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Sample records for simulated bwr environments

  1. Effect of nitrogen in austenitic stainless steel on deformation behavior and stress corrosion cracking susceptibility in BWR simulated environment

    International Nuclear Information System (INIS)

    Roychowdhury, S.; Kain, V.; Dey, G.K.

    2012-01-01

    Intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) components in boiling water reactor (BWR has been a serious issue and is generic in nature. Initial cracking incidences were attributed to weld induced sensitisation and low temperature sensitisation which was mitigated by the use of low carbon grade of SS and molybdenum and nitrogen containing nuclear grade SS. However, IGSCC has occurred in these SS in the non-sensitised condition which was attributed to residual weld induced strain. Strain hardening in SS has been identified as a major cause for enhanced IGSCC susceptibility in BWR environment. Nitrogen in SS has a significant effect on the strain hardening characteristics and has potential to affect the IGSCC susceptibility in BWR environment. Type 304LN stainless steel is a candidate material for use in future reactors with long design life like the Advanced Heavy Water Reactor (AHWR), in which the operating conditions are similar to BWR. This study reports the effect of nitrogen in type 304LN stainless steel on the strain hardening behaviour and deformation characteristics and its effect on the IGSCC susceptibility in BWR/AHWR environment. Two heats of type 304LN stainless steel were used containing different levels of nitrogen, 0.08 and 0.16 wt % (SS alloys A and B, respectively). Both the SS was strain hardened by cross rolling at 200℃ to simulate the strain hardened regions having higher IGSCC susceptibility in BWRs. Tensile testing was done at both room temperature and 288℃(temperature simulating operating BWR conditions) and the effect of nitrogen on the tensile properties were established. Tensile testing was done at strain rates similar to the crack tip strain rates associated with a growing IGSCC in SS. Detailed transmission electron microscopic (TEM) studies were done to establish the effect of nitrogen on the deformation modes. Results indicated twinning was the major mode of deformation during cross rolling while

  2. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

  3. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  4. BWR-plant simulator and its neural network companion with programming under mat lab environment

    International Nuclear Information System (INIS)

    Ghenniwa, Fatma Suleiman

    2008-01-01

    Stand alone nuclear power plant simulators, as well as building blocks based nuclear power simulator are available from different companies throughout the world. In this work, a review of such simulators has been explored for both types. Also a survey of the possible authoring tools for such simulators development has been performed. It is decided, in this research, to develop prototype simulator based on components building blocks. Further more, the authoring tool (Mat lab software) has been selected for programming. It has all the basic tools required for the simulator development similar to that developed by specialized companies for simulator like MMS, APROS and others. Components simulations, as well as integrated components for power plant simulation have been demonstrated. Preliminary neural network reactor model as part of a prepared neural network modules library has been used to demonstrate module order shuffling during simulation. The developed components library can be refined and extended for further development. (author)

  5. Corrosion of pre-oxidized nickel alloy X-750 in simulated BWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Tuzi, Silvia, E-mail: silvia.tuzi@chalmers.se [Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Lai, Haiping [Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Göransson, Kenneth [Westinghouse Electric Sweden AB, SE-721 63 Västerås (Sweden); Thuvander, Mattias; Stiller, Krystyna [Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

    2017-04-01

    Samples of pre-oxidized Alloy X-750 were exposed to a simulated boiling water reactor environment in an autoclave at a temperature of 286 °C and a pressure of 80 bar for four weeks. The effect of alloy iron content on corrosion was investigated by comparing samples with 5 and 8 wt% Fe, respectively. In addition, the effect of two different surface pre-treatments was investigated. The microstructure of the formed oxide scales was studied using mainly electron microscopy. The results showed positive effects of an increased Fe content and of removing the deformed surface layer by pickling. After four weeks of exposure the oxide scale consists of oxides formed in three different ways. The oxide formed during pre-oxidization at 700 °C, mainly consisting of chromia, is partly still present. There is also an outer oxide consisting of NiFe{sub 2}O{sub 4} crystals, reaching a maximum size of 3 μm, which has formed by precipitation of dissolved metal ions. Finally, there is an inner nanocrystalline and porous oxide, with a metallic content reflecting the alloy composition, which has formed by corrosion.

  6. TEM/STEM study of Zircaloy-2 with protective FeAl(Cr) layers under simulated BWR environment and high-temperature steam exposure

    Science.gov (United States)

    Park, Donghee; Mouche, Peter A.; Zhong, Weicheng; Mandapaka, Kiran K.; Was, Gary S.; Heuser, Brent J.

    2018-04-01

    FeAl(Cr) thin-film depositions on Zircaloy-2 were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) with respect to oxidation behavior under simulated boiling water reactor (BWR) conditions and high-temperature steam. Columnar grains of FeAl with Cr in solid solution were formed on Zircaloy-2 coupons using magnetron sputtering. NiFe2O4 precipitates on the surface of the FeAl(Cr) coatings were observed after the sample was exposed to the simulated BWR environment. High-temperature steam exposure resulted in grain growth and consumption of the FeAl(Cr) layer, but no delamination at the interface. Outward Al diffusion from the FeAl(Cr) layer occurred during high-temperature steam exposure (700 °C for 3.6 h) to form a 100-nm-thick alumina oxide layer, which was effective in mitigating oxidation of the Zircaloy-2 coupons. Zr intermetallic precipitates formed near the FeAl(Cr) layer due to the inward diffusion of Fe and Al. The counterflow of vacancies in response to the Al and Fe diffusion led to porosity within the FeAl(Cr) layer.

  7. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  8. BWR nuclear plant maintenance simulation

    International Nuclear Information System (INIS)

    Stuart, I.F.

    1985-01-01

    As early as 1977, the General Electric Company, USA, Nuclear Energy Operation was making plans to construct a maintenance-type simulator to support Training and Services. The Company's pioneering experience with control room simulators started in 1968 with the Dresden simulator and showed clearly the benefits of having such facilities for training, checkout of procedures and, in the case of maintenance, match-up of equipment or tools as needed. Since the dedication of the facility, it has proved to be an invaluable resource in the training of refuelling and servicing crews. The facility has also been extensively used as developmental and test facility for in-vessel servicing equipment and procedures. (author)

  9. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  10. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  11. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  12. Effects of Cr and Nb contents on the susceptibility of Alloy 600 type Ni-base alloys to stress-corrosion cracking in a simulated BWR environment

    International Nuclear Information System (INIS)

    Akashi, Masatsune

    1995-01-01

    In order to discuss the effects of chromium and niobium contents on the susceptibility of Alloy 600 type nickel-base alloys to stress-corrosion cracking in the BWR primary coolant environment, a series of creviced bent-beam (CBB) tests were conducted in a high-temperature, high-purity water environment. Chromium, niobium, and titanium as alloying elements improved the resistivity to stress-corrosion cracking, whereas carbon enhanced the susceptibility to it. Alloy-chemistry-based correlations have been defined to predict the relative resistances of alloys to stress-corrosion cracking. A strong correlation was found, for several heats of alloys, between grain-boundary chromium depletion and the susceptibility to stress-corrosion cracking

  13. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  14. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

  15. Studies of Corrosion of Cladding Materials in Simulated BWR-environment Using Impedance Measurements. Part I: Measurements in the Pre-transition Region

    International Nuclear Information System (INIS)

    Forsberg, Stefan; Ahlberg, Elisabet; Andersson, Ulf

    2004-09-01

    The corrosion of three Zircaloy 2 cladding materials, LK2, LK2+ and LK3, have been studied in-situ in an autoclave using electrochemical impedance spectroscopy. Measurements were performed in simulated BWR water at temperatures up to 288 deg C. The impedance spectra were successfully modelled using equivalent circuits. When the oxide grew thicker during the experiments, a change-over from one to two time constants was seen, showing that a layered structure was formed. Oxide thickness, oxide conductivity and effective donor density were evaluated from the impedance data. The calculated oxide thickness at the end of the experiments was consistent with the value obtained from SEM. It was shown that the difference in oxide growth rate between the investigated materials is small in the pre-transition region. The effective donor density, which is a measure of electronic conductivity, was found to be lower for the LK3 material compared to the other two materials

  16. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

    Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

    1998-02-01

    The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

  17. Sophistication of operator training using BWR plant simulator

    International Nuclear Information System (INIS)

    Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

    1986-01-01

    In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

  18. IGSCC in cold worked austenitic stainless steel in BWR environment

    International Nuclear Information System (INIS)

    Persson, B.; Lindblad, B.

    1989-09-01

    The survey shows that austenitic stainless steels in a cold worked condition can exhibit IGSCC in BWR environment. It is also found that IGSCC often is initiated as a transgranular crack. Local stresses and surface defects very often acts as starting points for IGSCC. IGSCC due to cold working requires a cold working magnitude of at leas 5%. During cold working a formation of mechanical martensite can take place. The transgranular corrosion occurs in the martensitic phase due to sensitation. The crack propagates integranularly due to anodic solvation of α'-martensite. Sensitation of the martensitic phase is fasten in BCC-structures than in a FCC-structures mainly due to faster diffusion of chromium and carbon which cause precipitation of chromium carbides. Experiments show that a carbon content as low as 0.008% is enough for the formation of 68% martensite and for sensitation. Hydrogen induced cracking is regarded as a mechanism which can accelerate IGSCC. Such cracking requires a hydrostatic stress near the crack tip. Since the oxide in the crack tip is relatively impermeable to hydrogen, cracks in the oxide layer are required for such embrittlement. Hydrogen induced embrittlement of the martensitic phase, at the crack tip, can cause crack propagation. Solution heat treated unstabilized stainless steels are regarded to have a good resistance to IGSCC if they have not undergone cold working. In general, though, Mo-alloyed steels have a better resistance to IGSCC in BWR environment. Regarding the causes for IGSCC, the present literature survey shows that many mechanisms are suggested. To provide a safer ground for the estimation of crack propagation rates, SA recommends SKI to finance a project with the aim to determine the crack propagation rate on proper material. (authors) (65 refs.)

  19. Modeling of SCC initiation and propagation mechanisms in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmeister, Hans, E-mail: Hans.Hoffmeister@hsu-hh.de [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany); Klein, Oliver [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer We show that SSC in BWR environments includes anodic crack propagation and hydrogen assisted cracking. Black-Right-Pointing-Pointer Hydrogen cracking is triggered by crack tip acidification following local impurity accumulations and subsequent phase precipitations. Black-Right-Pointing-Pointer We calculate effects of pH, chlorides, potentials and stress on crack SCC growth rates at 288 Degree-Sign C. - Abstract: During operation of mainly BWRs' (Boiling Water Reactors) excursions from recommended water chemistries may provide favorite conditions for stress corrosion cracking (SCC). Maximum levels for chloride and sulfate ion contents for avoiding local corrosion are therefore given in respective water specifications. In a previously published deterministic 288 Degree-Sign C - corrosion model for Nickel as a main alloying element of BWR components it was demonstrated that, as a theoretically worst case, bulk water chloride levels as low as 30 ppb provide local chloride ion accumulation, dissolution of passivating nickel oxide and precipitation of nickel chlorides followed by subsequent local acidification. In an extension of the above model to SCC the following work shows that, in a first step, local anodic path corrosion with subsequent oxide breakdown, chloride salt formation and acidification at 288 Degree-Sign C would establish local cathodic reduction of accumulated hydrogen ions inside the crack tip fluid. In a second step, local hydrogen reduction charges and increasing local crack tip strains from increasing crack lengths at given global stresses are time stepwise calculated and related to experimentally determined crack critical cathodic hydrogen charges and fracture strains taken from small scale SSRT tensile tests pieces. As a result, at local hydrogen equilibrium potentials higher than those of nickel in the crack tip solution, hydrogen ion reduction initiates hydrogen crack propagation that is enhanced with

  20. BWR Full Integral Simulation Test (FIST). Phase I test results

    International Nuclear Information System (INIS)

    Hwang, W.S.; Alamgir, M.; Sutherland, W.A.

    1984-09-01

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report

  1. Cobra-TF simulation of BWR bundle dry out experiments

    Energy Technology Data Exchange (ETDEWEB)

    Frepoli, C.; Ireland, A.; Hochreiter, L.; Ivanov, K. [Penn State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, PA (United States); Velten, R. [Siemens Nuclear Power GmbH, Erlangen (Germany)

    2001-07-01

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

  2. Upgrading BWR training simulators for annual outage operation training

    International Nuclear Information System (INIS)

    Yamakabe, K.; Nakajima, A.; Shiyama, H.; Noji, K.; Okabe, N.; Murata, F.

    2006-01-01

    Based upon the recently developed quality assurance program by the Japanese electric companies, BWR Operator Training Center (BTC) identified the needs to enhance operators' knowledge and skills for operations tasks during annual outage, and started to develop a dedicated operator training course specialized for them. In this paper, we present the total framework of the training course for annual outage operations and the associated typical three functions of our full-scope simulators specially developed and upgraded to conduct the training; namely, (1) Simulation model upgrade for the flow and temperature behavior concerning residual heat removal (RHR) system with shutdown cooling mode, (2) Addition of malfunctions for DC power supply equipment, (3) Simulation model upgrade for water filling operation for reactor pressurization (future development). We have implemented a trial of the training course by using the upgraded 800MW full-scope training simulator with functions (1) and (2) above. As the result of this trial, we are confident that the developed training course is effective for enhancing operators' knowledge and skills for operations tasks during annual outage. (author)

  3. General model for Pc-based simulation of PWR and BWR plant components

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

    1995-10-01

    In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

  4. Numerical simulation of boron injection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

    2010-02-15

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

  5. Numerical simulation of boron injection in a BWR

    International Nuclear Information System (INIS)

    Tinoco, Hernan; Buchwald, Przemyslaw; Frid, Wiktor

    2010-01-01

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several

  6. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

  7. An A BWR demonstration simulator for training and developing technical staff

    International Nuclear Information System (INIS)

    Powers, J.; Yonezawa, H.; Aoyagi, Y.; Kataoka, K.

    2015-09-01

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  8. An A BWR demonstration simulator for training and developing technical staff

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  9. TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

  10. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    International Nuclear Information System (INIS)

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-01-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The

  11. On the fast estimation of transit times application to BWR simulated data

    International Nuclear Information System (INIS)

    Antonopoulos-Domis, M.; Marseguerra, M.; Padovani, E.

    1996-01-01

    Real time estimators of transit times are proposed. BWR noise is simulated including a global component due to rod vibration. The time obtained form the simulation is used to investigate the robustness and noise immunity of the estimators. It is found that, in presence of a coincident (global) signal, the cross-correlation function is the worst estimator. (authors)

  12. BWR full integral simulation test (FIST) pretest predictions with TRACBO2

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.

    1984-01-01

    The Full Integral Simulation Test program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An analytical method development program is underway to extend the BWR-TRAC computer code to model reactor kinetics and major interfacing systems, including balance-of-plant, to improve application modeling flexibility, and to reduce computer running time. An experimental program is underway in a new single bundle system test facility to extend the large break loss-of-coolant accident LOCA data base to small breaks and operational transients. And a method qualification program is underway to test TRACBO2 against experiments in the FIST facility. The recently completed Phase 1 period included a series of LOCA and power transient tests, and successful pretest analysis of the large and small break LOCA tests with TRACBO2. These comparisons demonstrate BWR-TRAC capability for small and large break analysis, and provide detailed understanding of the phenomena

  13. Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly

    International Nuclear Information System (INIS)

    Mitsutake, T.; Terasaka, H.; Yoshimura, K.; Oishi, M.; Inoue, A.; Akiyama, M.

    1990-01-01

    Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8x8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F 'multi-fluid' computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, though comparisons, between the prediction and the obtained test data. (orig.)

  14. Effect of yield strength on stress corrosion crack propagation under PWR and BWR environments of hardened stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Castano, M.L.; Garcia, M.S.; Diego, G. de; Gomez-Briceno, D. [CIEMAT, Nuclear Fission Department, Structural Materials Program, Avda. Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    Core components of light water reactor (LWR), mainly made of austenitic stainless steels (SS), subjected to stress and exposed to relatively high fast neutron flux may suffer a cracking process termed as Irradiation Assisted Stress Corrosion Cracking (IASCC). Neutron radiation leads to critical modifications in material characteristics, which can modify their stress corrosion cracking (SCC) response. Current knowledge highlights three fundamental factors, induced by radiation, as primary contributors to IASCC of core materials: Radiation Induced Segregation (RIS) at grain boundaries, Radiation Hardening and Radiolysis. Most of the existing literature on IASCC is focussed on the influence of RIS, mainly chromium depletion, which can promote IASCC in oxidizing environments, such a Boiling Water Reactor (BWR) under normal water chemistry. However, in non-oxidizing environments, such as primary water of Pressurized Water Reactor (PWR) or BWR hydrogen water chemistry, the role played by chromium depletion at grain boundary on IASCC behaviour of highly irradiated material is irrelevant. One important issue with limited study is the effect of radiation induced hardening. The role of hardening on IASCC is became stronger considered, especially for environments where other factors, like micro-chemistry, have no significant influence. To formulate the mechanism of IASCC, a well-established method is to isolate and quantify the effect of individual parameters. The use of unirradiated material and the simulation of the irradiation effects is a procedure used with success for evaluating the influence of irradiation effects. Radiation hardening can be simulated by mechanical deformation and, although some differences exist in the types of defects produced, it is believed that the study of the SCC behaviour of unirradiated materials with different hardening levels would contribute to the understanding of IASCC mechanism. In order to evaluate the influence of hardening on the

  15. Efficient method for simulation of BWR severe accident sequence events before core uncovery

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1984-01-01

    BWR-LACP has been a versatile tool for the ORNL SASA program. The development effort was minimal, and the code is fast running and economical. Operator actions are easily simulated and the complete scope of both reactor vessel and primary containment are modeled. Valuable insights have been gained into accident sequences. A Fortran version is under development and it will be modified for application to Mark II plants

  16. Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR

    International Nuclear Information System (INIS)

    Arganis J, C. R.

    2010-01-01

    The stress corrosion cracking susceptibility was studied in sensitized, solution annealed 304 steel, and in 304-L welded with a heat treatment that simulated the radiation induced segregation, by the slow strain rate test technique, in a similar environment of a boiling water reactor (BWR), 288 C, 8 MPa, low conductivity and a electrochemical corrosion potential near 200 mV. vs. standard hydrogen electrode (She). The electrochemical noise technique was used for the detection of the initiation and propagation of the cracking. The steels were characterized by metallographic studies with optical and scanning electronic microscopy and by the electrochemical potentiodynamic reactivation of single loop and double loop. In all the cases, the steels present delta ferrite. The slow strain rate tests showed that the 304 steel in the solution annealed condition is susceptible to transgranular stress corrosion cracking (TGSCC), such as in a normalized condition showed granulated. In the sensitized condition the steel showed intergranular stress corrosion cracking, followed by a transition to TGSCC. The electrochemical noise time series showed that is possible associated different time sequences to different modes of cracking and that is possible detect sequentially cracking events, it is means, one after other, supported by the fractographic studies by scanning electron microscopy. The parameter that can distinguish between the different modes of cracking is the re passivation rate, obtained by the current decay rate -n- in the current transients. This is due that the re passivation rate is a function of the microstructure and the sensitization. Other statistic parameters like the localized index, Kurtosis, Skew, produce results that are related with mixed corrosion. (Author)

  17. Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

    Energy Technology Data Exchange (ETDEWEB)

    Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

    1976-07-01

    Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

  18. Parametric tests of the effects of water chemistry impurities on corrosion of Zr-alloys under simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, S; Ito, K [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Lin, C C [GE Nucklear Energy (United States); Cheng, B [Electric Power Research Inst. (United States); Ikeda, T [Toshiba Corp. (Japan); Oguma, M [Hitachi, Ltd (Japan); Takei, T [Tokyo Electric Power Co., Inc. (Japan); Vitanza, C; Karlsen, T M [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-02-01

    The Halden BWR corrosion test loop was constructed to evaluate the impact of water chemistry variables, heat flux and boiling condition on corrosion performance of Zr-alloys in a simulated BWR environment. The loop consists of two in-core rigs, one for testing fuel rod segments and the other for evaluating water chemistry variables utilizing four miniautoclaves. Ten coupon specimens are enclosed in each miniautoclave. The Zr-alloys for the test include Zircaloy-2 having different nodular corrosion resistance and five new alloys. The first and second of the six irradiation tests planned in this program were completed. Post-irradiation examination of those test specimens have shown that the test loop is capable of producing nodular corrosion on the fuel rod cladding tested under the reference chemistry condition. The miniautoclave tests showed that nodular corrosion could be formed without flux and boiling under some water chemistry conditions and the new alloys, generally, had higher corrosion resistance than the Zircaloy in high oxygen environments. (author). 5 refs, 4 figs, 5 tabs.

  19. Fracture toughness of irradiated wrought and cast austenitic stainless steels in BWR environment

    International Nuclear Information System (INIS)

    Chopra, O.K.; Gruber, E.E.; Shack, W.J.

    2007-01-01

    Experimental data are presented on the fracture toughness of wrought and cast austenitic stainless steels (SSs) that were irradiated to a fluence of ∼ 1.5 x 10 21 n/cm 2 (E > 1 MeV) * (∼ 2.3 dpa) at 296-305 o C. To evaluate the possible effects of test environment and crack morphology on the fracture toughness of these steels, all tests were conducted in normal-water-chemistry boiling water reactor (BWR) environments at ∼ 289 o C. Companion tests were also conducted in air on the same material for comparison. The fracture toughness J-R curves for SS weld heat-affected-zone materials in BWR water were found to be comparable to those in air. However, the results of tests on sensitized Type 304 SS and thermally aged cast CF-8M steel suggested a possible effect of water environment. The available fracture toughness data on irradiated austenitic SSs were reviewed to assess the potential for radiation embrittlement of reactor-core internal components. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components are also discussed. (author)

  20. BWR thermohydraulics simulation on the AD-10 peripheral processor

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

    1983-01-01

    This presentation demonstrates the feasibility of simulating plant transients and severe abnormal transients in nuclear power plants at much faster than real-time computing speeds in a low-cost, dedicated, interactive minicomputer. This is achieved by implementing advanced modeling techniques in modern, special-purpose peripheral processors for high-speed system simulation. The results of this demonstration will impact safety analyses and parametric studies, studies on operator responses and control system failures and it will make possible the continuous on-line monitoring of plant performance and the detection and diagnosis of system or component failures

  1. Numerical simulation of progressive BWR fuel inlet orifices

    International Nuclear Information System (INIS)

    Sara Lundgren; Hernan Tinoco; Aleksander Pohl; Wiktor Frid

    2005-01-01

    Full text of publication follows: A 'progressive' orifice is characterized by an edge-shaped hole that gives a Reynolds number dependent resistance coefficient. For Reynolds numbers smaller than a critical one, the resistance coefficient has a high constant value that drops to a much lower value for Reynolds numbers greater than this critical value. A similar effect is widely known for external flows around bodies of different shapes, i. e. spheres, cylinders, etc., and the sudden drop in drag coefficient is due to the shift from laminar to turbulent boundary-layer flow. Experimentally, progressive orifices have been investigated under high-pressure and high-temperature conditions by Akiba et al. (2001) for a reduced set of geometrical parameters. Using the sparse experimental data, a core stability study was carried out by Forsmaks Kraftgrupp AB that showed an improvement in core stability but without the expected reduction in pump power at normal operation. The reason for this partial success was the impossibility of optimizing the fuel inlet pressure drop owing to the limited amount of available data. Due to the high costs associated with the experimental generation of high-pressure, high-temperature data, it was considered that, if possible, the lacking data could be generated numerically at much lower cost. Therefore, the present work deals with the possibility of numerically simulate the flow through progressive orifices, and with the conditions under which to reproduce and generate resistance coefficient data by means of a commercial CFD-code. The results obtained with a two-dimensional, axisymmetric approximation show that Reynolds Averaged Navier-Stokes (RANS) turbulence models are able to qualitatively capture the physics of the phenomenon but with an earlier transition to turbulent boundary-layer flow and with an underestimation of the resistance coefficient by approximately 20 %. This underestimation of the resistance coefficient is related to the two

  2. Numerical simulation of progressive BWR fuel inlet orifices

    Energy Technology Data Exchange (ETDEWEB)

    Sara Lundgren; Hernan Tinoco [Forsmarks Kraftgrupp AB, 742 03 Oesthammar (Sweden); Aleksander Pohl; Wiktor Frid [The Royal Institute of Technology, Dept. Energy Technology, SE-100 44 Stockholm (Sweden)

    2005-07-01

    Full text of publication follows: A 'progressive' orifice is characterized by an edge-shaped hole that gives a Reynolds number dependent resistance coefficient. For Reynolds numbers smaller than a critical one, the resistance coefficient has a high constant value that drops to a much lower value for Reynolds numbers greater than this critical value. A similar effect is widely known for external flows around bodies of different shapes, i. e. spheres, cylinders, etc., and the sudden drop in drag coefficient is due to the shift from laminar to turbulent boundary-layer flow. Experimentally, progressive orifices have been investigated under high-pressure and high-temperature conditions by Akiba et al. (2001) for a reduced set of geometrical parameters. Using the sparse experimental data, a core stability study was carried out by Forsmaks Kraftgrupp AB that showed an improvement in core stability but without the expected reduction in pump power at normal operation. The reason for this partial success was the impossibility of optimizing the fuel inlet pressure drop owing to the limited amount of available data. Due to the high costs associated with the experimental generation of high-pressure, high-temperature data, it was considered that, if possible, the lacking data could be generated numerically at much lower cost. Therefore, the present work deals with the possibility of numerically simulate the flow through progressive orifices, and with the conditions under which to reproduce and generate resistance coefficient data by means of a commercial CFD-code. The results obtained with a two-dimensional, axisymmetric approximation show that Reynolds Averaged Navier-Stokes (RANS) turbulence models are able to qualitatively capture the physics of the phenomenon but with an earlier transition to turbulent boundary-layer flow and with an underestimation of the resistance coefficient by approximately 20 %. This underestimation of the resistance coefficient is related to

  3. BWR Full Integral Simulation Test (FIST) program: facility description report

    International Nuclear Information System (INIS)

    Stephens, A.G.

