteria all intermediate breaks. Bechtel's rationale for the proposal and support for their position are presented

  18. The problem of symmetry breaking hierarchy

    International Nuclear Information System (INIS)

    Natale, A.A.

    1983-01-01

    The problem of symmetry breaking hierarchy in grand unified theories is discussed, proving the impossibility to get a big hierarchy of interactions, in a natural way within the framework of perturbation theory. (L.C.) [pt

  19. Phil Anderson and Gauge Symmetry Breaking

    Science.gov (United States)

    Witten, Edward

    In this article, I describe the celebrated paper that Phil Anderson wrote in 1962 with early contributions to the idea of gauge symmetry breaking in particle physics. To set the stage, I describe the work of Julian Schwinger to which Anderson was responding, and also some of Anderson's own work on superconductivity that provided part of the context. After describing Anderson's work I describe the later work of others, leading to the modern understanding of gauge symmetry breaking in weak interactions...

  20. Hydraulic Response of Caisson Breakwaters in Multidirectional Breaking and Non-Breaking Waves

    DEFF Research Database (Denmark)

    Grønbech, J.; Kofoed, Jens Peter; Hald, Tue

    1998-01-01

    The present paper concerns the results and findings of a physical study on wave impacts on vertical caisson breakwaters situated in irregular, multidirectional breaking seas. The study has taken place as part of the framework programme "Dynamic of Structures" financially supported by the Danish...... induced loading and overtopping on caisson breakwaters situated in breaking seas. Regarding the wave forces only minor differences between breaking and non breaking waves in deep water were observed, and it was found that the prediction formula of Goda also seems to apply well for multidirectionally...