    1984-09-01

    A new boiling water reactor safety test facility (FIST, Full Integral Simulation Test) is described. It will be used to investigate small breaks and operational transients and to tie results from such tests to earlier large-break test results determined in the TLTA. The new facility's full height and prototypical components constitute a major scaling improvement over earlier test facilities. A heated feedwater system, permitting steady-state operation, and a large increase in the number of measurements are other significant improvements. The program background is outlined and program objectives defined. The design basis is presented together with a detailed, complete description of the facility and measurements to be made. An extensive component scaling analysis and prediction of performance are presented

  4. Quantitative evaluation for training results of nuclear plant operator on BWR simulator

    International Nuclear Information System (INIS)

    Sato, Takao; Sato, Tatsuaki; Onishi, Hiroshi; Miyakita, Kohji; Mizuno, Toshiyuki

    1985-01-01

    Recently, the reliability of neclear power plants has largely risen, and the abnormal phenomena in the actual plants are rarely encountered. Therefore, the training using simulators becomes more and more important. In BWR Operator Training Center Corp., the training of the operators of BWR power plants has been continued for about ten years using a simulator having the nearly same function as the actual plants. The recent high capacity ratio of nuclear power plants has been mostly supported by excellent operators trained in this way. Taking the opportunity of the start of operation of No.2 simulator, effort has been exerted to quantitatively grasp the effect of training and to heighten the quality of training. The outline of seven training courses is shown. The technical ability required for operators, the items of quantifying the effect of training, that is, operational errors and the time required for operation, the method of quantifying, the method of collecting the data and the results of the application to the actual training are described. It was found that this method is suitable to quantify the effect of training. (Kako, I.)

  5. BISEN: Biochemical simulation environment

    NARCIS (Netherlands)

    Vanlier, J.; Wu, F.; Qi, F.; Vinnakota, K.C.; Han, Y.; Dash, R.K.; Yang, F.; Beard, D.A.

    2009-01-01

    The Biochemical Simulation Environment (BISEN) is a suite of tools for generating equations and associated computer programs for simulating biochemical systems in the MATLAB® computing environment. This is the first package that can generate appropriate systems of differential equations for

  6. Core heat transfer analysis during a BWR LOCA simulation experiment at ROSA-III

    International Nuclear Information System (INIS)

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

    1987-01-01

    The ROSA-III test facility is a 1/424-th volumetrically scaled BWR/6 simulator with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA). Heat transfer analyses for 5, 15, 50 and 200% break tests were conducted to understand the basic heat transfer behavior in the core under BWR LOCA conditions and to obtain a data base of post-critical heat flux (CHF) heat transfer coefficients and quench temperature. The results show that the convective heat transfer coefficient of dried-out rods at the core midplane during a steam cooling period is less than approximately 120 W/m 2 K. It is larger than existing data measured at lower pressures during a spray cooling period. Bottom-up quench temperatures are given by a simple equations: The sum of the saturation temperature and a constant of 262 K. Then the heat transfer model in the RELAP4/MOD6/U4/J3 code was revised using the present results. The rod surface temperature behavior in the 200% break test was calculated better by using the revised model although the model is very simple. (orig.)

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

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Suzuki, Mitsuhiro; Koizumi, Yasuo

    1985-10-01

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

  8. An interactive simulation-based education system for BWR emergency, procedure guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Tanikawa, Naoshi; Shida, Touichi [Hitachi Ltd (Japan). Hitachi Works; Ujita, Hiroshi; Yokota, Takeshi; Kato, Kanji [Hitachi Ltd, (Japan). Energy Research Lab.

    1994-12-31

    When applying EPGs (Emergency Procedure Guidelines), an operator decides the operational procedure by predicting the change of parameters from the plant status, because EPGs are described in a symptom style for emergency conditions. Technical knowledge of the plant behavior and its operation are necessary for operator to understand the EPGs. An interactive simulation-based education system, EPG-ICAI (Intelligent Computer Assisted Instruction), has been developed for BWR plant operators to acquire the knowledge of EPGs. EPG-ICAI is designed to realize an effective education by the step-by-step study by using an interactive real time simulator and an individual education by applying an intelligent tutoring function. (orig.) (2 refs., 7 figs., 1 tab.).

  9. An interactive simulation-based education system for BWR emergency, procedure guidelines

    International Nuclear Information System (INIS)

    Tanikawa, Naoshi; Shida, Touichi; Ujita, Hiroshi; Yokota, Takeshi; Kato, Kanji

    1994-01-01

    When applying EPGs (Emergency Procedure Guidelines), an operator decides the operational procedure by predicting the change of parameters from the plant status, because EPGs are described in a symptom style for emergency conditions. Technical knowledge of the plant behavior and its operation are necessary for operator to understand the EPGs. An interactive simulation-based education system, EPG-ICAI (Intelligent Computer Assisted Instruction), has been developed for BWR plant operators to acquire the knowledge of EPGs. EPG-ICAI is designed to realize an effective education by the step-by-step study by using an interactive real time simulator and an individual education by applying an intelligent tutoring function. (orig.) (2 refs., 7 figs., 1 tab.)

  10. Development of neural network simulating power distribution of a BWR fuel bundle

    International Nuclear Information System (INIS)

    Tanabe, A.; Yamamoto, T.; Shinfuku, K.; Nakamae, T.

    1992-01-01

    A neural network model is developed to simulate the precise nuclear physics analysis program code for quick scoping survey calculations. The relation between enrichment and local power distribution of BWR fuel bundles was learned using two layers neural network (ENET). A new model is to introduce burnable neutron absorber (Gadolinia), added to several fuel rods to decrease initial reactivity of fresh bundle. The 2nd stages three layers neural network (GNET) is added on the 1st stage network ENET. GNET studies the local distribution difference caused by Gadolinia. Using this method, it becomes possible to survey of the gradients of sigmoid functions and back propagation constants with reasonable time. Using 99 learning patterns of zero burnup, good error convergence curve is obtained after many trials. This neural network model is able to simulate no learned cases fairly as well as the learned cases. Computer time of this neural network model is about 100 times faster than a precise analysis model. (author)

  11. U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 7. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been recently expanded for BWR out-of-phase behavior. Out-of-phase oscillation is a phenomenon that occurs at BWRs. During this kind of event, half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. The HRS will be used for development and validation of stability monitoring and control techniques as part of an ongoing U.S. Department of Energy Nuclear Engineering Education and Research grant. The Penn State TRIGA reactor is used to simulate BWR fundamental mode power dynamics. The first harmonic mode power, together with detailed thermal hydraulics of boiling channels of both fundamental mode and first harmonic mode, is simulated digitally in real time with a computer. Simulations of boiling channels provide reactivity feedback to the TRIGA reactor, and the TRIGA reactor's power response is in turn fed into the channel simulations and the first harmonic mode power simulation. The combination of reactor power response and the simulated first harmonic power response with spatial distribution functions thus mimics the stability phenomena actually encountered in BWRs. The digital simulations of the boiling channels are performed by solving conservation equations for different regions in the channel with C-MEX S-functions. A fast three-dimensional (3-D) reactor power display of modal BWR power distribution was implemented using MATLAB graphics capability. Fundamental mode, first harmonic, together with the total power distribution over the reactor cross section, are displayed. Because of the large amount of computation for BWR boiling channel simulation and real-time data processing and graph generation, one computer is not sufficient to handle these jobs in the hybrid reactor simulation environment. A new three-computer setup has been

  12. Peach Bottom Turbine Trip Simulations with RETRAN Using INER/TPC BWR Transient Analysis Method

    International Nuclear Information System (INIS)

    Kao Lainsu; Chiang, Show-Chyuan

    2005-01-01

    The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations are described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT

  13. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  14. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  15. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  16. BWR simulation in a stationary state for the evaluation of fuel cell design

    International Nuclear Information System (INIS)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A.

    2014-10-01

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  17. Investigation on the electrochemical properties and crack growth rates of stainless steels in BWR alkaline environments

    International Nuclear Information System (INIS)

    Wang, L.H.; Hsu, T.Y.; Huang, C.S.

    2000-01-01

    Increasing pH of reactor water to mildly alkaline is considered as one of the mitigating water chemistry strategies to reduce the activity release of radioactive oxides and suppress the growth rate of stress corrosion cracking. However, only limited experimental data are currently available in the published literature, it is imperative to perform additional tests to verify the effectiveness of slightly alkaline reactor water. Because the electrochemical behavior and SCC are intricately related, this study will attempt to investigates the electrochemical properties and measures the crack growth rates (CGRs) of type 304 stainless steel (SS) in both normal water chemistry (200 ppb O 2 , neutral pH 25 ) and alkaline chemistry (200 ppb O 2 , pH 25 = 8.0). The additive for pH control is potassium hydroxide (KOH). The crack growth rate was monitored by reversing DC potential drop technique. The electrochemical measurements include AC impedance measurement and potential pulsing test to measure the repassivation behavior. The characteristics of electrochemical properties and its effect on stress corrosion crocking in BWR alkaline environments have been further examined. (author)

  18. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Ramirez G, C.; Chavez M, C.

    2012-10-01

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  19. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models

    International Nuclear Information System (INIS)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A.

    2003-01-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  20. Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

    International Nuclear Information System (INIS)

    Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

    2009-01-01

    The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

  1. Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)

    International Nuclear Information System (INIS)

    Burns, C. J.; Aumiler, D.L.

    2006-01-01

    The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF

  2. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models; SUN-RAH: simulador universitario de nucleoelectrica BWR basado en modelos de orden reducido

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2003-07-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  3. OECD/NRC BWR Turbine Trip Benchmark: Simulation by POLCA-T Code

    International Nuclear Information System (INIS)

    Panayotov, Dobromir

    2004-01-01

    Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and three-dimensional (3-D) neutron kinetics core models. Participation in the OECD/NRC BWR Turbine Trip (TT) Benchmark is a part of our efforts toward the code's validation. The paper describes the objectives for TT analyses and gives a brief overview of the developed plant system input deck and 3-D core model.The results of exercise 1, system model without netronics, are presented. Sensitivity studies performed cover the maximal time step, turbine stop valve position and mass flow, feedwater temperature, and steam bypass mass flow. Results of exercise 2, 3-D core neutronic and thermal-hydraulic model with boundary conditions, are also presented. Sensitivity studies include the core inlet temperature, cladding properties, and direct heating to core coolant and bypass.The entire plant model was validated in the framework of the benchmark's phase 3. Sensitivity studies include the effect of SCRAM initialization and carry-under. The results obtained - transient fission power and its initial axial distribution and steam dome, core exit, lower and upper plenum, main steam line, and turbine inlet pressures - showed good agreement with measured data. Thus, the POLCA-T code capabilities for correct simulation of pressurizing transients with very fast power were proved

  4. Electromagnetic Environments Simulator (EMES)

    International Nuclear Information System (INIS)

    Varnado, G.B.

    1975-11-01

    A multipurpose electromagnetic environments simulator has been designed to provide a capability for performing EMR, EMP, and lightning near stroke testing of systems, subsystems and components in a single facility. This report describes the final facility design and presents the analytical and experimental verification of the design

  5. 3D simulation of a core operation cycle of a BWR using Serpent

    International Nuclear Information System (INIS)

    Barrera Ch, M. A.; Del Valle G, E.; Gomez T, A. M.

    2016-09-01

    This work had the main goal to develop a methodology to obtain the length of an operating cycle of the core of a BWR under different operating states using the Serpent code. The reactor core modeled in Serpent is composed of 444 fuel assemblies (120 with fresh fuels and 324 fuels from previous cycles), 109 cruciform control rods and light water as moderator and coolant. Once the core of the reactor was modeled in Serpent (Three-dimensional) without considering the cruciform control rods, a simulation was carried out with different steps of burning in the operational state with the average values of the fuel temperature (900 K), moderator temperature (600 K) and voids fraction equal to 0.4. In addition, the thermal power considered was 2017 MWt. This operational state was chosen because a previous analysis (not shown in this work) was carried out in 4 types of control cells. The first and second control cell has all of its natural uranium fuel pellets, with control rod and without control rod respectively. The third and fourth control cell types have various types of enrichment, both natural uranium and gadolinium in their fuel pellets, with control rod and without control rod. The conclusion of this previous analysis was that the behavior of the effective multiplication factor along the fuel burnout within the four control cell types was almost unaffected by the fuel temperature but was affected by the voids fraction. Thus, for this operating cycle in the operating state defined above, its length was 14,63052 GW t/Tm. In addition, at the end of this cycle, the decay heat obtained was equal to 116.71 MWt and the inventory of the most important isotopes to be considered was obtained, such as some isotopes of uranium, neptune, plutonium, americium and curio. (Author)

  6. Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  7. CAPS Simulation Environment Development

    Science.gov (United States)

    Murphy, Douglas G.; Hoffman, James A.

    2005-01-01

    The final design for an effective Comet/Asteroid Protection System (CAPS) will likely come after a number of competing designs have been simulated and evaluated. Because of the large number of design parameters involved in a system capable of detecting an object, accurately determining its orbit, and diverting the impact threat, a comprehensive simulation environment will be an extremely valuable tool for the CAPS designers. A successful simulation/design tool will aid the user in identifying the critical parameters in the system and eventually allow for automatic optimization of the design once the relationships of the key parameters are understood. A CAPS configuration will consist of space-based detectors whose purpose is to scan the celestial sphere in search of objects likely to make a close approach to Earth and to determine with the greatest possible accuracy the orbits of those objects. Other components of a CAPS configuration may include systems for modifying the orbits of approaching objects, either for the purpose of preventing a collision or for positioning the object into an orbit where it can be studied or used as a mineral resource. The Synergistic Engineering Environment (SEE) is a space-systems design, evaluation, and visualization software tool being leveraged to simulate these aspects of the CAPS study. The long-term goal of the SEE is to provide capabilities to allow the user to build and compare various CAPS designs by running end-to-end simulations that encompass the scanning phase, the orbit determination phase, and the orbit modification phase of a given scenario. Herein, a brief description of the expected simulation phases is provided, the current status and available features of the SEE software system is reported, and examples are shown of how the system is used to build and evaluate a CAPS detection design. Conclusions and the roadmap for future development of the SEE are also presented.

  8. Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

    2011-07-01

    Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

  9. 3D simulation of a core operation cycle of a BWR using Serpent; Simulacion 3D de un ciclo de operacion del nucleo de un BWR usando SERPENT

    Energy Technology Data Exchange (ETDEWEB)

    Barrera Ch, M. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: rionchez@icloud.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This work had the main goal to develop a methodology to obtain the length of an operating cycle of the core of a BWR under different operating states using the Serpent code. The reactor core modeled in Serpent is composed of 444 fuel assemblies (120 with fresh fuels and 324 fuels from previous cycles), 109 cruciform control rods and light water as moderator and coolant. Once the core of the reactor was modeled in Serpent (Three-dimensional) without considering the cruciform control rods, a simulation was carried out with different steps of burning in the operational state with the average values of the fuel temperature (900 K), moderator temperature (600 K) and voids fraction equal to 0.4. In addition, the thermal power considered was 2017 MWt. This operational state was chosen because a previous analysis (not shown in this work) was carried out in 4 types of control cells. The first and second control cell has all of its natural uranium fuel pellets, with control rod and without control rod respectively. The third and fourth control cell types have various types of enrichment, both natural uranium and gadolinium in their fuel pellets, with control rod and without control rod. The conclusion of this previous analysis was that the behavior of the effective multiplication factor along the fuel burnout within the four control cell types was almost unaffected by the fuel temperature but was affected by the voids fraction. Thus, for this operating cycle in the operating state defined above, its length was 14,63052 GW t/Tm. In addition, at the end of this cycle, the decay heat obtained was equal to 116.71 MWt and the inventory of the most important isotopes to be considered was obtained, such as some isotopes of uranium, neptune, plutonium, americium and curio. (Author)

  10. Simulation of the operational monitoring of a BWR with Simulate-3

    International Nuclear Information System (INIS)

    Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  11. Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

    International Nuclear Information System (INIS)

    Griffin, F.P.

    1995-01-01

    A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B 4 C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence

  12. Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.; Nuenighoff, K.; Allelein, H.J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energie- und Klimaforschung (IEK), Sicherheitsforschung und Reaktortechnik (IEK-6)

    2011-07-01

    Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

  13. The corrosion potential of stainless steel in BWR environment comparison of data and modeling results

    International Nuclear Information System (INIS)

    Molander, Anders; Ullberg, Mats

    2004-01-01

    Corrosion potential measurements have been performed in Swedish BWRs during 25 years using commercially available monitoring equipment. Today, such measurements are performed on a routine basis in the BWRs on hydrogen water chemistry in Sweden. Measurements are usually performed at several monitoring locations in the plants. During the years, variations in the corrosion potential between different reactor cycles have been observed. Also, the corrosion potential can vary significantly during the reactor year. The changes have not always been easy to explain. Examples of in-plant data are given, demonstrating the need for a better understanding and for improved modeling tools. These examples were used as starting points for developing improved methods for corrosion potential modeling. A new tool recently developed, The Virtual ECP Laboratory, is described and applications to BWR conditions including some unexpected experimental corrosion potential responses are given. (author)

  14. Automatic determination of BWR fuel loading patterns based on K.E. technique with core physics simulation

    International Nuclear Information System (INIS)

    Ikehara, T.; Tsuiki, M.; Takeshita, T.

    1990-01-01

    On the basis oof a computerized search method, a prototype for a fuel loading pattern expert system has been developed to support designers in core design for BWRs. The method was implemented by coupling rules and core physics simulators into an inference engine to establish an automated generate-and-test cycle. A search control mechanism, which prunes paths to be searched and selects appropriate rules through the interaction with the user, was also introduced to accomplish an effective search. The constraints in BWR core design are: (1) cycle length more than L, (2) core shutdown margin more than S, and (3) thermal margin more than T. Here L, S, and T are the specified minimum values. In this system, individual rules contain the manipulation to improve the core shutdown margin explicitly. Other items were taken into account only implicitly. Several applications to the test cases were carried out. It was found that the results were comparable with those obtained by human expert engineers. Broad applicability of the present method in the BWR core design domain was proved

  15. CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor

    International Nuclear Information System (INIS)

    Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L.; Tijerina S, F.; Tapia M, R.

    2016-09-01

    In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

  16. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

    International Nuclear Information System (INIS)

    Hernandez, N.; Alonso, G.; Valle, E. del

    2004-01-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  17. Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic–neutronic simulations

    International Nuclear Information System (INIS)

    Ánchel, F.; Barrachina, T.; Miró, R.; Verdú, G.; Juanas, J.; Macián-Juan, R.

    2012-01-01

    Highlights: ► Best-estimate codes are affected by the uncertainty in the methods and the models. ► Influence of the uncertainty in the macroscopic cross-sections in a BWR and PWR RIA accidents analysis. ► The fast diffusion coefficient, the scattering cross section and both fission cross sections are the most influential factors. ► The absorption cross sections very little influence. ► Using a normal pdf the results are more “conservative” comparing the power peak reached with uncertainty quantified with a uniform pdf. - Abstract: The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether–Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.

  18. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

    1990-03-01

    Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

  19. Simulation of hydrogen deflagration and detonation in a BWR reactor building

    International Nuclear Information System (INIS)

    Manninen, M.; Silde, A.; Lindholm, I.; Huhtanen, R.; Sjoevall, H.

    2002-01-01

    A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm 2 leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis

  20. In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

    Science.gov (United States)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

  1. A pneumatic bellows-driven setup for controlled-distance electrochemical impedance measurements of Zircaloy-2 in simulated BWR conditions

    International Nuclear Information System (INIS)

    Arilahti, E.; Bojinov, M.; Hansson-Lyyra, L.

    2004-01-01

    This paper describes a novel pneumatic bellows-driven arrangement designed for controlled distance electrochemistry (CDE) measurements. The feasibility of the new arrangement has been verified by performing contact electric impedance measurements to study corrosion of Zircaloy-2 in a re-circulation loop simulating the BWR conditions. Until now, the measurements have been carried out using a step-motor driven controlled-distance electrochemistry (CDE) arrangement. The electrical and electrochemical properties of the pre transition oxide on Zircaloy-2 determined from these measurements were in good agreement with those estimated from measurements with a step-motor driven CDE. Furthermore, the results indicate that the bellows-driven CDE device is less sensitive to the contact pressure variation than the step-motor driven arrangement. This property combined with the bellows driven displacement mechanism provides a clear advantage for future in-core corrosion studies of fuel cladding materials. (Author)

  2. Low cycle corrosion fatigue properties of F316Ti in simulated LWR primary environment

    International Nuclear Information System (INIS)

    Xu Xuelian; Ding Yaping; Katada, Y.; Sato, S.

    1998-11-01

    Environment effect on fatigue performance of materials used for Pressurized boundary, including fatigue life and crack growth rate, are of importance to nuclear safety. To predict the fatigue life of nuclear materials and to improve the design of nuclear materials, it is necessary to investigated the material fatigue performances in corrosive environment and to get the fatigue data under its environment to be used in. Low cycle corrosion fatigue (CF) performance investigation of domestic F316Ti in simulated BWR and PWR primary environment was carried out. The result shows that the high temperature water environment is one of the most important factors on CF properties. For the same material, the low cycle fatigue life in high temperature air is longer than that in simulated BWR and PWR primary environments. In high temperature water, domestic F316Ti has almost the same low cycle corrosion fatigue performance as F316 (made in Japan). All of the fatigue data are scattered within ASME best-fit curve and ASME design fatigue curve. In high strain range, there is no significant difference of the CF performance for F316Ti in both of BWR and PWR primary environments. With the decrease of strain amplitude, the difference appears gradually. The data is located at the short life side of the fatigue data in simulated BWR primary environment. Titanium is distributed uniformly in F316Ti manufactured in Fushun Steel Factory. Ni, Cr, Mo in this material are located at the high side of the alloy chemical composition range. So, F316Ti has a better CF property in high temperature water

  3. Characterization of oxide films formed on steels in a BWR environment

    International Nuclear Information System (INIS)

    Honda, Takashi; Ohashi, Kenya; Kashimura, Eiji; Furutani, Yasumasa

    1988-01-01

    Environmental effects on corrosion bahaviors and properties of oxide films were evaluated for austenitic stainless and carbon steels in high-temperature water simulating a Boiling Water Reactor condition. The existence ratios of Cr and OH - in oxide films formed on stainless steel decreased and those of Fe, Ni and O 2- increased with increases of temperature and dissolved oxygen concentration. Changes of pH in the test region did not affect the composition of these species. These results indicated that Cr tended to combine with OH - , i.e., Cr existed as hydroxides or oxyhydroxides. Further, Fe and Ni tended to form spinel type oxides, which were indentified by XRD. In addition, the corrosion resistance of stainless steel was higher than that of carbon steel in all environments. The protectivity of magnetite films on carbon steel increased with temperature, dissolved oxygen concentration and pH. However, Ni ferrite, formed on stainless steel, further improved the corrosion resistance under such conditions. On the other hand, as the solubility of magnetite increased with decreases in the above mentioned factors, the corrosion resistance of carbon steel decreased. But, even under such conditions Cr, contained in stainless steel, tended to form stable films and suppressed corrosion. (author)

  4. BWR shutdown analyzer using artificial intelligence (AI) techniques

    International Nuclear Information System (INIS)

    Cain, D.G.

    1986-01-01

    A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

  5. Instructional environments for simulations.

    NARCIS (Netherlands)

    van Berkum, J.J.A.; de Jong, T.

    1991-01-01

    The use of computer simulations in education and training can have substantial advantages over other approaches. In comparison with alternatives such as textbooks, lectures, and tutorial courseware, a simulation-based approach offers the opportunity to learn in a relatively realistic problem-solving

  6. Instructional environments for simulations

    NARCIS (Netherlands)

    van Berkum, Jos J.A.; de Jong, Anthonius J.M.

    1991-01-01

    The use of computer simulations in education and training can have substantial advantages over other approaches. In comparison with alternatives such as textbooks, lectures, and tutorial courseware, a simulation-based approach offers the opportunity to learn in a relatively realistic problem-solving

  7. A nonlinear 3D real-time model for simulation of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.

    1982-02-01

    A nonlinear transient model for BWR nuclear power plants which consists of a 3D-core (subdivided into a number of superboxes, and with parallel flow and subcooled boiling), a top plenum, steam removal and feed water systems and main coolant recirculation pumps is given. The model describes the local core and global plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. The model which forms the basis for the digital code GARLIC-B (Er et al. 82) is aimed to be used on an on-site process computer in parallel to the actual reactor process (or even in predictive mode). Thus, special measures had to be taken into account in order to increase the computational speed and reduce the necessary computer storage. This could be achieved by - separating the neutron and power kinetics from the xenon-iodine dynamics, - treating the neutron kinetics and most of the thermodynamics and hydrodynamics in a pseudostationary way, - developing a special coupling coefficient concept to describe the neutron diffusion, calculating the coupling coefficients from a basic neutron kinetics code, - combining coarse mesh elements into superboxes, taking advantage of the symmetry properties of the core and - applying a sparse matrix technique for solving the resulting algebraic power equation system. (orig.) [de

  8. An improved one-and-a-half group BWR core simulator for a new-generation core management system

    International Nuclear Information System (INIS)

    Iwamoto, Tatsuya; Yamamoto, Munenari

    2000-01-01

    An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good. (author)

  9. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Cardenas V, J.; Filio L, C.

    2016-09-01

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  10. Weightless environment simulation test; Mujuryo simulation shiken

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, K.; Yamamoto, T.; Kato, F. [Kawasaki Heavy Industries, Ltd., Kobe (Japan)

    1997-07-20

    Kawasaki Heavy Industries, Ltd., delivered a Weightless Environment Test System (WETS) to National Space Development Agency of Japan in 1994. This system creates a weightless environment similar to that in space by balancing gravity and buoyancy in the water, and is constituted of a large water tank, facilities to supply air and cooling water to space suits worn in the water, etc. In this report, a weightless environment simulation test and the facilities to supply air and cooling water are described. In the weightless environment simulation test, the astronaut to undergo tests and training wears a space suit quite similar to the suit worn on the orbit, and performs EVA/IVA (extravehicular activities/intravehicular activities) around a JEM (Japanese Experimental Module) mockup installed in the water verifying JEM design specifications, preparing manuals for operations on the orbit, or receives basic space-related drill and training. An EVA weightless environment simulation test No. 3 was accomplished with success in January, 1997, when the supply of breathing water and cooling water to the space suit, etc., were carried out with safety and reliability. 2 refs., 8 figs., 2 tabs.

  11. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  12. Impact on radioecological conditions in the environment of a BWR power station

    International Nuclear Information System (INIS)

    Bhat, I.S.; Hegde, A.G.; Kamath, P.R.

    1979-01-01

    The Environmental Survey Laboratory at Tarapur has monitored environment of the Tarapur Atomic Power Station (TAPS) right from the preoperational period in 1965 and then during the operational phase. Impact of release of radioactive effluents - liquid wastes to the sea and gaseous wastes through stack - of the TAPS on environmental radioactivity levels is described. 60 Co is absent from the TAPS colony air particulate sample, but is found to the extent of 5x10 -3 pCi/M 3 in samples at site since 1972. 131 I is found in the range of 1 to 5 pCi/g in the goat thyroids in the downwind area from 1971. Cumulative radiation dose in population centres as measured by thermoluminescent dosemeters has not shown any detectable increase, except the two villages in direction of prevailing wind 2 km away from TAPS. The cumulative dose in these two villages showed an increase of 5 to 10 mr/year from 1971. No detectable increase in radioactivity is found in vegetation close to the TAPS. Offshore seawaters beyond 5 km have not shown any signjficant concentrations of radionuclides, but the near shore waters along the coast showed increased activity of radioiodine and radiocesium. The silt has shown an increased 60 Co activity. The near shore sea food o organisms have shown the pick-up and build-up of sup(131)I, sup(134,137)Cs and sup(60)Co. Internal dose to the populations in the vicinity is above the natural preoperational background but within the recommended limits. The waste treatment processes at TAPS were augmented in 1973-74 by addition of storage tanks for: (1) the decay of short-lived nuclides and (2) removal of radiocobalt and radiocesium by flocculation vermiculite column absorption process. With this augmentation, levels of radioactivity ir sea water, silt and seafood have shown a declining trend. (M.G.B.)

  13. IGSCC crack propagation rate measurement in BWR environments. Executive summary of a Round Robin study

    International Nuclear Information System (INIS)

    Andresen, Peter L.

    1998-01-01

    that a complete transition is made from the transgranular fatigue pre-crack to an intergranular stress corrosion crack. Retarded or completely stalled crack growth was best addressed by imposing very gentle unloading cycles to re-initiate and sustain crack growth. While comparatively little of the total testing time successfully produced meaningful crack growth rates, the data exhibit the expected high crack growth rates in high dissolved oxygen environments. This is an important conclusion, as U.S. industry efforts have shown that a remarkably different growth rate is predicted based on statistical analyses of a broader collection of scattered crack growth rate data in sensitized type 304 stainless steel. The scatter in the data clearly dilutes all trends in SCC response, as the correlation, e.g., with corrosion potential (all other effects normalized in the correlation model) is quite weak - in addition to the dependence on crack growth being shallow. The origin of the weak correlation, shallow dependence, and poor agreement with other sets of well-controlled data is a myriad of experimental and interpretational complexities and flaws, so that the mean of such data is the mean of the flaws, not the mean of the true SCC response. In focusing on a single stress intensity and high dissolved oxygen / corrosion potential conditions, this program obviously does not address the broad range of important stress corrosion cracking dependencies on stress intensity, corrosion potential, aqueous impurities, temperature, degree of sensitization, irradiation, material type, etc. However, it invaluably elucidates the complexities involved in generating and interpreting stress corrosion cracking data. It also underscores the crucial overall role of developing a fundamental understanding of SCC and a recognition of the common elements or 'linkages' among SCC susceptible materials. These are necessary because of the sophistication required to generate high quality SCC data, and the

  14. IGSCC crack propagation rate measurement in BWR environments. Executive summary of a Round Robin study

    Energy Technology Data Exchange (ETDEWEB)

    Andresen, Peter L. [GE Corporate Research and Development, Schenectady, NY (United States)

    1998-12-31

    that a complete transition is made from the transgranular fatigue pre-crack to an intergranular stress corrosion crack. Retarded or completely stalled crack growth was best addressed by imposing very gentle unloading cycles to re-initiate and sustain crack growth. While comparatively little of the total testing time successfully produced meaningful crack growth rates, the data exhibit the expected high crack growth rates in high dissolved oxygen environments. This is an important conclusion, as U.S. industry efforts have shown that a remarkably different growth rate is predicted based on statistical analyses of a broader collection of scattered crack growth rate data in sensitized type 304 stainless steel. The scatter in the data clearly dilutes all trends in SCC response, as the correlation, e.g., with corrosion potential (all other effects normalized in the correlation model) is quite weak - in addition to the dependence on crack growth being shallow. The origin of the weak correlation, shallow dependence, and poor agreement with other sets of well-controlled data is a myriad of experimental and interpretational complexities and flaws, so that the mean of such data is the mean of the flaws, not the mean of the true SCC response. In focusing on a single stress intensity and high dissolved oxygen / corrosion potential conditions, this program obviously does not address the broad range of important stress corrosion cracking dependencies on stress intensity, corrosion potential, aqueous impurities, temperature, degree of sensitization, irradiation, material type, etc. However, it invaluably elucidates the complexities involved in generating and interpreting stress corrosion cracking data. It also underscores the crucial overall role of developing a fundamental understanding of SCC and a recognition of the common elements or `linkages` among SCC susceptible materials. These are necessary because of the sophistication required to generate high quality SCC data, and the

  15. High-speed BWR power plant simulations on the special-purpose peripheral processor AD10

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

    1985-01-01

    A newly developed technique is described for fast, on-line simulations of normal and accidental transients in nuclear power plants. The technique is based on the utilization of the special-purpose peripheral processor AD10, which is specifically designed for high-speed systems simulations through integration of large systems of nonlinear ordinary differential equations. The Peach Bottom-II Boiling Water Reactor power plant has been simulated and results are presented. It is shown that the new technique not only advances safety analyses but also supports plant monitoring, failure diagnosis and accident mitigation, as well as the training of nuclear power plant operators. (author)

  16. Kohonen mapping of the crack growth under fatigue loading conditions of stainless steels in BWR environments and of nickel alloys in PWR environments

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, Mirna

    2008-01-01

    In this study, crack growth rate data under fatigue loading conditions generated by Argonne National Laboratories and published in 2006 were analyzed [O.K. Chopra, B. Alexandreanu, E.E. Gruber, R.S. Daum, W.J. Shack, Argonne National Laboratory, NUREG CR 6891-series ANL 04/20, Crack Growth Rates of Austenitic Stainless Steel Weld Heat Affected Zone in BWR Environments, January, 2006; B. Alexandreanu, O.K. Chopra, H.M. Chung, E.E. Gruber, W.K. Soppet, R.W. Strain, W.J. Shack, Environmentally Assisted Cracking in Light Water Reactors, vol. 34 in the NUREG/CR-4667 series annual report of Argonne National Laboratory program studies for Calendar (Annual Report 2003). Manuscript Completed: May 2005, Date Published: May 2006], and reported by DoE [B. Alexandreanu, O.K. Chopra, W.J. Shack, S. Crane, H.J. Gonzalez, NRC, Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments, NUREG/CR-6964, May 2008]. The data collected were measured on austenitic stainless steels in BWR (boiling water reactor) environments and on nickel alloys in PWR (pressurized water reactor) environments. The data collected contained information on material composition, temperature, conductivity of the environment, oxygen concentration, irradiated sample information, weld information, electrochemical potential, load ratio, rise time, hydrogen concentration, hold time, down time, maximum stress intensity factor (K max ), stress intensity range (ΔK max ), crack length, and crack growth rates (CGR). Each position on that Kohonen map is called a cell. A Kohonen map clusters vectors of information by 'similarities.' Vectors of information were formed using the metal composition, followed by the environmental conditions used in each experiments, and finally followed by the crack growth rate (CGR) measured when a sample of pre-cracked metal is set in an environment and the sample is cyclically loaded

  17. BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  18. Plant analyzer development for high-speed interactive simulation of BWR plant transients

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.

    1986-01-01

    Advanced modeling techniques have been combined with modern, special-purpose peripheral minicomputer technology to develop a plant analyzer which provides realistic and accurate predictions of plant transients and severe off-normal events in nuclear power plants through on-line simulations at speeds of approximately 10 times faster than actual process speeds. The new simulation technology serves not only for carrying out routinely and efficiently safety analyses, optimizations of emergency procedures and design changes, parametric studies for obtaining safety margins and for generic training but also for assisting plant operations. Five modeling principles are presented which serve to achieve high-speed simulation of neutron kinetics, thermal conduction, nonhomogeneous and nonequilibrium two-phase flow coolant dynamics, steam line acoustical effects, and the dynamics of the balance of plant and containment systems, control systems and plant protection systems. 21 refs

  19. General Electric Company analytical model for loss-of-coolant analysis in accordance with 10CFR50 appendix K, amendment No. 3: effect of steam environment on BWR core spray distribution

    International Nuclear Information System (INIS)

    1977-04-01

    The core spray sparger designs of the BWR/2 through BWR/5 product lines were verified by means of full-scale mock-ups tested in air at various flow conditions. In 1974, an overseas technical partner of General Electric reported that a steam environment changed the individual core spray nozzle patterns when compared to patterns measured in air. This document describes preliminary findings of how a steam environment alters the core spray nozzle pattern, and the actions which General Electric is pursuing to quantify the steam effects

  20. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R.

    2015-09-01

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm 2 and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  1. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5

    International Nuclear Information System (INIS)

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

    2014-10-01

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm 2 ) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm 2 ), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  2. Plant analyzer for high-speed interactive simulation of BWR plant transients

    International Nuclear Information System (INIS)

    Cheng, H.S.; Lekach, S.V.; Mallen, A.N.; Wulff, W.; Cerbone, R.J.

    1984-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology was utilized to develop a plant analyzer which affords realistic predictions of plant transients and severe off-normal events in LWR power plants through on-line simulations at speeds up to 10 times faster than actual process speeds. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the entire balance of the plant. Reactor core models include point kinetics with reactivity feedback due to void fraction, fuel temperature, coolant temperature, and boron concentration as well as a conduction model for predicting fuel and clad temperatures. Control systems and trip logic for plant protection systems are also simulated. The AD10 of Applied Dynamics International, a special-purpose peripheral processor, is used as the principal hardware of the plant analyzer

  3. The BWR core simulator COSIMA with 2 group nodal flux expansion and control rod history

    International Nuclear Information System (INIS)

    Hoejerup, C.F.

    1989-08-01

    The boiling water simulator NOTAM has been modified and improved in several aspects: - The ''1 1/2'' energy group TRILUX nodal flux solution method has been exchanged with a 2 group modal expansion method. - Control rod ''history'' has been introduced. - Precalculated instrument factors have been introduced. The paper describes these improvements, which were considered sufficiently large to justify a new name to the programme: COSIMA. (author)

  4. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H. P

    2002-03-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

  5. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  6. FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients

    International Nuclear Information System (INIS)

    1988-01-01

    1 - Description of test facility: In the FIX-II pump trip experiments, mass flow and power transients were simulated subsequent to a total loss of power to the recirculation pumps in an internal pump boiling water reactor. The aim was to determine the initial power limit to give dryout in the fuel bundle for the specified transient. In addition, the peak cladding temperature was measured and the rewetting was studied. 2 - Description of test: Pump trip experiment 2032 was a part of test group 2, i.e. the mass flow transient was to simulate the pump coast down with a pump inertia of 11.3 kg.m -2 . The initial power in the 36-rod bundle was 4.44 MW which gave dryout after 1.4 s from the start of the flow transient. A maximum rod cladding temperature of 457 degrees C was measured. Rewetting was obtained after 7.6 s. 3 - Experimental limitations or shortcomings: No ECCS injection systems

  7. Numerical Simulations and Design Optimization of the PHT Loop of Natural Circulation BWR

    Directory of Open Access Journals (Sweden)

    G. V. Durga Prasad

    2008-01-01

    Full Text Available Mathematical modeling and numerical simulation of natural circulation boiling water reactor (NCBWR are very important in order to study its performance for different designs and various off-design conditions and for design optimization. In the present work, parametric studies of the primary heat transport loop of NCBWR have been performed using lumped parameter models and RELAP5/MOD3.4 code. The lumped parameter models are based on the drift flux model and homogeneous equilibrium mixture (HEM model of two-phase flow. Numerical simulations are performed with both models. Compared to the results obtained from the HEM model, those obtained from the drift flux model are closer to RELAP5. The variations of critical heat flux with various geometric parameters and operating conditions are thoroughly investigated. The material required to construct the primary heat transport (PHT loop of NCBWR has been minimized using sequential quadratic programming. The stability of NCBWR has also been verified at the optimum point.

  8. A method of taking control rod history into account in core simulation calculations for BWR'S

    International Nuclear Information System (INIS)

    Hojerup, C.F.; Nonbol, E.

    1990-01-01

    The problem of taking control rod history into account in core simulator codes using precalculated cross sections has been examined, and two methods have been devised and tested. The very demanding first method, using the accumulated control rod in burn-up as a parameter, turned out to be even more inaccurate than the much less demanding second method, which only requires two full burn-up histories, one with the control rod in all the time, and another with the control rod out all the time. From the analysis it can be seen that the proper treatment of the control rod history is quite important, both for the cross sections, as several per cent on the reactivity are at stake, as for the pin powers, which for some pins are very much affected

  9. Simulation of the electro-hydraulic control system of a BWR-5

    International Nuclear Information System (INIS)

    Acosta, M.; Montoya, J.; Chavez, H.

    1986-01-01

    The methodology used to develop the mathematical models for the simulation of the principal turbine electro-hydraulic control of the Laguna Verde Nuclear Plant (LVNP) is presented in this report. The development of the systems mathematical model is based on the response curves of each of its elements. Therefore, little error is expected with respect to real results. On the other hand, due to the fact that the greater part of the systems dynamics is governed by first order differential equations the explicit solution method is used allowing to solve the equations algebraically. The model is validated by comparing real valves and the ones obtained through our model. The analogical and logical parts will be tested considering transitory and steady state situations. The results are presented as computer graphs

  10. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2016-09-15

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  11. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  12. Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Urquidi-Macdonald, Mirna [Penn State University, 212 Earth-Engineering Science Building, University Park, PA 16801 (United States)

    2004-07-01

    Many publications have shown that crack growth rates (CGR) due to intergranular stress corrosion cracking (IGSCC) of metals is dependent on many parameters related to the manufacturing process of the steel and the environment to which the steel is exposed. Those parameters include, but are not restricted to, the concentration of chloride, fluoride, nitrates, and sulfates, pH, fluid velocity, electrochemical potential (ECP), electrolyte conductivity, stress and sensitization applied to the steel during its production and use. It is not well established how combinations of each of these parameters impact the CGR. Many different models and beliefs have been published, resulting in predictions that sometimes disagree with experimental observations. To some extent, the models are the closest to the nature of IGSCC, however, there is not a model that fully describes the entire range of observations, due to the difficulty of the problem. Among the models, the Fracture Environment Model, developed by Macdonald et al., is the most physico-chemical model, accounting for experimental observations in a wide range of environments or ECPs. In this work, we collected experimental data on BWR environments and designed a data mining pattern recognition model to learn from that data. The model was used to generate CGR estimations as a function of ECP on a BWR environment. The results of the predictive model were compared to the Fracture Environment Model predictions. The results from those two models are very close to the experimental observations of the area corresponding to creep and IGSCC controlled by diffusion. At more negative ECPs than the potential corresponding to creep, the pattern recognition predicts an increase of CGR with decreasing ECP, while the Fracture Environment Model predicts the opposite. The results of this comparison confirm that the pattern recognition model covers 3 phenomena: hydrogen embrittlement at very negative ECP, creep at intermediate ECP, and IGSCC

  13. Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWR environments

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, Mirna

    2004-01-01

    Many publications have shown that crack growth rates (CGR) due to intergranular stress corrosion cracking (IGSCC) of metals is dependent on many parameters related to the manufacturing process of the steel and the environment to which the steel is exposed. Those parameters include, but are not restricted to, the concentration of chloride, fluoride, nitrates, and sulfates, pH, fluid velocity, electrochemical potential (ECP), electrolyte conductivity, stress and sensitization applied to the steel during its production and use. It is not well established how combinations of each of these parameters impact the CGR. Many different models and beliefs have been published, resulting in predictions that sometimes disagree with experimental observations. To some extent, the models are the closest to the nature of IGSCC, however, there is not a model that fully describes the entire range of observations, due to the difficulty of the problem. Among the models, the Fracture Environment Model, developed by Macdonald et al., is the most physico-chemical model, accounting for experimental observations in a wide range of environments or ECPs. In this work, we collected experimental data on BWR environments and designed a data mining pattern recognition model to learn from that data. The model was used to generate CGR estimations as a function of ECP on a BWR environment. The results of the predictive model were compared to the Fracture Environment Model predictions. The results from those two models are very close to the experimental observations of the area corresponding to creep and IGSCC controlled by diffusion. At more negative ECPs than the potential corresponding to creep, the pattern recognition predicts an increase of CGR with decreasing ECP, while the Fracture Environment Model predicts the opposite. The results of this comparison confirm that the pattern recognition model covers 3 phenomena: hydrogen embrittlement at very negative ECP, creep at intermediate ECP, and IGSCC

  14. Simulation of BWR stability following an ATWS with boron injection using TRAC-BF1 with one-dimensional kinetics

    International Nuclear Information System (INIS)

    Lider, S.; Maclan, R.; Baratta, A.J.; Mahaffy, J.; Robinson, G.E.

    2004-01-01

    The scenario following an ATWS is characterized by the necessity to reduce the power in the reactor as fast as possible. The only means to insert a significant amount of negative reactivity in a BWR during an ATWS are the natural reactor negative void coefficient, and the injection of highly enriched boron through the SLCS. The ATWS management strategy suggested by BWR owner's group contemplates an initial rapid decrease in power as a result of the recirculation pump trip. This is followed by lowering of vessel water level and the injection of borated water into the lower plenum. A recent paper of Dias, et al. reports that reducing core power and lowering water level causes a reduction in boron mixing efficiency and the net effect is a longer time to shut down and an increase in Suppression Pool (SP) temperature. In the present paper, a series of analyses are made to address this issue. The preliminary results for the water level positions at TAF, TAF+1.5 m (TAF+5') and TAF+3 m (TAF+10') support the similar findings of Dias, et al. (author)

  15. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    Lopez R, A.

    2004-01-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  16. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    International Nuclear Information System (INIS)

    Thakre, S.; Ma, W.

    2013-08-01

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  17. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  18. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  19. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  20. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

    1989-01-01

    BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

  1. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  2. Virtual environments simulation in research reactor

    Science.gov (United States)

    Muhamad, Shalina Bt. Sheik; Bahrin, Muhammad Hannan Bin

    2017-01-01

    Virtual reality based simulations are interactive and engaging. It has the useful potential in improving safety training. Virtual reality technology can be used to train workers who are unfamiliar with the physical layout of an area. In this study, a simulation program based on the virtual environment at research reactor was developed. The platform used for virtual simulation is 3DVia software for which it's rendering capabilities, physics for movement and collision and interactive navigation features have been taken advantage of. A real research reactor was virtually modelled and simulated with the model of avatars adopted to simulate walking. Collision detection algorithms were developed for various parts of the 3D building and avatars to restrain the avatars to certain regions of the virtual environment. A user can control the avatar to move around inside the virtual environment. Thus, this work can assist in the training of personnel, as in evaluating the radiological safety of the research reactor facility.

  3. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

    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. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  4. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  5. Simulation of mechanical shock environments

    International Nuclear Information System (INIS)

    Lalanne, Christian.

    1975-07-01

    Shocks can produce a severe mechanical environment which must be taken into account when designing and developing new equipments. After some mathematical (Laplace and Fourier transforms) and mechanical recalls (response of a one degree freedom system to a sinusoidal excitation), different analysis methods are compared, these methods being the most used now to compare relative severities of tests and establish specifications. A few chapter deal with the different properties of simple, easy to produce, shock shapes. Then some now-in-use programmators or shock-machines specifications are shown. A final chapter concerns acceleration transducers [fr

  6. CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor; Simulacion CFD de los venteos rigidos de la contencion de un reactor BWR-5 Mark-II

    Energy Technology Data Exchange (ETDEWEB)

    Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F.; Tapia M, R., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

    2016-09-15

    In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

  7. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  8. Effect of a Chloride Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 2)

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.-P.

    2002-11-01

    Within the CASTOC-project (5 t h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the second test of working package (WP) 3 with a NaCl transient, performed at Paul Scherrer Institut (PSI). In the first part of the experiment, an actively growing EAC crack with a crack growth rate (CGR) in the range of the 'low-sulphur SCC line' of the GE-model was generated by periodical partial unloading (PPU) in oxygenated high-temperature, high-purity water (T = 288 o C, DO = 8 ppm). Then a chloride transient of 49 ppb Cl - was applied for ∼40 h. After this transient, the load level was reduced and the loading conditions were changed to pure cyclic loading. Thereupon a second transient with a chloride concentration of 49 ppb was applied. In both RPV steels, the first chloride transient of 49 ppb Cl - resulted in an acceleration of the EAC crack growth by more than one order of magnitude and in fast, stationary SCC crack growth during the constant load phase of the PPU cycles at K I values 1/2 . 3 h after adding chloride to the high-purity water, the EAC CGR started to increase in the high-sulphur RPV steel during the constant load phase of a PPU cycle and after 20 h a stationary EAC CGR value in the range of the 'high-sulphur SCC curve' of the GE-model was reached. After 5 h in high-purity water, the crack growth began to slow down after a partial unloading cycle and 15 h later it reached again a stationary CGR value in the range of the 'low-sulphur SCC curve' of the GE-model. The second chloride transient did not result in an acceleration of the crack growth in both investigated specimens. This was explained by crack closure effects, which occurred in both specimens after the reduction of the load. The CGR

  9. The AutoAssociative Neural Network in signal analysis: II. Application to on-line monitoring of a simulated BWR component

    International Nuclear Information System (INIS)

    Marseguerra, M.; Zoia, A.

    2005-01-01

    In this paper, Robust AutoAssociative Neural Networks (RAANN) are applied to a series of signals produced by the Halden simulator of the 1200MWe BWR Forsmark-3 plant in Sweden. The applications concern: - correction of drifts and gross errors in sensors, for diagnostic and control purposes, - cluster analysis, to individuate a failed component and the intensity of the failure, - forecasting system signals, for safety or economic purposes, - reconstruction of unmeasured signals (virtual sensors). In the attainment of the above results, the geometric interpretation of the mapping performed by the network, propounded in Part I of this work, has provided a reasoned choice of the most critical free parameter, i.e., the number f of nodes of the bottleneck layer, thus allowing a deep understanding of the network functioning and also avoiding the traditional and troubling procedure of selection by trial-and-error. The theoretical basis of this analysis, discussed in details in the companion paper, is founded on the idea of dimension and in particular of fractal dimension, which has been used as a numerical estimator of f

  10. Virtual Environments for Advanced Trainers and Simulators

    NARCIS (Netherlands)

    Jense, G.J.; Kuijper, F.

    1993-01-01

    Virtual environment technology is expected to make a big impact on future training and simulation systems. Direct stimulation of human senses (eyesight, auditory, tactile) and new paradigms for user input will improve the realism of simulations and thereby the effectiveness of training systems.

  11. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator

    International Nuclear Information System (INIS)

    Sanchez S, R.A.

    2003-01-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  12. BWR stability: history and state-of-the-art

    International Nuclear Information System (INIS)

    Yadigaroglu, George

    2014-01-01

    The paper briefly recalls the historical developments, reviews the important phenomena, the analytical and simulation tools that are used for the analysis of BWR stability focussing on the linear, frequency domain methods

  13. Broadband electromagnetic environments simulator (EMES)

    International Nuclear Information System (INIS)

    Pollard, N.

    1977-01-01

    A new test facility has been developed by Sandia Laboratories for determining the effects of electromagnetic environments on systems and components. The facility is capable of producing uniform, vertically polarized, continuous wave (CW) and pulsed fields over the frequency range of dc to 10 GHz. This broadband capability addresses the electromagnetic radiation (EMR) threat and is ideally suited to computer controlled sweeping and data acquisition. EMES is also capable of producing uniform transient fields having the wave shape and magnitude characteristic of a nuclear electromagnetic pulse (EMP) and near lightning. The design consists of a truncated, triplate, rectangular coaxial transmission line. The spacing between the flat center conductor and the ground planes is 4 meters. The line is terminated in its characteristic impedance of 50 ohms. At frequencies below the first resonance of the facility it behaves as a typical coaxial system. Above resonance, a wall of electromagnetic absorbing material provides a nonreflecting termination. Thus, EMES essentially combines the elements of a transmission line and an anechoic chamber. It will not radiate electromagnetic energy into the surrounding area because it is a shielded transmission line

  14. Kinematics of two-phase mixture level motion in BWR pressure vessels

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

    1985-01-01

    A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

  15. Simulation of the BWR experiments CORA-17 and CORA-28 using ATHLET-CD and assessment of BWR modelling. 1{sup st} Technical report. Validation and interpretation of the ATHLET-CD model basis; Simulation der SWR-Versuche CORA-17 und CORA-28 mit dem Programmsystem ATHLET-CD und Bewertung der SWR-Modellbasis. 1. Technischer Fachbericht. Validierung und Interpretation der ATHLET-CD Modellbasis

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, M.; Gremme, F.; Koch, M.K.

    2013-08-15

    The 1st Technical Report was prepared for the research project ''Validation and Interpretation of the ATHLET-CD Model Basis'' funded by the Federal Ministry of Economics and Technology (BMWi1501385) and carried out at the Chair of Energy Systems and Energy Economics at Ruhr-Universitaet Bochum (RUB). This report provides results of the simulation of the Boiling Water Reactor (BWR) experiments CORA-17 and -28 with ATHLET-CD Mod. 2.2A. The system code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) is developed by the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH. Code results are compared to measurements in order to assess and to analyze the capabilities of the current code version with regard to the modeling of BWR components. The CORA test series was carried out between the years 1987 and 1993 at the former Kernforschungszentrum Karlsruhe (KfK), now Karlsruhe Institute of Technology (KIT). The investigations provided experimental data regarding the material behavior during the early phase of core degradation in Light Water Reactors (LWR). The tests CORA17 and -28 represented a typical BWR arrangement of the fuel rod bundle and provided insights about the bundle behavior during the quenching process (CORA-17) and regarding the influence of a preoxidized bundle (CORA-28), respectively. The simulation results are analyzed and discussed in terms of the thermal bundle behavior, the zirconium oxidation in steam and the resulting hydrogen generation as well as the material relocation. In particular, the recently extended modeling capabilities of the code in terms of the relocation of BWR components like the absorber blade and the canister wall are assessed. The analysis shows that the code captures the thermal behavior in good agreement in both experiments. An even enhanced reproduction of the test CORA-28 is obtained in comparison to a calculation using the previous code version ATHLET-CD Mod

  16. Residual stress analysis in BWR pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Leung, C.P.; Pont, D.

    1992-06-01

    Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

  17. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Membrillo G, O. E.; Chavez M, C.

    2012-10-01

    The present work presents the modeling and simulation of the aspersion system to low pressure of reactor of the nuclear power plant of Laguna Verde using the nuclear code RELAP/SCDAP. The objective of the emergency systems inside a nuclear reactor is the cooling of the core, nor caring the performance of any other emergency system in the case of an accident design base for coolant loss. To obtain a simulation of the system is necessary to have a model based on their main components, pipes, pumps, valves, etc. This article describes the model for the simulation of the main line and the test line for the HPCS. At the moment we have the simulation of the reactor vessel and their systems associated to the nuclear power plant of Laguna Verde, this work will allow to associate the emergency system model LPCS to the vessel model. The simulation of the vessel and the emergency systems will allow knowing the behavior of the reactor in the stage of the coolant loos, giving the possibility to analyze diverse scenarios. The general model will provide an auxiliary tool for the training in classroom and at distance in the operation of nuclear power plants. (Author)

  18. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  19. Hard X-ray photoelectron spectroscopy study for transport behavior of CsI in heating test simulating a BWR severe accident condition: Chemical effects of boron vapors

    Energy Technology Data Exchange (ETDEWEB)

    Okane, T., E-mail: okanet@spring8.or.jp [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Kobata, M. [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Sato, I. [Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Kobayashi, K. [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Osaka, M. [Nuclear Science and Engineering Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Yamagami, H. [Quantum Beam Science Center, Japan Atomic Energy Agency, 1-1-1 Kouto, Sayo-cho, Hyogo, 679-5148 (Japan); Faculty of Science, Kyoto Sangyo University, Motoyama, Kamigamo, Kita-ku, Kyoto, 603-8555 (Japan)

    2016-02-15

    Highlights: • We have clarified the temperature-dependent chemical forms of Cs/I products. • We have examined the CsI-decomposing effects of B{sub 2}O{sub 3} vapor. • The possibility of Cs re-evaporation from CsI-deposited surface is suggested. • We have demonstrated the usefulness of HAXPES on FP chemistry. - Abstract: Transport behavior of CsI in the heating test, which simulated a BWR severe accident, was investigated by hard X-ray photoelectron spectroscopy (HAXPES) with an emphasis on the chemical effect of boron vapors. CsI deposited on metal tube at temperatures ranging from 150 °C to 750 °C was reacted with vapor/aerosol B{sub 2}O{sub 3}, and the chemical form of reaction products on the sample surface was examined from the HAXPES spectra of core levels, e.g., Ni 2p, Cs 3d and I 3d levels, and valence band. For the samples at ∼300 °C, while the chemical form of major product on the sample surface without an exposure to B{sub 2}O{sub 3} was suggested to be CsI from the HAXPES spectra, an intensity ratio of Cs/I was dramatically reduced at the sample surface after the reaction with B{sub 2}O{sub 3}. The results suggest the possibility of significant decomposition of deposited CsI induced by the chemical reaction with B{sub 2}O{sub 3} at specific temperatures.

  20. Effects of cold working ratio and stress intensity factor on intergranular stress corrosion cracking susceptibility of non-sensitized austenitic stainless steels in simulated BWR and PWR primary water

    International Nuclear Information System (INIS)

    Yaguchi, Seiji; Yonezawa, Toshio

    2012-01-01

    To evaluate the effects of cold working ratio, stress intensity factor and water chemistry on an IGSCC susceptibility of non-sensitized austenitic stainless steel, constant displacement DCB specimens were applied to SCC tests in simulated BWR and PWR primary water for the three types of austenitic stainless steels, Types 316L, 347 and 321. IGSCC was observed on the test specimens in simulated BWR and PWR primary water. The observed IGSCC was categorized into the following two types. The one is that the IGSCC observed on the same plane of the pre-fatigue crack plane, and the other is that the IGSCC observed on a plane perpendicular to the pre-fatigue crack plane. The later IGSCC fractured plane is parallel to the rolling plane of a cold rolled material. Two types of IGSCC fractured planes were changed according to the combination of the testing conditions (cold working ratio, stress intensity factor and simulated water). It seems to suggest that the most susceptible plane due to fabrication process of materials might play a significant role of IGSCC for non-sensitized cold worked austenitic stainless steels, especially, in simulated PWR primary water. Based upon evaluating on the reference crack growth rate (R-CGR) of the test specimens, the R-CGR seems to be mainly affected by cold working ratio. In case of simulated PWR primary water, it seems that the effect of metallurgical aspects dominates IGSCC susceptibility. (author)

  1. Development of BWR [boiling water reactor] and PWR [pressurized water reactor] event descriptions for nuclear facility simulator training

    International Nuclear Information System (INIS)

    Carter, R.J.; Bovell, C.R.

    1987-01-01

    A number of tools that can aid nuclear facility training developers in designing realistic simulator scenarios have been developed. This paper describes each of the tools, i.e., event lists, events-by-competencies matrices, and event descriptions, and illustrates how the tools can be used to construct scenarios

  2. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  3. Best-estimate analysis development for BWR systems

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

    1986-01-01

    The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

  4. Hydraulic modeling and simulation of a System Division of Essential Service Water in a BWR plant with Flow master

    International Nuclear Information System (INIS)

    Vegazo Juzgado, L.; Rodriguez Garcia, G. M.; Mota Coloma, M.

    2012-01-01

    At the conclusion of the project can say that Flow master is a simulation tool that allows you to create your model from a library of components and obtain useful results from the point of view of the operation, engineering and maintenance. Compared to previous software from the point of view of use, can comment that Flow master is a tool which has an intuitive and user-friendly interaction between the user and the program thus facilitating the modeling of the system and definition of the components of same.

  5. BWR modeling capability and Scale/Triton lattice-to-core integration of the Nestle nodal simulator - 331

    International Nuclear Information System (INIS)

    Galloway, J.; Hernandez, H.; Maldonado, G.I.; Jessee, M.; Popov, E.; Clarno, K.

    2010-01-01

    This article reports the status of recent and substantial enhancements made to the NESTLE nodal core simulator, a code originally developed at North Carolina State University (NCSU) of which version 5.2.1 has been available for several years through the Oak Ridge National Laboratory (ORNL) Radiation Safety Information Computational Center (RSICC) software repository. In its released and available form, NESTLE is a seasoned, well-developed and extensively tested code system particularly useful to model PWRs. In collaboration with NCSU, University of Tennessee (UT) and ORNL researchers have recently developed new enhancements for the NESTLE code, including the implementation of a two-phase drift-flux thermal hydraulic and flow redistribution model to facilitate modeling of Boiling Water Reactors (BWRs) as well as the development of an integrated coupling of SCALE/TRITON lattice physics to NESTLE so to produce an end-to-end capability for reactor simulations. These latest advancements implemented into NESTLE as well as an update of other ongoing efforts of this project are herein reported. (authors)

  6. 3D pin-by-pin power density profiles with high spatial resolution in the vicinity of a BWR control blade tip simulated with coupled neutronics/burn-up calculations

    International Nuclear Information System (INIS)

    Li, J.; Nünighoff, K.; Allelein, H.-J.

    2011-01-01

    Highlights: ► High spatial resolution neutronic and burn-up calculations of quarter BWR fuel element section. ► Coupled MCNP(X)–ORIGEN2.2 simulation using VESTA. ► Control blade history effect was taken into account. ► Determining local power excursion after instantaneous control rod movement. ► Correlation between control blade geometry and occurrence of local power excursions. - Abstract: Pellet cladding interaction (PCI) as well as pellet cladding mechanical interaction (PCMI) are well-known fuel failures in light water reactors, especially in boiling water reactors (BWR). Whereas the thermo-mechanical processes of PCI effects have been intensively investigated in the last decades, only rare information is available on the role of neutron physics. However, each power transient is primary due to neutron physics effects and thus knowledge of the neutron physical background is mandatory to better understand the occurrence of PCI effects in BWRs. This paper will focus on a study of local power excursions in a typical BWR fuel assembly during control rod movements. Burn-up and energy deposition were simulated with high spatial granularity, especially in the vicinity of the control blade tip. It could be shown, that the design of the control blade plays a dominant role for the occurrence of local power peaks while instantaneously moving down the control rod. The main result is, that the largest power peak occurs at the interface between steel handle and absorber rods. A full width half maximum (FWHM) of ±2.5 cm was observed. This means, the local power excursion due to neutron physics phenomena involve approximately five pellets. With the VESTA code coupled MCNP(X)/ORIGEN2.2 calculations were performed with more than 3400 burn-up zones in order to take history effects into account.

  7. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

  8. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  9. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  10. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant; Implementacion digital, simulacion y pruebas en MATLAB de los modelos de la linea de vapor, las turbinas y el regulador de presion de una central Nucleoelectrica tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)

    2004-07-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  11. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Suzuki, H.; Hatamiya, S.; Murase, M.

    1988-01-01

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

  12. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  13. A simulation environment for ITER PCS development

    International Nuclear Information System (INIS)

    Walker, M.L.; Ambrosino, G.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neu, G.; Raupp, G.; Treutterer, W.; Winter, A.

    2014-01-01

    Highlights: • Describes task to develop simulation tool to aid development/testing of ITER PCS. • Requirements and use cases and preliminary architecture have been delivered. • Detailed design is now being developed. • Provides overview of use cases and requirements. • Provides overview of architecture and status of development. - Abstract: A simulation environment known as the Plasma Control System Simulation Platform (PCSSP), specifically designed to support development of the ITER Plasma Control System (PCS), is currently under construction by an international team encompassing a cross-section of expertise in simulation and exception handling for plasma control. The proposed design addresses the challenging requirements of supporting the PCS design. This paper provides an overview of the PCSSP project and a discussion of some of the major features of its design. Plasma control for the ITER tokamak will be significantly more challenging than for existing fusion devices. An order of magnitude greater performance (e.g. [1,2]) is needed for some types of control, which together with limited actuator authority, implies that optimized individual controllers and nonlinear saturation logic are required. At the same time, consequences of control failure are significantly more severe, which implies a conflicting requirement for robust control. It also implies a requirement for comprehensive and robust exception handling. Coordinated control of multiple competing objectives with significant interactions, together with many shared uses of actuators to control multiple variables, implies that highly integrated control logic and shared actuator management will be required. It remains a challenge for the integrated technologies to simultaneously address these multiple and often competing requirements to be demonstrated on existing fusion devices and adapted for ITER in time to support its operational schedule. We describe ways in which the PCSSP will help address

  14. Cell survival in a simulated Mars environment

    Science.gov (United States)

    Todd, Paul; Kurk, Michael Andy; Boland, Eugene; Thomas, David

    2016-07-01

    The most ancient life forms on earth date back comfortably to the time when liquid water was believed to be abundant on Mars. These ancient life forms include cyanobacteria, contemporary autotrophic earth organisms believed to have descended from ancestors present as long as 3.5 billion years ago. Contemporary cyanobacteria have adapted to the earth environment's harshest conditions (long-term drying, high and low temperature), and, being autotrophic, they are among the most likely life forms to withstand space travel and the Mars environment. However, it is unlikely that humans would unwittingly contaminate a planetary spacecraft with these microbes. One the other hand, heterotrophic microbes that co-habit with humans are more likely spacecraft contaminants, as history attests. Indeed, soil samples from the Atacama desert have yielded colony-forming organisms resembling enteric bacteria. There is a need to understand the survivability of cyanobacteria (likely survivors, unlikely contaminants) and heterotrophic eubacteria (unlikely survivors, likely contaminants) under simulated planetary conditions. A 35-day test was performed in a commercial planetary simulation system (Techshot, Inc., Greenville, IN) in which the minimum night-time temperature was -80 C, the maximum daytime temperature was +26 C, the simulated day-night light cycle in earth hours was 12-on and 12-off, and the total pressure of the pure CO _{2} atmosphere was maintained below 11 mbar. Any water present was allowed to equilibrate with the changing temperature and pressure. The gas phase was sampled into a CR1-A low-pressure hygrometer (Buck Technologies, Boulder, CO), and dew/frost point was measured once every hour and recorded on a data logger, along with the varying temperature in the chamber, from which the partial pressure of water was calculated. According to measurements there was no liquid water present throughout the test except during the initial pump-down period when aqueous specimens

  15. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  16. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  17. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

  18. Full scope upgrade project for the Fermi 2 simulator

    International Nuclear Information System (INIS)

    Bollacasa, D.; Gonsalves, J.B.; Newcomb, P.C.

    1994-01-01

    The Detroit Edison company (DECO) concentrated the Simulation Division of Asea Brown Boveri (ABB) to perform a full scope upgrade of the Fermi 2 simulator. The Fermi 2 plant is a BWR 6 generation Nuclear Steam Supply System (NSSS). The project included the complete replacement of the existing simulation model sofware with ABB's high fidelity BWR models, addition of an advanced instructor station facility and new simulation computers. Also provided on the project were ABB's advanced simulation environment (CETRAN), a comprehensive configuration management system based on a modern relational database system and a new computer interface to the input/output system. (8 refs., 2 figs.)

  19. Computer simulation of spacecraft/environment interaction

    International Nuclear Information System (INIS)

    Krupnikov, K.K.; Makletsov, A.A.; Mileev, V.N.; Novikov, L.S.; Sinolits, V.V.

    1999-01-01

    This report presents some examples of a computer simulation of spacecraft interaction with space environment. We analysed a set data on electron and ion fluxes measured in 1991-1994 on geostationary satellite GORIZONT-35. The influence of spacecraft eclipse and device eclipse by solar-cell panel on spacecraft charging was investigated. A simple method was developed for an estimation of spacecraft potentials in LEO. Effects of various particle flux impact and spacecraft orientation are discussed. A computer engineering model for a calculation of space radiation is presented. This model is used as a client/server model with WWW interface, including spacecraft model description and results representation based on the virtual reality markup language

  20. Computer simulation of spacecraft/environment interaction

    CERN Document Server

    Krupnikov, K K; Mileev, V N; Novikov, L S; Sinolits, V V

    1999-01-01

    This report presents some examples of a computer simulation of spacecraft interaction with space environment. We analysed a set data on electron and ion fluxes measured in 1991-1994 on geostationary satellite GORIZONT-35. The influence of spacecraft eclipse and device eclipse by solar-cell panel on spacecraft charging was investigated. A simple method was developed for an estimation of spacecraft potentials in LEO. Effects of various particle flux impact and spacecraft orientation are discussed. A computer engineering model for a calculation of space radiation is presented. This model is used as a client/server model with WWW interface, including spacecraft model description and results representation based on the virtual reality markup language.

  1. Human Performance in Simulated Reduced Gravity Environments

    Science.gov (United States)

    Cowley, Matthew; Harvill, Lauren; Rajulu, Sudhakar

    2014-01-01

    NASA is currently designing a new space suit capable of working in deep space and on Mars. Designing a suit is very difficult and often requires trade-offs between performance, cost, mass, and system complexity. Our current understanding of human performance in reduced gravity in a planetary environment (the moon or Mars) is limited to lunar observations, studies from the Apollo program, and recent suit tests conducted at JSC using reduced gravity simulators. This study will look at our most recent reduced gravity simulations performed on the new Active Response Gravity Offload System (ARGOS) compared to the C-9 reduced gravity plane. Methods: Subjects ambulated in reduced gravity analogs to obtain a baseline for human performance. Subjects were tested in lunar gravity (1.6 m/sq s) and Earth gravity (9.8 m/sq s) in shirt-sleeves. Subjects ambulated over ground at prescribed speeds on the ARGOS, but ambulated at a self-selected speed on the C-9 due to time limitations. Subjects on the ARGOS were given over 3 minutes to acclimate to the different conditions before data was collected. Nine healthy subjects were tested in the ARGOS (6 males, 3 females, 79.5 +/- 15.7 kg), while six subjects were tested on the C-9 (6 males, 78.8 +/- 11.2 kg). Data was collected with an optical motion capture system (Vicon, Oxford, UK) and was analyzed using customized analysis scripts in BodyBuilder (Vicon, Oxford, UK) and MATLAB (MathWorks, Natick, MA, USA). Results: In all offloaded conditions, variation between subjects increased compared to 1-g. Kinematics in the ARGOS at lunar gravity resembled earth gravity ambulation more closely than the C-9 ambulation. Toe-off occurred 10% earlier in both reduced gravity environments compared to earth gravity, shortening the stance phase. Likewise, ankle, knee, and hip angles remained consistently flexed and had reduced peaks compared to earth gravity. Ground reaction forces in lunar gravity (normalized to Earth body weight) were 0.4 +/- 0.2 on

  2. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  3. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  4. FIST small break accident analysis with BWR TRACBO2-pretest predictions

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

  5. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  6. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  7. Hydraulic modeling and simulation of a System Division of Essential Service Water in a BWR plant with Flow master; Modelo hidraulico y simulacion de una division del Sistema de Agua de Servicio Esencial de una central BWR con Flowmaster

    Energy Technology Data Exchange (ETDEWEB)

    Vegazo Juzgado, L.; Rodriguez Garcia, G. M.; Mota Coloma, M.

    2012-07-01

    At the conclusion of the project can say that Flow master is a simulation tool that allows you to create your model from a library of components and obtain useful results from the point of view of the operation, engineering and maintenance. Compared to previous software from the point of view of use, can comment that Flow master is a tool which has an intuitive and user-friendly interaction between the user and the program thus facilitating the modeling of the system and definition of the components of same.

  8. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  9. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. EIS pitting temperature determination of A182 nickel based alloy in simulated BWR environment containing dilute seawater

    International Nuclear Information System (INIS)

    Lavigne, Olivier; Shoji, Tetsuo; Takeda, Yoichi

    2014-01-01

    Graphical abstract: - Highlights: • Stable pitting events in function of the temperature are monitored by electrochemical impedance spectroscopy. • The pitting temperature for the nickel based alloy A182 in solution containing 450 ppm Cl − is defined as above 160 °C. • The presented method can be applied for others passive alloys as stainless steel in solution containing aggressive anions. - Abstract: A method based on electrochemical impedance spectroscopy (EIS) measurements to monitor the pitting temperature of passive alloys in a given media is developed in this communication. The pitting corrosion behavior of the nickel based alloy 182 in water containing 450 ppm by weight of chloride is presented in this study. The analysis of the EIS fit parameters from the proposed equivalent electrical circuit allows to determine the temperature from which stable pitting event occurs at open circuit potential. For the A182 sample this temperature is measured above 160 °C

  13. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator; Modelado de la dinamica de la vasija y circuitos de recirculacion de una nucleoelectrica tipo BWR como parte del simulador universitario SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R A [DEPFI, Campus Morelos, en IMTA, Jiutepec, Morelos (Mexico)

    2003-07-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  14. An intelligent dynamic simulation environment: An object-oriented approach

    International Nuclear Information System (INIS)

    Robinson, J.T.; Kisner, R.A.

    1988-01-01

    This paper presents a prototype simulation environment for nuclear power plants which illustrates the application of object-oriented programming to process simulation. Systems are modeled using this technique as a collection of objects which communicate via message passing. The environment allows users to build simulation models by selecting iconic representations of plant components from a menu and connecting them with the aid of a mouse. Models can be modified graphically at any time, even as the simulation is running, and the results observed immediately via real-time graphics. This prototype illustrates the use of object-oriented programming to create a highly interactive and automated simulation environment. 9 refs., 4 figs

  15. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Uruma, Y.; Osato, T.; Yamazaki, K.

    2002-01-01

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  16. Effect of Loading Transients on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Tests 3 and 4)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2003-04-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the third and fourth test of working package (WP) 3 with loading transients, performed at Paul Scherrer Institut (PSI). Two different low-alloy steels (20 MnMoNi 5 5, 0.015 wt.% S and 22 NiMoCr 3 7, 0.007 wt. %S) were investigated in oxygenated high-temperature, high-purity water (T = 240 {sup o}C, DO = 400 ppb) in a daisy chain at two different load ratios (R = 0.8 and 0.2). In the first part of the experiments, asymmetrical saw tooth loading with different rise times {delta}t{sub R} of the load and different loading frequencies were applied. Then the loading conditions were changed to an asymmetrical trapezoid waveform loading (periodical partial unloading, PPU) and the hold time {delta}t{sub H} at maximum load was varied. In the final phase of WP 3 PSI tests 3 and 4 the SCC behaviour was investigated under constant load. With decreasing loading frequency the corrosion fatigue (CF) crack advance per cycle {delta}a/{delta}N{sub EAC} of material A increased. Sustained EAC crack growth could be maintained down to low frequencies of 10{sup -5} Hz. The time-based crack growth rate (CGR) da/dt{sub EAC} decreased with decreasing frequency. In material B no effect of the loading frequency could be resolved. Up to a hold time of 1 h at maximum constant load the CGR da/dt{sub EAC} seemed to be independent of the hold time. Above hold times of 1 h the CGR decreased and dropped down to CGR values in the range or below the BWR VIP 60 SCC disposition lines. This behaviour was observed in both investigated materials. The cycle-based CGR {delta}a/{delta}N{sub EAC} remained approximately constant with increasing hold time. The

  17. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  18. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  19. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  20. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  1. Virtual Collaborative Simulation Environment for Integrated Product and Process Development

    Science.gov (United States)

    Gulli, Michael A.

    1997-01-01

    Deneb Robotics is a leader in the development of commercially available, leading edge three- dimensional simulation software tools for virtual prototyping,, simulation-based design, manufacturing process simulation, and factory floor simulation and training applications. Deneb has developed and commercially released a preliminary Virtual Collaborative Engineering (VCE) capability for Integrated Product and Process Development (IPPD). This capability allows distributed, real-time visualization and evaluation of design concepts, manufacturing processes, and total factory and enterprises in one seamless simulation environment.

  2. Using IMPRINT to Guide Experimental Design with Simulated Task Environments

    Science.gov (United States)

    2015-06-18

    USING IMPRINT TO GUIDE EXPERIMENTAL DESIGN OF SIMULATED TASK ENVIRONMENTS THESIS Gregory...ENG-MS-15-J-052 USING IMPRINT TO GUIDE EXPERIMENTAL DESIGN WITH SIMULATED TASK ENVIRONMENTS THESIS Presented to the Faculty Department...Civilian, USAF June 2015 DISTRIBUTION STATEMENT A. APPROVED FOR PUBLIC RELEASE; DISTRIBUTION UNLIMITED. AFIT-ENG-MS-15-J-052 USING IMPRINT

  3. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  4. Development of a Smart Grid Simulation Environment

    OpenAIRE

    Delamare, J; Bitachon, B.; Peng, Z.; Wang, Y.; Haverkort, Boudewijn R.H.M.; Jongerden, M.R.

    2015-01-01

    With the increased integration of renewable energy sources the interaction between energy producers and consumers has become a bi-directional exchange. Therefore, the electrical grid must be adapted into a smart grid which effectively regulates this two-way interaction. With the aid of simulation, stakeholders can obtain information on how to properly develop and control the smart grid. In this paper, we present the development of an integrated smart grid simulation model, using the Anylogic ...

  5. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

  6. CHAVIR: A virtual site simulation environment

    International Nuclear Information System (INIS)

    Leservot, Arnauld; Chodorge, Laurent

    2006-01-01

    In nuclear field, any companies involved in the management and/or the design and performance of an intervention aim at preparing it, by finding the most appropriate scenario(s) under several needs: - Technical requirements: feasibility, kind of means to engage, operating modes, tasks scheduling; - economical requirements: global mission cost minimization; - Environmental requirements: take into account the individual and collective dose rate received by the human operators involved in the intervention(s), according to the ALARA principle. Today, they also must answer complex questions to design their interventions with increasing reactivity and always lowering costs. Besides, they must be brought to answer unexpected situations during the effective realization of their nuclear interventions, and naturally to consolidate their experience feedback of the missions. An interesting way to help them in these different needs consists in taking advantage of simulation. The paper has the following contents: - Introduction; - CHAVIR project; - Goal; - Simulation and virtual reality; - Strategy; - Interactive dose evaluation; - Requirements; - Physical algorithm; - Objects representation; - Calculation optimization; - Interactive mechanical simulation; - First study cases; - Conclusion - prospects. To summarize, the authors succeeded in developing a software simulation tool, helping the users from nuclear field to prepare their interventions. CHAVIR allows interactive evaluation of dose rate, when taking into account real industrial models coming from CAD world. One can also perform mechanical simulations, to address accessibilities issues and design scenario involving either manual tasks of robotic interventions. CHAVIR is already entered the industrialization process. It aims at becoming shortly a commercial software tool for dismantling site simulation, adapted to the professional needs in order to respect the ALARA principle. It should efficiently contribute to optimize

  7. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

    The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  8. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor

    International Nuclear Information System (INIS)

    Sandoval V, S.

    2006-01-01

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

  9. Development of a Smart Grid Simulation Environment

    NARCIS (Netherlands)

    Delamare, J; Bitachon, B.; Peng, Z.; Wang, Y.; Haverkort, Boudewijn R.H.M.; Jongerden, M.R.

    2015-01-01

    With the increased integration of renewable energy sources the interaction between energy producers and consumers has become a bi-directional exchange. Therefore, the electrical grid must be adapted into a smart grid which effectively regulates this two-way interaction. With the aid of simulation,

  10. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  11. Cloning simulation in the cage environment.

    OpenAIRE

    Douthart, R J; Thomas, J J; Rosier, S D; Schmaltz, J E; West, J W

    1986-01-01

    The CAGE/GEM(TM) software toolkit for genetic engineering is briefly described. The system functionally uses color graphics and is menu driven. It integrates genetics and features information ("Overlays") with information based on sequence analysis ("Representations"). The system is structured around CAD (Computer Aided Design) principles. The CAGE (Computer Aided Genetic Engineering) aspects of the software are emphasized and illustrated by a simulated cloning of the hepatitis B core antigen...

  12. Simulating tumor growth in confined heterogeneous environments

    International Nuclear Information System (INIS)

    Gevertz, Jana L; Torquato, Salvatore; Gillies, George T

    2008-01-01

    The holy grail of computational tumor modeling is to develop a simulation tool that can be utilized in the clinic to predict neoplastic progression and propose individualized optimal treatment strategies. In order to develop such a predictive model, one must account for many of the complex processes involved in tumor growth. One interaction that has not been incorporated into computational models of neoplastic progression is the impact that organ-imposed physical confinement and heterogeneity have on tumor growth. For this reason, we have taken a cellular automaton algorithm that was originally designed to simulate spherically symmetric tumor growth and generalized the algorithm to incorporate the effects of tissue shape and structure. We show that models that do not account for organ/tissue geometry and topology lead to false conclusions about tumor spread, shape and size. The impact that confinement has on tumor growth is more pronounced when a neoplasm is growing close to, versus far from, the confining boundary. Thus, any clinical simulation tool of cancer progression must not only consider the shape and structure of the organ in which a tumor is growing, but must also consider the location of the tumor within the organ if it is to accurately predict neoplastic growth dynamics

  13. A Collaborative Extensible User Environment for Simulation and Knowledge Management

    Energy Technology Data Exchange (ETDEWEB)

    Freedman, Vicky L.; Lansing, Carina S.; Porter, Ellen A.; Schuchardt, Karen L.; Guillen, Zoe C.; Sivaramakrishnan, Chandrika; Gorton, Ian

    2015-06-01

    In scientific simulation, scientists use measured data to create numerical models, execute simulations and analyze results from advanced simulators executing on high performance computing platforms. This process usually requires a team of scientists collaborating on data collection, model creation and analysis, and on authorship of publications and data. This paper shows that scientific teams can benefit from a user environment called Akuna that permits subsurface scientists in disparate locations to collaborate on numerical modeling and analysis projects. The Akuna user environment is built on the Velo framework that provides both a rich client environment for conducting and analyzing simulations and a Web environment for data sharing and annotation. Akuna is an extensible toolset that integrates with Velo, and is designed to support any type of simulator. This is achieved through data-driven user interface generation, use of a customizable knowledge management platform, and an extensible framework for simulation execution, monitoring and analysis. This paper describes how the customized Velo content management system and the Akuna toolset are used to integrate and enhance an effective collaborative research and application environment. The extensible architecture of Akuna is also described and demonstrates its usage for creation and execution of a 3D subsurface simulation.

  14. Applying virtual environments to training and simulation (abstract)

    NARCIS (Netherlands)

    Jense, G.J.; Kuijper, F.

    1993-01-01

    Virtual environment (VE) technology is expected to make a big impact on future training and simulation systems. Direct stimulation of human-senses (eyesight, auditory, tactile) and new paradigms for user input will improve the realism of simulations and thereby the effectiveness of training systems.

  15. NECTAR: Simulation and Visualization in a 3D Collaborative Environment

    NARCIS (Netherlands)

    Law, Y.W.; Chan, K.Y.

    For simulation and visualization in a 3D collaborative environment, an architecture called the Nanyang Experimental CollaboraTive ARchitecture (NECTAR) has been developed. The objective is to support multi-user collaboration in a virtual environment with an emphasis on cost-effectiveness and

  16. Simulation environment for algorithms and agents evaluation.

    Directory of Open Access Journals (Sweden)

    Pablo CHAMOSO

    2016-06-01

    Full Text Available This article presents an adaptive platform that can simulate the centralized control of different smart city areas. For example, public lighting and intelligent management, public zones of buildings, energy distribution, etc. It can operate the hardware infrastructure and perform optimization both in energy consumption and economic control from a modular architecture which is fully adaptable to most cities. Machine-to-machine (M2M permits connecting all the sensors of the city so that they provide the platform with a perfect perspective of the global city status. To carry out this optimization, the platform offers the developers a software that operates on the hardware infrastructure and merges various techniques of artificial intelligence (AI and statistics, such as artificial neural networks (ANN, multi-agent systems (MAS or a Service Oriented Approach (SOA, forming an Internet of Services (IoS. Different case studies were tested by using the presented platform, and further development is still underway with additional case studies.

  17. Environments for online maritime simulators with cloud computing capabilities

    Science.gov (United States)

    Raicu, Gabriel; Raicu, Alexandra

    2016-12-01

    This paper presents the cloud computing environments, network principles and methods for graphical development in realistic naval simulation, naval robotics and virtual interactions. The aim of this approach is to achieve a good simulation quality in large networked environments using open source solutions designed for educational purposes. Realistic rendering of maritime environments requires near real-time frameworks with enhanced computing capabilities during distance interactions. E-Navigation concepts coupled with the last achievements in virtual and augmented reality will enhance the overall experience leading to new developments and innovations. We have to deal with a multiprocessing situation using advanced technologies and distributed applications using remote ship scenario and automation of ship operations.

  18. A Simulated Learning Environment for Teaching Medicine Dispensing Skills.

    Science.gov (United States)

    McDowell, Jenny; Styles, Kim; Sewell, Keith; Trinder, Peta; Marriott, Jennifer; Maher, Sheryl; Naidu, Som

    2016-02-25

    To develop an authentic simulation of the professional practice dispensary context for students to develop their dispensing skills in a risk-free environment. A development team used an Agile software development method to create MyDispense, a web-based simulation. Modeled on virtual learning environments elements, the software employed widely available standards-based technologies to create a virtual community pharmacy environment. Assessment. First-year pharmacy students who used the software in their tutorials, were, at the end of the second semester, surveyed on their prior dispensing experience and their perceptions of MyDispense as a tool to learn dispensing skills. The dispensary simulation is an effective tool for helping students develop dispensing competency and knowledge in a safe environment.

  19. An intelligent simulation environment for control system design

    International Nuclear Information System (INIS)

    Robinson, J.T.

    1989-01-01

    The Oak Ridge National Laboratory is currently assisting in the development of advanced control systems for the next generation of nuclear power plants. This paper presents a prototype interactive and intelligent simulation environment being developed to support this effort. The environment combines tools from the field of Artificial Intelligence; in particular object-oriented programming, a LISP programming environment, and a direct manipulation user interface; with traditional numerical methods for simulating combined continuous/discrete processes. The resulting environment is highly interactive and easy to use. Models may be created and modified quickly through a window oriented direct manipulation interface. Models may be modified at any time, even as the simulation is running, and the results observed immediately via real-time graphics. 8 refs., 3 figs

  20. Development of long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-01-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  1. Results of modeling advanced BWR fuel designs using CASMO-4

    International Nuclear Information System (INIS)

    Knott, D.; Edenius, M.

    1996-01-01

    Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

  2. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  3. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  4. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  5. Development of next BWR plant

    International Nuclear Information System (INIS)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

    1995-01-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

  6. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1995-04-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

  7. Conducting Simulation Studies in the R Programming Environment.

    Science.gov (United States)

    Hallgren, Kevin A

    2013-10-12

    Simulation studies allow researchers to answer specific questions about data analysis, statistical power, and best-practices for obtaining accurate results in empirical research. Despite the benefits that simulation research can provide, many researchers are unfamiliar with available tools for conducting their own simulation studies. The use of simulation studies need not be restricted to researchers with advanced skills in statistics and computer programming, and such methods can be implemented by researchers with a variety of abilities and interests. The present paper provides an introduction to methods used for running simulation studies using the R statistical programming environment and is written for individuals with minimal experience running simulation studies or using R. The paper describes the rationale and benefits of using simulations and introduces R functions relevant for many simulation studies. Three examples illustrate different applications for simulation studies, including (a) the use of simulations to answer a novel question about statistical analysis, (b) the use of simulations to estimate statistical power, and (c) the use of simulations to obtain confidence intervals of parameter estimates through bootstrapping. Results and fully annotated syntax from these examples are provided.

  8. Corrosion fatigue crack growth behaviour of low-alloy RPV steels at different temperatures and loading frequencies under BWR/NWC environment

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.

    2004-01-01

    The strain-induced corrosion cracking or low-frequency corrosion fatigue (LFCF) crack growth behaviour of different reactor pressure vessel (RPV) steels and of a RPV weld filler/weld heat-affected zone (HAZ) material were characterized under simulated transient boiling water reactor/normal water chemistry conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated high-temperature water at temperatures of either 288, 250, 200, or 150 deg. C. Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographic analysis by SEM were used to quantify the cracking response. Under low-flow and highly oxidising conditions (ECP > 0 mV SHE , O 2 = 0.4 ppm) the cycle-based LFCF crack growth rates (CGR) Δa/ΔN increased with decreasing loading frequency and increasing temperature with a maximum/plateau at/above 250 deg. C. Sustained environmentally-assisted crack growth could be maintained down to low frequencies of 10 -5 Hz. The LFCF CGR of low- and high-sulphur steels and of the weld filler/HAZ material were comparable over a wide range of loading conditions and conservatively covered by the 'high-sulphur line' of the General Electric-model. The 'ASME XI wet fatigue CGR curves' could be significantly exceeded in all materials by cyclic fatigue loading at low frequencies ( -2 Hz) at high and low load ratios R. (authors)

  9. Conducting Simulation Studies in the R Programming Environment

    Directory of Open Access Journals (Sweden)

    Kevin A. Hallgren

    2013-10-01

    Full Text Available Simulation studies allow researchers to answer specific questions about data analysis, statistical power, and best-practices for obtainingaccurate results in empirical research. Despite the benefits that simulation research can provide, many researchers are unfamiliar with available tools for conducting their own simulation studies. The use of simulation studies need not be restricted toresearchers with advanced skills in statistics and computer programming, and such methods can be implemented by researchers with a variety of abilities and interests. The present paper provides an introduction to methods used for running simulationstudies using the R statistical programming environment and is written for individuals with minimal experience running simulation studies or using R. The paper describes the rationale and benefits of using simulations and introduces R functions relevant for many simulation studies. Three examples illustrate different applications for simulation studies, including (a the use of simulations to answer a novel question about statistical analysis, (b the use of simulations to estimate statistical power, and (c the use of simulations to obtain confidence intervals of parameter estimates throughbootstrapping. Results and fully annotated syntax from these examples are provided.

  10. Simulation based virtual learning environment in medical genetics counseling

    DEFF Research Database (Denmark)

    Makransky, Guido; Bonde, Mads T.; Wulff, Julie S. G.

    2016-01-01

    BACKGROUND: Simulation based learning environments are designed to improve the quality of medical education by allowing students to interact with patients, diagnostic laboratory procedures, and patient data in a virtual environment. However, few studies have evaluated whether simulation based...... the perceived relevance of medical educational activities. The results suggest that simulations can help future generations of doctors transfer new understanding of disease mechanisms gained in virtual laboratory settings into everyday clinical practice....... learning environments increase students' knowledge, intrinsic motivation, and self-efficacy, and help them generalize from laboratory analyses to clinical practice and health decision-making. METHODS: An entire class of 300 University of Copenhagen first-year undergraduate students, most with a major...

  11. A simulation and training environment for robotic radiosurgery

    Energy Technology Data Exchange (ETDEWEB)

    Schlaefer, Alexander [University of Luebeck, Institute for Robotics and Cognitive Systems, Luebeck (Germany); Stanford University, Department of Radiation Oncology, Stanford, CA (United States); Gill, Jakub; Schweikard, Achim [University of Luebeck, Institute for Robotics and Cognitive Systems, Luebeck (Germany)

    2008-09-15

    To provide a software environment for simulation of robotic radiosurgery, particularly to study the effective robot workspace with respect to the treatment plan quality, and to illustrate the concepts of robotic radiosurgery. A simulation environment for a robotic radiosurgery system was developed using Java and Java3D. The kinematics and the beam characteristics were modeled and linked to a treatment planning module. Simulations of different robot workspace parameters for two example radiosurgical patient cases were performed using the novel software tool. The first case was an intracranial lesion near the left inner ear, the second case was a spinal lesion. The planning parameters for both cases were visualized with the novel simulation environment. An incremental extension of the robot workspace had limited effect for the intracranial case, where the original workspace already covered the left side of the patient. For the spinal case, a larger workspace resulted in a noticeable improvement in plan quality and a large portion of the beams being delivered from the extended workspace. The new software environment is useful to simulate and analyze parameters and configurations for robotic radiosurgery. An enlarged robot workspace may result in improved plan quality depending on the location of the target region. (orig.)

  12. A simulation and training environment for robotic radiosurgery

    International Nuclear Information System (INIS)

    Schlaefer, Alexander; Gill, Jakub; Schweikard, Achim

    2008-01-01

    To provide a software environment for simulation of robotic radiosurgery, particularly to study the effective robot workspace with respect to the treatment plan quality, and to illustrate the concepts of robotic radiosurgery. A simulation environment for a robotic radiosurgery system was developed using Java and Java3D. The kinematics and the beam characteristics were modeled and linked to a treatment planning module. Simulations of different robot workspace parameters for two example radiosurgical patient cases were performed using the novel software tool. The first case was an intracranial lesion near the left inner ear, the second case was a spinal lesion. The planning parameters for both cases were visualized with the novel simulation environment. An incremental extension of the robot workspace had limited effect for the intracranial case, where the original workspace already covered the left side of the patient. For the spinal case, a larger workspace resulted in a noticeable improvement in plan quality and a large portion of the beams being delivered from the extended workspace. The new software environment is useful to simulate and analyze parameters and configurations for robotic radiosurgery. An enlarged robot workspace may result in improved plan quality depending on the location of the target region. (orig.)

  13. Visual simulation study of equipment maintenance in dangerous environment

    International Nuclear Information System (INIS)

    Zhu Bo; Yang Yanhua; Li Shiting

    2010-01-01

    The maintenance characteristics in dangerous environments are analyzed, and the application characteristics of visualized maintenance technology are introduced. The interactive method to implement maintenance simulation is presented using EON simulation platform. Then an interacted Virtual Maintenance Training System (VMTS) is further developed, and the composition and function are described in details. The VMTS can be used in extensive array of application scopes, and it is well compatible to the hardware of virtual reality. (author)

  14. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Vargas O, D.; Chavez M, C.

    2012-10-01

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  15. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  16. BWR alloy 182 stress Corrosion Cracking Experience

    International Nuclear Information System (INIS)

    Horn, R.M.; Hickling, J.

    2002-01-01

    Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

  17. IMPETUS - Interactive MultiPhysics Environment for Unified Simulations.

    Science.gov (United States)

    Ha, Vi Q; Lykotrafitis, George

    2016-12-08

    We introduce IMPETUS - Interactive MultiPhysics Environment for Unified Simulations, an object oriented, easy-to-use, high performance, C++ program for three-dimensional simulations of complex physical systems that can benefit a large variety of research areas, especially in cell mechanics. The program implements cross-communication between locally interacting particles and continuum models residing in the same physical space while a network facilitates long-range particle interactions. Message Passing Interface is used for inter-processor communication for all simulations. Copyright © 2016 Elsevier Ltd. All rights reserved.

  18. A SIMULATION ENVIRONMENT FOR AUTOMATIC NIGHT DRIVING AND VISUAL CONTROL

    OpenAIRE

    Arroyo Rubio, Fernando

    2012-01-01

    This project consists on developing an automatic night driving system in a simulation environment. The simulator I have used is TORCS. TORCS is an Open Source car racing simulator written in C++. It is used as an ordinary car racing game, as a IA racing game and as a research platform. The goal of this thesis is to implement an automatic driving system to control the car under night conditions using computer vision. A camera is implemented inside the vehicle and it will detect the reflective ...

  19. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  20. Large bundle BWR test CORA-18: Test results

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

    1998-04-01

    The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

  1. Experimental investigation of cooling by top spray and bottom flooding of a simulated 64 rod bundle for a BWR. Pt. 2. Main experiment with modified test section

    International Nuclear Information System (INIS)

    Nilsson, L.; Gustafson, L.; Harju, R.

    1978-06-01

    The cooling of an electrically heated, full scale 64-rod bundle has been investigated under simulated emergency core cooling conditions. Emphasis was laid on measurements of rod cladding and canister temperatures. By means of difference pressure measurements the levels in bundle, by-pass and downcomer could be estimated and thus the effective reflooding velocity. The test section was modified compared to the pre-tests, in order to improve system effects simulation. A new rod bundle was installed including a hollow, water, rod and 63 indirectly heated rods. Parameter effects of coolant mass flow rate and distribution, initial cladding temperature, pressure and power were studied. The effect of the way the test section was vented was also investigated and turned out to be very significant. (author)

  2. MARS: An Educational Environment for Multiagent Robot Simulations

    Directory of Open Access Journals (Sweden)

    Marco Casini

    2016-01-01

    Full Text Available Undergraduate robotics students often find it difficult to design and validate control algorithms for teams of mobile robots. This is mainly due to two reasons. First, very rarely, educational laboratories are equipped with large teams of robots, which are usually expensive, bulky, and difficult to manage and maintain. Second, robotics simulators often require students to spend much time to learn their use and functionalities. For this purpose, a simulator of multiagent mobile robots named MARS has been developed within the Matlab environment, with the aim of helping students to simulate a wide variety of control algorithms in an easy way and without spending time for understanding a new language. Through this facility, the user is able to simulate multirobot teams performing different tasks, from cooperative to competitive ones, by using both centralized and distributed controllers. Virtual sensors are provided to simulate real devices. A graphical user interface allows students to monitor the robots behaviour through an online animation.

  3. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C.

    2012-10-01

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

  4. POLCA-T simulation of OECD/NRC BWR turbine trip benchmark exercise 3 best estimate scenario TT2 test and four extreme scenarios

    International Nuclear Information System (INIS)

    Panayotov, D.

    2004-01-01

    Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and the 3D neutron kinetics core model. Code validation plan includes the calculations of Peach Bottom end of cycle 2 turbine trip transients and low-flow stability tests. The paper describes the objectives, method, and results of analyses performed in the final phase of OECD/NRC Peach Bottom 2 Boiling Water Reactor Turbine Trip Benchmark. Brief overview of the code features, the method of simulation, the developed 3D core model and system input deck for Peach Bottom 2 are given. The paper presents the results of benchmark exercise 3 best estimate scenario: coupled 3D core neutron kinetics with system thermal-hydraulics analyses. Performed sensitivity studies cover the SCRAM initiation, carry-under, and decay power. Obtained results including total power, steam dome, core exit, lower and upper plenum, main steam line and turbine inlet pressures showed good agreement with measured plant data Thus the POLCA-T code capabilities for correct simulation of turbine trip transients were proved The performed calculations and obtained results for extreme cases demonstrate the POLCA-T code wide range capabilities to simulate transients when scram, steam bypass, and safety and relief valves are not activated. The code is able to handle such transients even when the reactor power and pressure reach values higher than 600 % of rated power, and 10.8 MPa. (authors)

  5. Open Source Power Plant Simulator Development Under Matlab Environment

    International Nuclear Information System (INIS)

    Ratemi, W.M.; Fadilah, S.M.; Abonoor, N

    2008-01-01

    In this paper an open source programming approach is targeted for the development of power plant simulator under Matlab environment. With this approach many individuals can contribute to the development of the simulator by developing different orders of complexities of the power plant components. Such modules can be modeled based on physical principles, or using neural networks or other methods. All of these modules are categorized in Matlab library, of which the user can select and build up his simulator. Many international companies developed its own authoring tool for the development of its simulators, and hence it became its own property available for high costs. Matlab is a general software developed by mathworks that can be used with its toolkits as the authoring tool for the development of components by different individuals, and through the appropriate coordination, different plant simulators, nuclear, traditional , or even research reactors can be computerly assembled. In this paper, power plant components such as a pressurizer, a reactor, a steam generator, a turbine, a condenser, a feedwater heater, a valve, a pump are modeled based on physical principles. Also a prototype modeling of a reactor ( a scram case) based on neural networks is developed. These modules are inserted in two different Matlab libraries one called physical and the other is called neural. Furthermore, during the simulation one can pause and shuffle the modules selected from the two libraries and then proceed the simulation. Also, under the Matlab environment a PID controller is developed for multi-loop plant which can be integrated for the control of the appropriate developed simulator. This paper is an attempt to base the open source approach for the development of power plant simulators or even research reactor simulators. It then requires the coordination among interested individuals or institutions to set it to professionalism. (author)

  6. TACOP : A cognitive agent for a naval training simulation environment

    NARCIS (Netherlands)

    Doesburg, W.A. van; Heuvelink, A.; Broek, E.L. van den

    2005-01-01

    This paper describes how cognitive modeling can be exploited in the design of software agents that support naval training sessions. The architecture, specifications, and embedding of the cognitive agent in a simulation environment are described. Subsequently, the agent's functioning was evaluated in

  7. TACOP: A Cognitive Agent for a Naval Training Simulation Environment

    NARCIS (Netherlands)

    van Doesburg, W.A.; Verbeeck, K.; Heuvelink, A.; Tuyls, K.; Nowé, A.; van den Broek, Egon; Manderick, B.; Kuijpers, B.

    2005-01-01

    The full version of this paper appeared in: Doesburg, W. A. van, Heuvelink, A., and Broek, E. L. van den (2005). TACOP: A cognitive agent for a naval training simulation environment. In M. Pechoucek, D. Steiner, and S. Thompson (Eds.), Proceedings of the Industry Track of the Fourth International

  8. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  9. BWR steel containment corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Tan, C.P.; Bagchi, G.

    1996-04-01

    The report describes regulatory actions taken after corrosion was discovered in the drywell at the Oyster Creek Plant and in the torus at the Nine Mile Point 1 Plant. The report describes the causes of corrosion, requirements for monitoring corrosion, and measures to mitigate the corrosive environment for the two plants. The report describes the issuances of generic letters and information notices either to collect information to determine whether the problem is generic or to alert the licensees of similar plants about the existence of such a problem. Implementation of measures to enhance the containment performance under severe accident conditions is discussed. A study by Brookhaven National Laboratory (BNL) of the performance of a degraded containment under severe accident conditions is summarized. The details of the BNL study are in the appendix to the report.

  10. Status Report of Simulated Space Radiation Environment Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Phil Hyun; Nho, Young Chang; Jeun, Joon Pyo; Choi, Jae Hak; Lim, Youn Mook; Jung, Chan Hee; Jeon, Young Kyu

    2007-11-15

    The technology for performance testing and improvement of materials which are durable at space environment is a military related technology and veiled and securely regulated in advanced countries such as US and Russia. This core technology cannot be easily transferred to other country too. Therefore, this technology is the most fundamental and necessary research area for the successful establishment of space environment system. Since the task for evaluating the effects of space materials and components by space radiation plays important role in satellite lifetime extension and running failure percentage decrease, it is necessary to establish simulated space radiation facility and systematic testing procedure. This report has dealt with the status of the technology to enable the simulation of space environment effects, including the effect of space radiation on space materials. This information such as the fundamental knowledge of space environment and research status of various countries as to the simulation of space environment effects of space materials will be useful for the research on radiation hardiness of the materials. Furthermore, it will be helpful for developer of space material on deriving a better choice of materials, reducing the design cycle time, and improving safety.

  11. Status Report of Simulated Space Radiation Environment Facility

    International Nuclear Information System (INIS)

    Kang, Phil Hyun; Nho, Young Chang; Jeun, Joon Pyo; Choi, Jae Hak; Lim, Youn Mook; Jung, Chan Hee; Jeon, Young Kyu

    2007-11-01

    The technology for performance testing and improvement of materials which are durable at space environment is a military related technology and veiled and securely regulated in advanced countries such as US and Russia. This core technology cannot be easily transferred to other country too. Therefore, this technology is the most fundamental and necessary research area for the successful establishment of space environment system. Since the task for evaluating the effects of space materials and components by space radiation plays important role in satellite lifetime extension and running failure percentage decrease, it is necessary to establish simulated space radiation facility and systematic testing procedure. This report has dealt with the status of the technology to enable the simulation of space environment effects, including the effect of space radiation on space materials. This information such as the fundamental knowledge of space environment and research status of various countries as to the simulation of space environment effects of space materials will be useful for the research on radiation hardiness of the materials. Furthermore, it will be helpful for developer of space material on deriving a better choice of materials, reducing the design cycle time, and improving safety

  12. Influence of sulphate ions on the composition and structure of the oxide films on stainless steel and nickel alloys in simulated BWR crack conditions

    International Nuclear Information System (INIS)

    Bojinov, M.; Kinnunen, P.; Laitinen, E.; Maekelae, K.; Saario, T.; Sirkiae, P.; Toivonen, A.; Campbell, J.M.; Johansson, L.S.; Helin, M.; Muttilainen, E.; Reinvall, A.; Ollonqvist, T.; Vaeyrynen, J.

    2002-01-01

    The goal of the present work has been to clarify the influence of sulphate ions on the oxide films formed on stainless steel and Ni-based alloys in simulated crack chemistry conditions using different ex situ analytical techniques. The main observations of this work can be summarised as follows: The thickness of the films formed in simulated oxygen-free crack chemistry conditions during an exposure of circa 4 days varies roughly in the range 200..500 nm, which corresponds to observations reported in the literature [2]. The presence of 10000 ppb sulphate ions in simulated crack tip conditions seems to lead to a considerably lower thickness of the oxide films when compared to sulphate-free conditions. The presence of 10000 ppb sulphate ions leads also to considerable changes in the morphology of the oxide crystals on the material samples. In the absence of sulphate the outer oxide layer contains elongated round-edged crystals, while in the presence of sulphate ions the crystals are longish and needle-like. No visible difference can be observed in the outlook of the crystals formed on stainless steel and Inconel alloy surfaces. A small amount of sulphur in the form of sulphate can be found on the oxide surface on all the studied materials after exposure to the 10000 ppb solution. Sulphur seems to become incorporated inside the oxide film on AISI 316 L(NG). It is not clear at this stage, whether the observed influence of the sulphate ions can be ascribed to the lower pH, to a possible effect on solubility or to a direct influence of the anionic species. (authors)

  13. Classroom Simulation for Trainee Teachers Using 3D Virtual Environments and Simulated Smartbot Student Behaviours

    OpenAIRE

    Alotaibi, Fahad Mazaed

    2014-01-01

    his thesis consists of an analysis of a classroom simulation using a Second Life (SL) experiment that aims to investigate the teaching impact on smartbots (virtual students) from trainee teacher avatars with respect to interaction, simulated behaviour, and observed teaching roles. The classroom-based SL experiments’ motivation is to enable the trainee teacher to acquire the necessary skills and experience to manage a real classroom environment through simulations of a real classroom. This ty...

  14. A Multiagent Modeling Environment for Simulating Work Practice in Organizations

    Science.gov (United States)

    Sierhuis, Maarten; Clancey, William J.; vanHoof, Ron

    2004-01-01

    In this paper we position Brahms as a tool for simulating organizational processes. Brahms is a modeling and simulation environment for analyzing human work practice, and for using such models to develop intelligent software agents to support the work practice in organizations. Brahms is the result of more than ten years of research at the Institute for Research on Learning (IRL), NYNEX Science & Technology (the former R&D institute of the Baby Bell telephone company in New York, now Verizon), and for the last six years at NASA Ames Research Center, in the Work Systems Design and Evaluation group, part of the Computational Sciences Division (Code IC). Brahms has been used on more than ten modeling and simulation research projects, and recently has been used as a distributed multiagent development environment for developing work practice support tools for human in-situ science exploration on planetary surfaces, in particular a human mission to Mars. Brahms was originally conceived of as a business process modeling and simulation tool that incorporates the social systems of work, by illuminating how formal process flow descriptions relate to people s actual located activities in the workplace. Our research started in the early nineties as a reaction to experiences with work process modeling and simulation . Although an effective tool for convincing management of the potential cost-savings of the newly designed work processes, the modeling and simulation environment was only able to describe work as a normative workflow. However, the social systems, uncovered in work practices studied by the design team played a significant role in how work actually got done-actual lived work. Multi- tasking, informal assistance and circumstantial work interactions could not easily be represented in a tool with a strict workflow modeling paradigm. In response, we began to develop a tool that would have the benefits of work process modeling and simulation, but be distinctively able to

  15. Generic Simulator Environment for Realistic Simulation - Autonomous Entity Proof and Emotion in Decision Making

    Directory of Open Access Journals (Sweden)

    Mickaël Camus

    2004-10-01

    Full Text Available Simulation is usually used as an evaluation and testing system. Many sectors are concerned such as EUROPEAN SPACE AGENCY or the EUROPEAN DEFENCE. It is important to make sure that the project is error-free in order to continue it. The difficulty is to develop a realistic environment for the simulation and the execution of a scenario. This paper presents PALOMA, a Generic Simulator Environment. This project is based essantially on the Chaos Theory and Complex Systems to create and direct an environment for a simulation. An important point is the generic aspect. PALOMA will be able to create an environment for different sectors (Aero-space, Biology, Mathematic, .... PALOMA includes six components : the Simulation Engine, the Direction Module, the Environment Generator, the Natural Behavior Restriction, the Communication API and the User API. Three languages are used to develop this simulator. SCHEME for the Direction language, C/C++ for the development of modules and OZ/MOZART for the heart of PALOMA.

  16. Simulation Environment Synchronizing Real Equipment for Manufacturing Cell

    Science.gov (United States)

    Inukai, Toshihiro; Hibino, Hironori; Fukuda, Yoshiro

    Recently, manufacturing industries face various problems such as shorter product life cycle, more diversified customer needs. In this situation, it is very important to reduce lead-time of manufacturing system constructions. At the manufacturing system implementation stage, it is important to make and evaluate facility control programs for a manufacturing cell, such as ladder programs for programmable logical controllers (PLCs) rapidly. However, before the manufacturing systems are implemented, methods to evaluate the facility control programs for the equipment while mixing and synchronizing real equipment and virtual factory models on the computers have not been developed. This difficulty is caused by the complexity of the manufacturing system composed of a great variety of equipment, and stopped precise and rapid support of a manufacturing engineering process. In this paper, a manufacturing engineering environment (MEE) to support manufacturing engineering processes using simulation technologies is proposed. MEE consists of a manufacturing cell simulation environment (MCSE) and a distributed simulation environment (DSE). MCSE, which consists of a manufacturing cell simulator and a soft-wiring system, is emphatically proposed in detail. MCSE realizes making and evaluating facility control programs by using virtual factory models on computers before manufacturing systems are implemented.

  17. A virtual environment for simulation of radiological accidents

    International Nuclear Information System (INIS)

    Silva, Tadeu Augusto de Almeida; Farias, Oscar Luiz Monteiro de

    2013-01-01

    A virtual environment is a computer environment, representative of a subset of the real world, and where models of the real world entities, process and events are included in a virtual (three-dimensional) space. Virtual environments are ideal tools for simulation of certain critical processes, such as radiological accidents, where human beings or properties can suffer irreversible or long term damages. Radiological accidents are characterized by the significant exposure to radiation of specialized workers and general public. The early detection of a radiological accident and the determination of its possible extension are essential factors for the planning of prompt answers and emergency actions. This paper proposes the integration of georeferenced representation of the three-dimensional space and agent-based models, with the objective to construct virtual environments that have the capacity to simulate radiological accidents. The three-dimensional georeferenced representations of space candidates are: 1) the spatial representation of traditional geographical information systems (GIS); 2) the representation adopted by Google Maps®. Adding agents to these spatial representations allow us to simulate radiological accidents, quantify the doses received by members of the public, obtain a possible spatial distribution of people contaminated, estimate the number of contaminated individuals, estimate the impact on the health-network, estimate environmental impacts, generate exclusion zones, build alternative scenarios and train staff to deal with radiological accidents. (author)

  18. A virtual environment for simulation of radiological accidents

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Tadeu Augusto de Almeida, E-mail: tedsilva@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Farias, Oscar Luiz Monteiro de, E-mail: fariasol@eng.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    A virtual environment is a computer environment, representative of a subset of the real world, and where models of the real world entities, process and events are included in a virtual (three-dimensional) space. Virtual environments are ideal tools for simulation of certain critical processes, such as radiological accidents, where human beings or properties can suffer irreversible or long term damages. Radiological accidents are characterized by the significant exposure to radiation of specialized workers and general public. The early detection of a radiological accident and the determination of its possible extension are essential factors for the planning of prompt answers and emergency actions. This paper proposes the integration of georeferenced representation of the three-dimensional space and agent-based models, with the objective to construct virtual environments that have the capacity to simulate radiological accidents. The three-dimensional georeferenced representations of space candidates are: 1) the spatial representation of traditional geographical information systems (GIS); 2) the representation adopted by Google Maps®. Adding agents to these spatial representations allow us to simulate radiological accidents, quantify the doses received by members of the public, obtain a possible spatial distribution of people contaminated, estimate the number of contaminated individuals, estimate the impact on the health-network, estimate environmental impacts, generate exclusion zones, build alternative scenarios and train staff to deal with radiological accidents. (author)

  19. Influence of partial blockage of a BWR bundle on heat transfer, cladding temperature, and quenching during bottom flooding or top spraying under simulated LOCA conditions

    International Nuclear Information System (INIS)

    Brand, B.; Gaul, H.P.; Sarkar, J.

    1982-01-01

    In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m 3 /h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800 0 C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis

  20. Tritium in liquid phase in a BWR-5 like Laguna Verde; Tritio en fase liquida en un BWR-5 como Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Vargas A, A.; Cardenas J, J., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km 7.5, Veracruz (Mexico)

    2011-11-15

    In boiling water reactors (BWR), the tritium (H{sub 3}) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

  1. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

  2. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  3. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

  4. Qualified operator training in the simulated control room environment

    International Nuclear Information System (INIS)

    Ionescu, Teodor; Studineanu, Emil; Radulescu, Catalina; Bolocan, Gabriel

    2005-01-01

    Full text: Mainly designed for the training of the Cernavoda NPP Unit 2 operators, the virtual simulated environment allows the training of the already qualified operators for Cernavoda NPP Unit 1, adding to the already trained knowledge, the differences which has occurred in the Unit 2 design. Using state-of-the-art computers and displays and qualified software, the virtual simulated panels could offer a viable alternative to classic hardware-based training. This approach allows quick training of the new procedures required by the new configuration of the re-designed operator panels in the main control room of Cernavoda NPP Unit 2. (authors)

  5. Qualified operator training in the simulated control room environment

    International Nuclear Information System (INIS)

    Ionescu, Teodor; Studineanu, Emil; Radulescu, Catalina; Bolocan, Gabriel

    2005-01-01

    Mainly designed for the training of the Cernavoda NPP Unit 2 operators, the virtual simulated environment allows the training of the already qualified operators for Cernavoda NPP Unit 1, adding to the already trained knowledge, the differences which have occurred in the Unit 2 design. Using state-of-the-art computers and displays and qualified software, the virtual simulated panels could offer a viable alternative to classic hardware-based training. This approach allows quick training of the new procedures required by the new configuration of the re-designed operator panels in the main control room of Cernavoda NPP Unit 2. (authors)

  6. Electrophysiological measurement of interest during walking in a simulated environment.

    Science.gov (United States)

    Takeda, Yuji; Okuma, Takashi; Kimura, Motohiro; Kurata, Takeshi; Takenaka, Takeshi; Iwaki, Sunao

    2014-09-01

    A reliable neuroscientific technique for objectively estimating the degree of interest in a real environment is currently required in the research fields of neuroergonomics and neuroeconomics. Toward the development of such a technique, the present study explored electrophysiological measures that reflect an observer's interest in a nearly-real visual environment. Participants were asked to walk through a simulated shopping mall and the attractiveness of the shopping mall was manipulated by opening and closing the shutters of stores. During the walking task, participants were exposed to task-irrelevant auditory probes (two-stimulus oddball sequence). The results showed a smaller P2/early P3a component of task-irrelevant auditory event-related potentials and a larger lambda response of eye-fixation-related potentials in an interesting environment (i.e., open-shutter condition) than in a boring environment (i.e., closed-shutter condition); these findings can be reasonably explained by supposing that participants allocated more attentional resources to visual information in an interesting environment than in a boring environment, and thus residual attentional resources that could be allocated to task-irrelevant auditory probes were reduced. The P2/early P3a component and the lambda response may be useful measures of interest in a real visual environment. Copyright © 2014 Elsevier B.V. All rights reserved.

  7. Temperature field simulation of complex structures in fire environment

    International Nuclear Information System (INIS)

    Li Weifen; Hao Zhiming; Li Minghai

    2010-01-01

    In this paper, the typical model of the system of dangerous goods - steel - wood composite structure including components of explosives is used as the research object. Using MARC program, the temperature field of the structure in the fire environment is simulated. Radiation, conduction and convection heat transfer within the gap of the structure are taken into account, contact heat transfer is also considered. The phenomenon of thermal decomposition of wood in high temperature is deal with by equivalent method. The results show that the temperature of the explosives is not high in the fire environment. The timber inside the composite structure has played a very good insulation effect of explosives.

  8. Electrochemistry of lead in simulated ground water environments

    International Nuclear Information System (INIS)

    Joerg, E.A.; Devereux, O.F.

    1996-01-01

    Lead and lead alloys are used commonly as moisture barriers for underground cables. Lead exhibits excellent corrosion resistance in a variety of environments, but areas of localized attack have been found. These can result in able failures. The susceptibility of lead to pitting in several simulated ground water (SGW) environments was assessed using cyclic potentiodynamic pitting scans (PPS) and microscopy. Although general corrosion was observed, PPS demonstrated pitting did not occur in the same sense as in alloys known to be susceptible to pitting (i.e., very localized pit formation without general corrosion). However, areas of nonuniform general attack did occur, resulting in pitted surface morphologies

  9. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Porsmyr, Jan; Bodal, Terje; Beere, William H.

    2004-01-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

  10. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

    2004-07-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

  11. Experiences with a simulated learning environment - the SimuScape©: Virtual environments in medical education

    Directory of Open Access Journals (Sweden)

    Anna-Lena Thies

    2014-03-01

    Full Text Available INTRODUCTION: Simulation as a tool for medical education has gained considerable importance in the past years. Various studies have shown that the mastering of basic skills happens best if taught in a realistic and workplace-based context. It is necessary that simulation itself takes place in the realistic background of a genuine clinical or in an accordingly simulated learning environment. METHODS: A panoramic projection system that allows the simulation of different scenarios has been created at the medical school of the Westphalian Wilhelms-University  Muenster/Germany. The SimuScape© is a circular training room of six meters in diameter and has the capacity to generate pictures or moving images as well as the corresponding background noises for medical students, who are then able to interact with simulated patients inside a realistic environment. RESULTS: About 1,000 students have been instructed using the SimuScape© in the courses of emergency medicine, family medicine and anesthesia. The SimuScape©, with its 270°-panoramic projection, gives the students the impression “of being right in the center of action”.  It is a flexible learning environment that can be easily integrated into curricular teaching and which is in full operation for 10 days per semester. CONCLUSION: The SimuScape© allows the establishment of new medical areas outside the hospital and surgery for simulation and it is an extremely adaptable and cost-effective utilization of a lecture room. In this simulated environment it is possible to teach objectives like self-protection and patient care during disturbing environmental influences in practice.

  12. Simulation based virtual learning environment in medical genetics counseling

    DEFF Research Database (Denmark)

    Makransky, Guido; Bonde, Mads T; Wulff, Julie S G

    2016-01-01

    learning environments increase students' knowledge, intrinsic motivation, and self-efficacy, and help them generalize from laboratory analyses to clinical practice and health decision-making. METHODS: An entire class of 300 University of Copenhagen first-year undergraduate students, most with a major...... in medicine, received a 2-h training session in a simulation based learning environment. The main outcomes were pre- to post- changes in knowledge, intrinsic motivation, and self-efficacy, together with post-intervention evaluation of the effect of the simulation on student understanding of everyday clinical...... practice were demonstrated. RESULTS: Knowledge (Cohen's d = 0.73), intrinsic motivation (d = 0.24), and self-efficacy (d = 0.46) significantly increased from the pre- to post-test. Low knowledge students showed the greatest increases in knowledge (d = 3.35) and self-efficacy (d = 0.61), but a non...

  13. DIGITAL SIMULATIONS FOR IMPROVING EDUCATION: Learning Through Artificial Teaching Environments

    OpenAIRE

    Reviewed by Özlem OZAN

    2009-01-01

    DIGITAL SIMULATIONS FOR IMPROVING EDUCATION:Learning Through Artificial Teaching EnvironmentsGibson, David, Ed.D.; Information Science Reference, Hershey, PA,SBN-10: 1605663239, ISBN-13: 9781605663234, p.514 Jan 2009Reviewed byÖzlem OZANFaculty of Education, Eskişehir Osmangazi University,Eskisehir-TURKEYSimulations in education, both for children and adults,become popular with the development of computer technology, because they are fun and engaging and allow learners to internalize knowledg...

  14. Virtual environment display for a 3D audio room simulation

    Science.gov (United States)

    Chapin, William L.; Foster, Scott

    1992-06-01

    Recent developments in virtual 3D audio and synthetic aural environments have produced a complex acoustical room simulation. The acoustical simulation models a room with walls, ceiling, and floor of selected sound reflecting/absorbing characteristics and unlimited independent localizable sound sources. This non-visual acoustic simulation, implemented with 4 audio ConvolvotronsTM by Crystal River Engineering and coupled to the listener with a Poihemus IsotrakTM, tracking the listener's head position and orientation, and stereo headphones returning binaural sound, is quite compelling to most listeners with eyes closed. This immersive effect should be reinforced when properly integrated into a full, multi-sensory virtual environment presentation. This paper discusses the design of an interactive, visual virtual environment, complementing the acoustic model and specified to: 1) allow the listener to freely move about the space, a room of manipulable size, shape, and audio character, while interactively relocating the sound sources; 2) reinforce the listener's feeling of telepresence into the acoustical environment with visual and proprioceptive sensations; 3) enhance the audio with the graphic and interactive components, rather than overwhelm or reduce it; and 4) serve as a research testbed and technology transfer demonstration. The hardware/software design of two demonstration systems, one installed and one portable, are discussed through the development of four iterative configurations. The installed system implements a head-coupled, wide-angle, stereo-optic tracker/viewer and multi-computer simulation control. The portable demonstration system implements a head-mounted wide-angle, stereo-optic display, separate head and pointer electro-magnetic position trackers, a heterogeneous parallel graphics processing system, and object oriented C++ program code.

  15. Acoustic emission from fuel pellets in a simulated reactor environment

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Kennedy, C.R.; Reimann, K.J.

    1977-01-01

    Thermal-shock damage of nuclear reactor fuel pellets in a simulated reactor environment has been correlated with acoustic-emission data obtained from sensors placed on extensions of the electrical feedthroughs. Ringdown counts, rms output data, and event-location data has been acquired for experiments carried out with single pellets as well as multiple pellet stacks. These tests have shown that acoustic-emission monitoring can provide information indicating the onset and the extent of cracking

  16. Simulated learning environment experience in nursing students for paediatric practice.

    Science.gov (United States)

    Mendoza-Maldonado, Yessy; Barría-Pailaquilén, René Mauricio

    The training of health professionals requires the acquisition of clinical skills in a safe and efficient manner, which is facilitated by a simulated learning environment (SLE). It is also an efficient alternative when there are limitations for clinical practice in certain areas. This paper shows the work undertaken in a Chilean university in implementing paediatric practice using SLE. Over eight days, the care experience of a hospitalized infant was studied applying the nursing process. The participation of a paediatrician, resident physician, nursing technician, and simulated user was included in addition to the use of a simulation mannequin and equipment. Simulation of care was integral and covered interaction with the child and family and was developed in groups of six students by a teacher. The different phases of the simulation methodology were developed from a pedagogical point of view. The possibility of implementing paediatric clinical practice in an efficient and safe way was confirmed. The experience in SLE was highly valued by the students, allowing them to develop different skills and abilities required for paediatric nursing through simulation. Copyright © 2018 Elsevier España, S.L.U. All rights reserved.

  17. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  18. Simulated Space Environment Effects on a Candidate Solar Sail Material

    Science.gov (United States)

    Kang, Jin Ho; Bryant, Robert G.; Wilkie, W. Keats; Wadsworth, Heather M.; Craven, Paul D.; Nehls, Mary K.; Vaughn, Jason A.

    2017-01-01

    For long duration missions of solar sails, the sail material needs to survive harsh space environments and the degradation of the sail material controls operational lifetime. Therefore, understanding the effects of the space environment on the sail membrane is essential for mission success. In this study, we investigated the effect of simulated space environment effects of ionizing radiation, thermal aging and simulated potential damage on mechanical, thermal and optical properties of a commercial off the shelf (COTS) polyester solar sail membrane to assess the degradation mechanisms on a feasible solar sail. The solar sail membrane was exposed to high energy electrons (about 70 keV and 10 nA/cm2), and the physical properties were characterized. After about 8.3 Grad dose, the tensile modulus, tensile strength and failure strain of the sail membrane decreased by about 20 95%. The aluminum reflective layer was damaged and partially delaminated but it did not show any significant change in solar absorbance or thermal emittance. The effect on mechanical properties of a pre-cracked sample, simulating potential impact damage of the sail membrane, as well as thermal aging effects on metallized PEN (polyethylene naphthalate) film will be discussed.

  19. Expanding the modeling capabilities of the cognitive environment simulation

    International Nuclear Information System (INIS)

    Roth, E.M.; Mumaw, R.J.; Pople, H.E. Jr.

    1991-01-01

    The Nuclear Regulatory Commission has been conducting a research program to develop more effective tools to model the cognitive activities that underlie intention formation during nuclear power plant (NPP) emergencies. Under this program an artificial intelligence (AI) computer simulation called Cognitive Environment Simulation (CES) has been developed. CES simulates the cognitive activities involved in responding to a NPP accident situation. It is intended to provide an analytic tool for predicting likely human responses, and the kinds of errors that can plausibly arise under different accident conditions to support human reliability analysis. Recently CES was extended to handle a class of interfacing loss of coolant accidents (ISLOCAs). This paper summarizes the results of these exercises and describes follow-on work currently underway

  20. Virtual environment simulation as a tool to support evacuation planning

    International Nuclear Information System (INIS)

    Mol, Antonio C.; Grecco, Claudio H.S.; Santos, Isaac J.A.L.; Carvalho, Paulo V.R.; Jorge, Carlos A.F.; Sales, Douglas S.; Couto, Pedro M.; Botelho, Felipe M.; Bastos, Felipe R.

    2007-01-01

    This work is a preliminary study of the use of a free game-engine as a tool to build and to navigate in virtual environments, with a good degree of realism, for virtual simulations of evacuation from building and risk zones. To achieve this goal, some adjustments in the game engine have been implemented. A real building with four floors, consisting of some rooms with furniture and people, has been virtually implemented. Simulations of simple different evacuation scenarios have been performed, measuring the total time spent in each case. The measured times have been compared with their corresponding real evacuation times, measured in the real building. The first results have demonstrated that the virtual environment building with the free game engine is capable to reproduce the real situation with a satisfactory level. However, it is important to emphasize that such virtual simulations serve only as an aid in the planning of real evacuation simulations, and as such must never substitute the later. (author)

  1. Construction material processed using lunar simulant in various environments

    Science.gov (United States)

    Chase, Stan; Ocallaghan-Hay, Bridget; Housman, Ralph; Kindig, Michael; King, John; Montegrande, Kevin; Norris, Raymond; Vanscotter, Ryan; Willenborg, Jonathan; Staubs, Harry

    1995-01-01

    The manufacture of construction materials from locally available resources in space is an important first step in the establishment of lunar and planetary bases. The objective of the CoMPULSIVE (Construction Material Processed Using Lunar Simulant In Various Environments) experiment is to develop a procedure to produce construction materials by sintering or melting Johnson Space Center Simulant 1 (JSC-1) lunar soil simulant in both earth-based (1-g) and microgravity (approximately 0-g) environments. The characteristics of the resultant materials will be tested to determine its physical and mechanical properties. The physical characteristics include: crystalline, thermal, and electrical properties. The mechanical properties include: compressive tensile, and flexural strengths. The simulant, placed in a sealed graphite crucible, will be heated using a high temperature furnace. The crucible will then be cooled by radiative and forced convective means. The core furnace element consists of space qualified quartz-halogen incandescent lamps with focusing mirrors. Sample temperatures of up to 2200 C are attainable using this heating method.

  2. Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

    International Nuclear Information System (INIS)

    Wiles, L.E.; McCann, R.A.

    1981-09-01

    This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

  3. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  4. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  5. Creating pedestrian crash scenarios in a driving simulator environment.

    Science.gov (United States)

    Chrysler, Susan T; Ahmad, Omar; Schwarz, Chris W

    2015-01-01

    In 2012 in the United States, pedestrian injuries accounted for 3.3% of all traffic injuries but, disproportionately, pedestrian fatalities accounted for roughly 14% of traffic-related deaths (NHTSA 2014 ). In many other countries, pedestrians make up more than 50% of those injured and killed in crashes. This research project examined driver response to crash-imminent situations involving pedestrians in a high-fidelity, full-motion driving simulator. This article presents a scenario development method and discusses experimental design and control issues in conducting pedestrian crash research in a simulation environment. Driving simulators offer a safe environment in which to test driver response and offer the advantage of having virtual pedestrian models that move realistically, unlike test track studies, which by nature must use pedestrian dummies on some moving track. An analysis of pedestrian crash trajectories, speeds, roadside features, and pedestrian behavior was used to create 18 unique crash scenarios representative of the most frequent and most costly crash types. For the study reported here, we only considered scenarios where the car is traveling straight because these represent the majority of fatalities. We manipulated driver expectation of a pedestrian both by presenting intersection and mid-block crossing as well as by using features in the scene to direct the driver's visual attention toward or away from the crossing pedestrian. Three visual environments for the scenarios were used to provide a variety of roadside environments and speed: a 20-30 mph residential area, a 55 mph rural undivided highway, and a 40 mph urban area. Many variables of crash situations were considered in selecting and developing the scenarios, including vehicle and pedestrian movements; roadway and roadside features; environmental conditions; and characteristics of the pedestrian, driver, and vehicle. The driving simulator scenarios were subjected to iterative testing to

  6. Comparison of the corrosion potential for stainless steel measured in-plant and in laboratory during BWR normal water chemistry conditions

    International Nuclear Information System (INIS)

    Molander, A.; Pein, K.; Tarkpea, P.; Takagi, Junichi; Karlberg, G.; Gott, K.

    1998-01-01

    To obtain reliable crack growth rate date for stainless steel in BWR environments careful laboratory simulation of the environmental conditions is necessary. In the plant the BWR normal water chemistry environment contains hydrogen peroxide, oxygen and hydrogen. However, in crack growth rate experiments in laboratories, the environment is normally simulated by adding 200 ppb oxygen to the high temperature water. Thus, as hydrogen peroxide is a more powerful oxidant than oxygen, it is to be expected that a lower corrosion potential will be measured in the laboratory than in the plant. To resolve this issue this work has been performed. In-plant and laboratory measurements have often been performed with somewhat different equipment, due to the special requirements concerning in-plant measurements. In this work such differences have been avoided and two identical sets of equipment for electrochemical measurements were built and used for measurements in-plant in a Swedish BWR and in high purity water in the laboratory. The host plant was Barsebaeck 1. Corrosion potential monitoring in-plant was performed under both NWC (Normal Water Chemistry) and HWC (Hydrogen Water Chemistry) conditions. This paper is, however, focused on NWC conditions. This is due to the fact, that the total crack growth obtained during a reactor cycle, can be determined by NWC conditions, even for plants running with HWC due to periodic stops in the hydrogen addition for turbine inspections or failure of the dosage or hydrogen production equipment. Thus, crack growth data for NWC is of great importance both for BWRs operating with HWC and NWC. Measurements in-plant and in the laboratory were performed during additions of oxygen and hydrogen peroxide to the autoclave systems. The corrosion potentials were compared for various conditions in the autoclaves, as well as versus in-plant in-pipe corrosion potentials. (J.P.N.)

  7. Construction of the quantitative analysis environment using Monte Carlo simulation

    International Nuclear Information System (INIS)

    Shirakawa, Seiji; Ushiroda, Tomoya; Hashimoto, Hiroshi; Tadokoro, Masanori; Uno, Masaki; Tsujimoto, Masakazu; Ishiguro, Masanobu; Toyama, Hiroshi

    2013-01-01

    The thoracic phantom image was acquisitioned of the axial section to construct maps of the source and density with Monte Carlo (MC) simulation. The phantom was Heart/Liver Type HL (Kyoto Kagaku Co., Ltd.) single photon emission CT (SPECT)/CT machine was Symbia T6 (Siemence) with the collimator LMEGP (low-medium energy general purpose). Maps were constructed from CT images with an in-house software using Visual studio C Sharp (Microsoft). The code simulation of imaging nuclear detectors (SIMIND) was used for MC simulation, Prominence processor (Nihon Medi-Physics) for filter processing and image reconstruction, and the environment DELL Precision T7400 for all image processes. For the actual experiment, the phantom was given 15 MBq of 99m Tc assuming the uptake 2% at the dose of 740 MBq in its myocardial portion and SPECT image was acquisitioned and reconstructed with Butter-worth filter and filter back projection method. CT images were similarly obtained in 0.3 mm thick slices, which were filed in one formatted with digital imaging and communication in medicine (DICOM), and then processed for application to SIMIND for mapping the source and density. Physical and mensuration factors were examined in ideal images by sequential exclusion and simulation of those factors as attenuation, scattering, spatial resolution deterioration and statistical fluctuation. Gamma energy spectrum, SPECT projection and reconstructed images given by the simulation were found to well agree with the actual data, and the precision of MC simulation was confirmed. Physical and mensuration factors were found to be evaluable individually, suggesting the usefulness of the simulation for assessing the precision of their correction. (T.T.)

  8. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  9. BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

    International Nuclear Information System (INIS)

    Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

    2002-01-01

    Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

  10. Simulation Environment Based on the Universal Verification Methodology

    CERN Document Server

    AUTHOR|(SzGeCERN)697338

    2017-01-01

    Universal Verification Methodology (UVM) is a standardized approach of verifying integrated circuit designs, targeting a Coverage-Driven Verification (CDV). It combines automatic test generation, self-checking testbenches, and coverage metrics to indicate progress in the design verification. The flow of the CDV differs from the traditional directed-testing approach. With the CDV, a testbench developer, by setting the verification goals, starts with an structured plan. Those goals are targeted further by a developed testbench, which generates legal stimuli and sends them to a device under test (DUT). The progress is measured by coverage monitors added to the simulation environment. In this way, the non-exercised functionality can be identified. Moreover, the additional scoreboards indicate undesired DUT behaviour. Such verification environments were developed for three recent ASIC and FPGA projects which have successfully implemented the new work-flow: (1) the CLICpix2 65 nm CMOS hybrid pixel readout ASIC desi...

  11. Comparative analysis of the simulation of the instantaneous closing of the discharge valve of a recirculation loop of a BWR with a model of recirculation loop with 2 jet pumps and another model with 20 jet pumps using RELAP5/SCDAPSIM Mod. 3.4

    International Nuclear Information System (INIS)

    Araiza M, E.; Ortiz V, J.; Martinez C, E.; Amador G, R.; Castillo D, R.

    2016-09-01

    This work presents the results of the simulation of the instantaneous closing of the water hammer, of a recirculation loop using two different arrangements in the loops. One of these arrangements corresponds to the traditional model that uses only two jet pumps to simulate the twenty pumps of the two recirculation loops of a BWR. The second nodalization models each of the ten jet pumps of each recirculation loop. The results obtained from the execution of both models are compared, using important variables such as pressures and mass costs for the same components of both models. In addition, the maximum pressure value generated on the pipe located upstream of the water hammer, relative to the design pressure of the pipe, is compared for each arrangement. (Author)

  12. Improved climate risk simulations for rice in arid environments.

    Directory of Open Access Journals (Sweden)

    Pepijn A J van Oort

    Full Text Available We integrated recent research on cardinal temperatures for phenology and early leaf growth, spikelet formation, early morning flowering, transpirational cooling, and heat- and cold-induced sterility into an existing to crop growth model ORYZA2000. We compared for an arid environment observed potential yields with yields simulated with default ORYZA2000, with modified subversions of ORYZA2000 and with ORYZA_S, a model developed for the region of interest in the 1990s. Rice variety 'IR64' was sown monthly 15-times in a row in two locations in Senegal. The Senegal River Valley is located in the Sahel, near the Sahara desert with extreme temperatures during day and night. The existing subroutines underestimated cold stress and overestimated heat stress. Forcing the model to use observed spikelet number and phenology and replacing the existing heat and cold subroutines improved accuracy of yield simulation from EF = -0.32 to EF =0.70 (EF is modelling efficiency. The main causes of improved accuracy were that the new model subversions take into account transpirational cooling (which is high in arid environments and early morning flowering for heat sterility, and minimum rather than average temperature for cold sterility. Simulations were less accurate when also spikelet number and phenology were simulated. Model efficiency was 0.14 with new heat and cold routines and improved to 0.48 when using new cardinal temperatures for phenology and early leaf growth. The new adapted subversion of ORYZA2000 offers a powerful analytic tool for climate change impact assessment and cropping calendar optimisation in arid regions.

  13. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

    Andrade, G.G. de.

    1973-01-01

    First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

  14. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  15. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  16. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

    2010-01-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  17. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

    2010-07-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  18. Corrosion resistance improvement of ferritic steels through hydrogen additions to the BWR coolant

    International Nuclear Information System (INIS)

    Gordon, B.M.; Jewett, C.W.; Pickett, A.E.; Indig, M.E.

    1984-01-01

    Motivated by the success of oxygen suppression for mitigation of intergranular stress corrosion cracking (IGSCC) in weld sensitized austenitic materials used in Boiling Water Reactors (BWRs), oxygen suppression, through hydrogen additions to the feedwater was investigated to determine its affect on the corrosion resistance of ferritic and martensitic BWR structural materials. The results of these investigations are presented in this paper, where particular emphasis is placed on the corrosion performance of BWR pressure vessel low alloy steels, carbon steel piping materials and martensitic pump materials. It is important to note that the corrosion resistance of these materials in the BWR environment is excellent. Consequently this investigation was also motivated to determine whether there were any detrimental effects of hydrogen additions, as well as to identify any additional margin in ferritic/martensitic materials corrosion performance

  19. Tritium in liquid phase in a BWR-5 like Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Vargas A, A.; Cardenas J, J.

    2011-11-01

    In boiling water reactors (BWR), the tritium (H 3 ) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

  20. ESSE: Engineering Super Simulation Emulation for Virtual Reality Systems Environment

    International Nuclear Information System (INIS)

    Suh, Kune Y.; Yeon, Choul W.

    2008-01-01

    The trademark 4 + D Technology TM based Engineering Super Simulation Emulation (ESSE) is introduced. ESSE resorting to three-dimensional (3D) Virtual Reality (VR) technology pledges to provide with an interactive real-time motion, sound and tactile and other forms of feedback in the man machine systems environment. In particular, the 3D Virtual Engineering Neo cybernetic Unit Soft Power (VENUS) adds a physics engine to the VR platform so as to materialize a physical atmosphere. A close cooperation system and prompt information share are crucial, thereby increasing the necessity of centralized information system and electronic cooperation system. VENUS is further deemed to contribute towards public acceptance of nuclear power in general, and safety in particular. For instance, visualization of nuclear systems can familiarize the public in answering their questions and alleviating misunderstandings on nuclear power plants answering their questions and alleviating misunderstandings on nuclear power plants (NPPs) in general, and performance, security and safety in particular. An in-house flagship project Systemic Three-dimensional Engine Platform Prototype Engineering (STEPPE) endeavors to develop the Systemic Three-dimensional Engine Platform (STEP) for a variety of VR applications. STEP is home to a level system providing the whole visible scene of virtual engineering of man machine system environment. The system is linked with video monitoring that provides a 3D Computer Graphics (CG) visualization of major events. The database linked system provides easy access to relevant blueprints. The character system enables the operators easy access to visualization of major events. The database linked system provides easy access to relevant blueprints. The character system enables the operators to access the virtual systems by using their virtual characters. Virtually Engineered NPP Informative systems by using their virtual characters. Virtually Engineered NPP Informative

  1. Comparison of discrete event simulation tools in an academic environment

    Directory of Open Access Journals (Sweden)

    Mario Jadrić

    2014-12-01

    Full Text Available A new research model for simulation software evaluation is proposed consisting of three main categories of criteria: modeling and simulation capabilities of the explored tools, and tools’ input/output analysis possibilities, all with respective sub-criteria. Using the presented model, two discrete event simulation tools are evaluated in detail using the task-centred scenario. Both tools (Arena and ExtendSim were used for teaching discrete event simulation in preceding academic years. With the aim to inspect their effectiveness and to help us determine which tool is more suitable for students i.e. academic purposes, we used a simple simulation model of entities competing for limited resources. The main goal was to measure subjective (primarily attitude and objective indicators while using the tools when the same simulation scenario is given. The subjects were first year students of Master studies in Information Management at the Faculty of Economics in Split taking a course in Business Process Simulations (BPS. In a controlled environment – in a computer lab, two groups of students were given detailed, step-by-step instructions for building models using both tools - first using ExtendSim then Arena or vice versa. Subjective indicators (students’ attitudes were collected using an online survey completed immediately upon building each model. Subjective indicators primarily include students’ personal estimations of Arena and ExtendSim capabilities/features for model building, model simulation and result analysis. Objective indicators were measured using specialised software that logs information on user's behavior while performing a particular task on their computer such as distance crossed by mouse during model building, the number of mouse clicks, usage of the mouse wheel and speed achieved. The results indicate that ExtendSim is well preferred comparing to Arena with regards to subjective indicators while the objective indicators are

  2. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  3. A COMPUTATIONAL WORKBENCH ENVIRONMENT FOR VIRTUAL POWER PLANT SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Mike Bockelie; Dave Swensen; Martin Denison; Adel Sarofim; Connie Senior

    2004-12-22

    In this report is described the work effort to develop and demonstrate a software framework to support advanced process simulations to evaluate the performance of advanced power systems. Integrated into the framework are a broad range of models, analysis tools, and visualization methods that can be used for the plant evaluation. The framework provides a tightly integrated problem-solving environment, with plug-and-play functionality, and includes a hierarchy of models, ranging from fast running process models to detailed reacting CFD models. The framework places no inherent limitations on the type of physics that can be modeled, numerical techniques, or programming languages used to implement the equipment models, or the type or amount of data that can be exchanged between models. Tools are provided to analyze simulation results at multiple levels of detail, ranging from simple tabular outputs to advanced solution visualization methods. All models and tools communicate in a seamless manner. The framework can be coupled to other software frameworks that provide different modeling capabilities. Three software frameworks were developed during the course of the project. The first framework focused on simulating the performance of the DOE Low Emissions Boiler System Proof of Concept facility, an advanced pulverized-coal combustion-based power plant. The second framework targeted simulating the performance of an Integrated coal Gasification Combined Cycle - Fuel Cell Turbine (IGCC-FCT) plant configuration. The coal gasifier models included both CFD and process models for the commercially dominant systems. Interfacing models to the framework was performed using VES-Open, and tests were performed to demonstrate interfacing CAPE-Open compliant models to the framework. The IGCC-FCT framework was subsequently extended to support Virtual Engineering concepts in which plant configurations can be constructed and interrogated in a three-dimensional, user-centered, interactive

  4. Optical intensity scintillation in the simulated atmospherical environment

    Science.gov (United States)

    Hajek, Lukas; Latal, Jan; Vanderka, Ales; Vitasek, Jan; Bojko, Marian; Bednarek, Lukas; Vasinek, Vladimir

    2016-09-01

    There are several parameters of the atmospheric environment which have an effect on the optical wireless connection. Effects like fog, snow or rain are ones of the effects which appears tendentiously and which are bound by season, geographic location, etc. One of the effects that appear with various intensity for the whole time is airflow. The airflow changes the local refractive index of the air and areas with lower or higher refractive index form. The light going through these areas refracts and due to the optical intensity scintillates on the detector of the receiver. The airflow forms on the basis of two effects in the atmosphere. The first is wind cut and flowing over barriers. The other is thermal flow when warm air rises to the higher layers of the atmosphere. The heart of this article is creation such an environment that will form airflow and the refractive index will scintillate. For the experiment, we used special laboratory box with high-speed ventilators and heating units to simulate atmospheric turbulence. We monitor the impact of ventilator arrangement and air temperature on the scintillation of the gas laser with wavelength 633 nm/15 mW. In the experiment, there is watched the difference in behavior between real measurement and flow simulation with the same peripheral conditions of the airflow in the area of 500 x 500 cm.

  5. Comparative analysis of the simulation of the instantaneous closing of the discharge valve of a recirculation loop of a BWR with a model of recirculation loop with 2 jet pumps and another model with 20 jet pumps using RELAP5/SCDAPSIM Mod. 3.4; Analisis comparativo de la simulacion del cierre instantaneo de la valvula de descarga de un lazo de recirculacion de un BWR con un modelo de lazo de recirculacion con 2 bombas chorro y un modelo con 20 bombas chorro empleando RELAP5/SCDAPSIM Mod. 3.4

    Energy Technology Data Exchange (ETDEWEB)

    Araiza M, E.; Ortiz V, J.; Martinez C, E.; Amador G, R.; Castillo D, R., E-mail: enrique.araiza@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This work presents the results of the simulation of the instantaneous closing of the water hammer, of a recirculation loop using two different arrangements in the loops. One of these arrangements corresponds to the traditional model that uses only two jet pumps to simulate the twenty pumps of the two recirculation loops of a BWR. The second nodalization models each of the ten jet pumps of each recirculation loop. The results obtained from the execution of both models are compared, using important variables such as pressures and mass costs for the same components of both models. In addition, the maximum pressure value generated on the pipe located upstream of the water hammer, relative to the design pressure of the pipe, is compared for each arrangement. (Author)

  6. Augmenting Sand Simulation Environments through Subdivision and Particle Refinement

    Science.gov (United States)

    Clothier, M.; Bailey, M.

    2012-12-01

    Recent advances in computer graphics and parallel processing hardware have provided disciplines with new methods to evaluate and visualize data. These advances have proven useful for earth and planetary scientists as many researchers are using this hardware to process large amounts of data for analysis. As such, this has provided opportunities for collaboration between computer graphics and the earth sciences. Through collaboration with the Oregon Space Grant and IGERT Ecosystem Informatics programs, we are investigating techniques for simulating the behavior of sand. We are also collaborating with the Jet Propulsion Laboratory's (JPL) DARTS Lab to exchange ideas and gain feedback on our research. The DARTS Lab specializes in simulation of planetary vehicles, such as the Mars rovers. Their simulations utilize a virtual "sand box" to test how a planetary vehicle responds to different environments. Our research builds upon this idea to create a sand simulation framework so that planetary environments, such as the harsh, sandy regions on Mars, are more fully realized. More specifically, we are focusing our research on the interaction between a planetary vehicle, such as a rover, and the sand beneath it, providing further insight into its performance. Unfortunately, this can be a computationally complex problem, especially if trying to represent the enormous quantities of sand particles interacting with each other. However, through the use of high-performance computing, we have developed a technique to subdivide areas of actively participating sand regions across a large landscape. Similar to a Level of Detail (LOD) technique, we only subdivide regions of a landscape where sand particles are actively participating with another object. While the sand is within this subdivision window and moves closer to the surface of the interacting object, the sand region subdivides into smaller regions until individual sand particles are left at the surface. As an example, let's say

  7. A High-Fidelity Batch Simulation Environment for Integrated Batch and Piloted Air Combat Simulation Analysis

    Science.gov (United States)

    Goodrich, Kenneth H.; McManus, John W.; Chappell, Alan R.

    1992-01-01

    A batch air combat simulation environment known as the Tactical Maneuvering Simulator (TMS) is presented. The TMS serves as a tool for developing and evaluating tactical maneuvering logics. The environment can also be used to evaluate the tactical implications of perturbations to aircraft performance or supporting systems. The TMS is capable of simulating air combat between any number of engagement participants, with practical limits imposed by computer memory and processing power. Aircraft are modeled using equations of motion, control laws, aerodynamics and propulsive characteristics equivalent to those used in high-fidelity piloted simulation. Databases representative of a modern high-performance aircraft with and without thrust-vectoring capability are included. To simplify the task of developing and implementing maneuvering logics in the TMS, an outer-loop control system known as the Tactical Autopilot (TA) is implemented in the aircraft simulation model. The TA converts guidance commands issued by computerized maneuvering logics in the form of desired angle-of-attack and wind axis-bank angle into inputs to the inner-loop control augmentation system of the aircraft. This report describes the capabilities and operation of the TMS.

  8. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  9. Plasma environment of Titan: a 3-D hybrid simulation study

    Directory of Open Access Journals (Sweden)

    S. Simon

    2006-05-01

    Full Text Available Titan possesses a dense atmosphere, consisting mainly of molecular nitrogen. Titan's orbit is located within the Saturnian magnetosphere most of the time, where the corotating plasma flow is super-Alfvénic, yet subsonic and submagnetosonic. Since Titan does not possess a significant intrinsic magnetic field, the incident plasma interacts directly with the atmosphere and ionosphere. Due to the characteristic length scales of the interaction region being comparable to the ion gyroradii in the vicinity of Titan, magnetohydrodynamic models can only offer a rough description of Titan's interaction with the corotating magnetospheric plasma flow. For this reason, Titan's plasma environment has been studied by using a 3-D hybrid simulation code, treating the electrons as a massless, charge-neutralizing fluid, whereas a completely kinetic approach is used to cover ion dynamics. The calculations are performed on a curvilinear simulation grid which is adapted to the spherical geometry of the obstacle. In the model, Titan's dayside ionosphere is mainly generated by solar UV radiation; hence, the local ion production rate depends on the solar zenith angle. Because the Titan interaction features the possibility of having the densest ionosphere located on a face not aligned with the ram flow of the magnetospheric plasma, a variety of different scenarios can be studied. The simulations show the formation of a strong magnetic draping pattern and an extended pick-up region, being highly asymmetric with respect to the direction of the convective electric field. In general, the mechanism giving rise to these structures exhibits similarities to the interaction of the ionospheres of Mars and Venus with the supersonic solar wind. The simulation results are in agreement with data from recent Cassini flybys.

  10. Plasma environment of Titan: a 3-D hybrid simulation study

    Directory of Open Access Journals (Sweden)

    S. Simon

    2006-05-01

    Full Text Available Titan possesses a dense atmosphere, consisting mainly of molecular nitrogen. Titan's orbit is located within the Saturnian magnetosphere most of the time, where the corotating plasma flow is super-Alfvénic, yet subsonic and submagnetosonic. Since Titan does not possess a significant intrinsic magnetic field, the incident plasma interacts directly with the atmosphere and ionosphere. Due to the characteristic length scales of the interaction region being comparable to the ion gyroradii in the vicinity of Titan, magnetohydrodynamic models can only offer a rough description of Titan's interaction with the corotating magnetospheric plasma flow. For this reason, Titan's plasma environment has been studied by using a 3-D hybrid simulation code, treating the electrons as a massless, charge-neutralizing fluid, whereas a completely kinetic approach is used to cover ion dynamics. The calculations are performed on a curvilinear simulation grid which is adapted to the spherical geometry of the obstacle. In the model, Titan's dayside ionosphere is mainly generated by solar UV radiation; hence, the local ion production rate depends on the solar zenith angle. Because the Titan interaction features the possibility of having the densest ionosphere located on a face not aligned with the ram flow of the magnetospheric plasma, a variety of different scenarios can be studied. The simulations show the formation of a strong magnetic draping pattern and an extended pick-up region, being highly asymmetric with respect to the direction of the convective electric field. In general, the mechanism giving rise to these structures exhibits similarities to the interaction of the ionospheres of Mars and Venus with the supersonic solar wind. The simulation results are in agreement with data from recent Cassini flybys.

  11. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  12. Evaluation of thermal margin during BWR neutron flux oscillation

    International Nuclear Information System (INIS)

    Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

    1992-01-01

    Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

  13. A detailed BWR recirculation loop model for RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

    2017-01-15

    Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

  14. TRAB, a transient analysis program for BWR. Part 1

    International Nuclear Information System (INIS)

    Rajamaeki, Markku.

    1980-03-01

    TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

  15. Specification of requirements for the virtual environment for reactor applications simulation environment

    International Nuclear Information System (INIS)

    Hess, S. M.; Pytel, M.

    2012-01-01

    In 2010, the United States Dept. of Energy initiated a research and development effort to develop modern modeling and simulation methods that could utilize high performance computing capabilities to address issues important to nuclear power plant operation, safety and sustainability. To respond to this need, a consortium of national laboratories, academic institutions and industry partners (the Consortium for Advanced Simulation of Light Water Reactors - CASL) was formed to develop an integrated Virtual Environment for Reactor Applications (VERA) modeling and simulation capability. A critical element for the success of the CASL research and development effort was the development of an integrated set of overarching requirements that provides guidance in the planning, development, and management of the VERA modeling and simulation software. These requirements also provide a mechanism from which the needs of a broad array of external CASL stakeholders (e.g. reactor / fuel vendors, plant owner / operators, regulatory personnel, etc.) can be identified and integrated into the VERA development plans. This paper presents an overview of the initial set of requirements contained within the VERA Requirements Document (VRD) that currently is being used to govern development of the VERA software within the CASL program. The complex interdisciplinary nature of these requirements together with a multi-physics coupling approach to realize a core simulator capability pose a challenge to how the VRD should be derived and subsequently revised to accommodate the needs of different stakeholders. Thus, the VRD is viewed as an evolving document that will be updated periodically to reflect the changing needs of identified CASL stakeholders and lessons learned during the progress of the CASL modeling and simulation program. (authors)

  16. BWR Refill-Reflood Program. Final report

    International Nuclear Information System (INIS)

    Myers, L.L.

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

  17. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1983-01-01

    A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  18. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  19. Uncertainties in source term estimates for a station blackout accident in a BWR with Mark I containment

    International Nuclear Information System (INIS)

    Lee, M.; Cazzoli, E.; Liu, Y.; Davis, R.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.; Khatib-Rahbar, M.

    1988-01-01

    In this paper, attention is limited to a single accident progression sequence, namly a station blackout accident in a BWR with a Mark I containment building. Identified as an important accident in the draft version of NUREG-1150 a station blackout involves loss of both off-site power and dc power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation cooling systems. This paper illustrates the calculated uncertainties (Probability Density Functions) associated with the radiological releases into the environment for the nine fission product groups at 10 hours following the initiation of core-concrete interactions. Also shown are the results ofthe STCP base case simulation. 5 refs., 1 fig., 1 tab

  20. Home automation and simulation of presence in empty environments

    Directory of Open Access Journals (Sweden)

    Marques Israel

    2017-01-01

    Full Text Available Since their humble beginnings at the dawn of the 20th Century until contemporary age, automation and control systems have grown exponentially in both complexity and importance. Its relevance on human activities, be they mundane tasks or crucial processes, is self-evident. Among its many utilities, automated systems acquire a noble mission when put in service to protect life and property from aggressors of any kind. This paper discusses how home automation components can be utilized to implement an alternative domestic security strategy that consists in simulating the presence of an individual in an empty environment in the absence of its owner in order dissuade potential trespassing criminals, once they would feel highly discouraged to carry the criminal act should they believe the property is occupied.

  1. Calculation and simulation of atmospheric refraction effects in maritime environments

    Science.gov (United States)

    Dion, Denis, Jr.; Gardenal, Lionel; Lahaie, P.; Forand, J. Luc

    2001-01-01

    Near the sea surface, atmospheric refraction and turbulence affect both IR transmission and image quality. This produces an impact on both the detection and classification/identification of targets. With the financial participation of the U.S. Office of Naval Research (ONR), Canada's Defence Research Establishment Valcartier (DREV) is developing PRIME (Propagation Resources In the Maritime Environment), a computer model aimed at describing the overall atmospheric effects on IR imagery systems in the marine surface layer. PRIME can be used as a complement to MODTRAN to compute the effective transmittance in the marine surface layer, taking into account the lens effects caused by refraction. It also provides information on image degradation caused by both refraction and turbulence. This paper reviews the refraction phenomena that take place in the surface layer and discusses their effects on target detection and identification. We then show how PRIME can benefit detection studies and image degradation simulations.

  2. Flexible Environments for Grand-Challenge Simulation in Climate Science

    Science.gov (United States)

    Pierrehumbert, R.; Tobis, M.; Lin, J.; Dieterich, C.; Caballero, R.

    2004-12-01

    Current climate models are monolithic codes, generally in Fortran, aimed at high-performance simulation of the modern climate. Though they adequately serve their designated purpose, they present major barriers to application in other problems. Tailoring them to paleoclimate of planetary simulations, for instance, takes months of work. Theoretical studies, where one may want to remove selected processes or break feedback loops, are similarly hindered. Further, current climate models are of little value in education, since the implementation of textbook concepts and equations in the code is obscured by technical detail. The Climate Systems Center at the University of Chicago seeks to overcome these limitations by bringing modern object-oriented design into the business of climate modeling. Our ultimate goal is to produce an end-to-end modeling environment capable of configuring anything from a simple single-column radiative-convective model to a full 3-D coupled climate model using a uniform, flexible interface. Technically, the modeling environment is implemented as a Python-based software component toolkit: key number-crunching procedures are implemented as discrete, compiled-language components 'glued' together and co-ordinated by Python, combining the high performance of compiled languages and the flexibility and extensibility of Python. We are incrementally working towards this final objective following a series of distinct, complementary lines. We will present an overview of these activities, including PyOM, a Python-based finite-difference ocean model allowing run-time selection of different Arakawa grids and physical parameterizations; CliMT, an atmospheric modeling toolkit providing a library of 'legacy' radiative, convective and dynamical modules which can be knitted into dynamical models, and PyCCSM, a version of NCAR's Community Climate System Model in which the coupler and run-control architecture are re-implemented in Python, augmenting its flexibility

  3. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  4. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  5. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  6. Flight Simulation of ARES in the Mars Environment

    Science.gov (United States)

    Kenney, P. Sean; Croom, Mark A.

    2011-01-01

    A report discusses using the Aerial Regional- scale Environmental Survey (ARES) light airplane as an observation platform on Mars in order to gather data. It would have to survive insertion into the atmosphere, fly long enough to meet science objectives, and provide a stable platform. The feasibility of such a platform was tested using the Langley Standard Real- Time Simulation in C++. The unique features of LaSRS++ are: full, six-degrees- of-freedom flight simulation that can be used to evaluate the performance of the aircraft in the Martian environment; capability of flight analysis from start to finish; support of Monte Carlo analysis of aircraft performance; and accepting initial conditions from POST results for the entry and deployment of the entry body. Starting with a general aviation model, the design was tweaked to maintain a stable aircraft under expected Martian conditions. Outer mold lines were adjusted based on experience with the Martian atmosphere. Flight control was modified from a vertical acceleration control law to an angle-of-attack control law. Navigation was modified from a vertical acceleration control system to an alpha control system. In general, a pattern of starting with simple models with well-understood behaviors was selected and modified during testing.

  7. Simulation of machine-maintenance training in virtual environment

    International Nuclear Information System (INIS)

    Yoshikawa, Hidekazu; Tezuka, Tetsuo; Kashiwa, Ken-ichiro; Ishii, Hirotake

    1997-01-01

    The periodical inspection of nuclear power plants needs a lot of workforces with a high degree of technical skill for the maintenance of various sorts of machines. Therefore, a new type of maintenance training system is required, where trainees can get training safely, easily and effectively. In this study we developed a training simulation system for disassembling a check valve in virtual environment (VE). The features of this system are as follows: Firstly, the trainees can execute tasks even in wrong order, and can experience the resultant conditions. In order to realize this environment, we developed a new Petri-net model for representing the objects' states in VE. This Petri-net model has several original characteristics, which make it easier to manage the change of the objects' states. Furthermore, we made a support system for constructing the Petri-net model of machine-disassembling training, because the Petri-net model is apt to become of large size. The effectiveness of this support system is shown through the system development. Secondly, this system can perform appropriate tasks to be done next in VE whenever the trainee wants even after some mistakes have been made. The effectiveness of this function has also been confirmed by experiments. (author)

  8. Simulation environment based on the Universal Verification Methodology

    International Nuclear Information System (INIS)

    Fiergolski, A.

    2017-01-01

    Universal Verification Methodology (UVM) is a standardized approach of verifying integrated circuit designs, targeting a Coverage-Driven Verification (CDV). It combines automatic test generation, self-checking testbenches, and coverage metrics to indicate progress in the design verification. The flow of the CDV differs from the traditional directed-testing approach. With the CDV, a testbench developer, by setting the verification goals, starts with an structured plan. Those goals are targeted further by a developed testbench, which generates legal stimuli and sends them to a device under test (DUT). The progress is measured by coverage monitors added to the simulation environment. In this way, the non-exercised functionality can be identified. Moreover, the additional scoreboards indicate undesired DUT behaviour. Such verification environments were developed for three recent ASIC and FPGA projects which have successfully implemented the new work-flow: (1) the CLICpix2 65 nm CMOS hybrid pixel readout ASIC design; (2) the C3PD 180 nm HV-CMOS active sensor ASIC design; (3) the FPGA-based DAQ system of the CLICpix chip. This paper, based on the experience from the above projects, introduces briefly UVM and presents a set of tips and advices applicable at different stages of the verification process-cycle.

  9. Simulating cloud environment for HIS backup using secret sharing.

    Science.gov (United States)

    Kuroda, Tomohiro; Kimura, Eizen; Matsumura, Yasushi; Yamashita, Yoshinori; Hiramatsu, Haruhiko; Kume, Naoto

    2013-01-01

    In the face of a disaster hospitals are expected to be able to continue providing efficient and high-quality care to patients. It is therefore crucial for hospitals to develop business continuity plans (BCPs) that identify their vulnerabilities, and prepare procedures to overcome them. A key aspect of most hospitals' BCPs is creating the backup of the hospital information system (HIS) data at multiple remote sites. However, the need to keep the data confidential dramatically increases the costs of making such backups. Secret sharing is a method to split an original secret message so that individual pieces are meaningless, but putting sufficient number of pieces together reveals the original message. It allows creation of pseudo-redundant arrays of independent disks for privacy-sensitive data over the Internet. We developed a secret sharing environment for StarBED, a large-scale network experiment environment, and evaluated its potential and performance during disaster recovery. Simulation results showed that the entire main HIS database of Kyoto University Hospital could be retrieved within three days even if one of the distributed storage systems crashed during a disaster.

  10. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  11. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 {sup o}C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV

  12. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 o C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV operating experience

  13. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor; Post-procesador para simulaciones del programa ORIGEN y calculo de la composicion de la actividad de un nucleo de combustible quemado por un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval V, S. [IIE, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sandoval@iie.org.mx

    2006-07-01

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

  14. STSE: Spatio-Temporal Simulation Environment Dedicated to Biology

    Directory of Open Access Journals (Sweden)

    Gerber Susanne

    2011-04-01

    Full Text Available Abstract Background Recently, the availability of high-resolution microscopy together with the advancements in the development of biomarkers as reporters of biomolecular interactions increased the importance of imaging methods in molecular cell biology. These techniques enable the investigation of cellular characteristics like volume, size and geometry as well as volume and geometry of intracellular compartments, and the amount of existing proteins in a spatially resolved manner. Such detailed investigations opened up many new areas of research in the study of spatial, complex and dynamic cellular systems. One of the crucial challenges for the study of such systems is the design of a well stuctured and optimized workflow to provide a systematic and efficient hypothesis verification. Computer Science can efficiently address this task by providing software that facilitates handling, analysis, and evaluation of biological data to the benefit of experimenters and modelers. Results The Spatio-Temporal Simulation Environment (STSE is a set of open-source tools provided to conduct spatio-temporal simulations in discrete structures based on microscopy images. The framework contains modules to digitize, represent, analyze, and mathematically model spatial distributions of biochemical species. Graphical user interface (GUI tools provided with the software enable meshing of the simulation space based on the Voronoi concept. In addition, it supports to automatically acquire spatial information to the mesh from the images based on pixel luminosity (e.g. corresponding to molecular levels from microscopy images. STSE is freely available either as a stand-alone version or included in the linux live distribution Systems Biology Operational Software (SB.OS and can be downloaded from http://www.stse-software.org/. The Python source code as well as a comprehensive user manual and video tutorials are also offered to the research community. We discuss main concepts

  15. STSE: Spatio-Temporal Simulation Environment Dedicated to Biology.

    Science.gov (United States)

    Stoma, Szymon; Fröhlich, Martina; Gerber, Susanne; Klipp, Edda

    2011-04-28

    Recently, the availability of high-resolution microscopy together with the advancements in the development of biomarkers as reporters of biomolecular interactions increased the importance of imaging methods in molecular cell biology. These techniques enable the investigation of cellular characteristics like volume, size and geometry as well as volume and geometry of intracellular compartments, and the amount of existing proteins in a spatially resolved manner. Such detailed investigations opened up many new areas of research in the study of spatial, complex and dynamic cellular systems. One of the crucial challenges for the study of such systems is the design of a well stuctured and optimized workflow to provide a systematic and efficient hypothesis verification. Computer Science can efficiently address this task by providing software that facilitates handling, analysis, and evaluation of biological data to the benefit of experimenters and modelers. The Spatio-Temporal Simulation Environment (STSE) is a set of open-source tools provided to conduct spatio-temporal simulations in discrete structures based on microscopy images. The framework contains modules to digitize, represent, analyze, and mathematically model spatial distributions of biochemical species. Graphical user interface (GUI) tools provided with the software enable meshing of the simulation space based on the Voronoi concept. In addition, it supports to automatically acquire spatial information to the mesh from the images based on pixel luminosity (e.g. corresponding to molecular levels from microscopy images). STSE is freely available either as a stand-alone version or included in the linux live distribution Systems Biology Operational Software (SB.OS) and can be downloaded from http://www.stse-software.org/. The Python source code as well as a comprehensive user manual and video tutorials are also offered to the research community. We discuss main concepts of the STSE design and workflow. We

  16. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  17. Comparative Study of the Effectiveness of Three Learning Environments: Hyper-Realistic Virtual Simulations, Traditional Schematic Simulations and Traditional Laboratory

    Science.gov (United States)

    Martinez, Guadalupe; Naranjo, Francisco L.; Perez, Angel L.; Suero, Maria Isabel; Pardo, Pedro J.

    2011-01-01

    This study compared the educational effects of computer simulations developed in a hyper-realistic virtual environment with the educational effects of either traditional schematic simulations or a traditional optics laboratory. The virtual environment was constructed on the basis of Java applets complemented with a photorealistic visual output.…

  18. Moderator temperature coefficient in BWR core

    International Nuclear Information System (INIS)

    Naito, Yoshitaka

    1977-01-01

    Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

  19. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  20. Delivering high performance BWR fuel reliably

    International Nuclear Information System (INIS)

    Schardt, J.F.

    1998-01-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  1. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  2. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  3. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  4. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  5. Studies of fragileness in steels of vessels of BWR reactors

    International Nuclear Information System (INIS)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2003-01-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

  6. Study of transient turbine shot without bypass in a BWR

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  7. Impact of advanced BWR core physics method on BWR core monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

    2000-07-01

    Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

  8. The Development and Evaluation of a Computer-Simulated Science Inquiry Environment Using Gamified Elements

    Science.gov (United States)

    Tsai, Fu-Hsing

    2018-01-01

    This study developed a computer-simulated science inquiry environment, called the Science Detective Squad, to engage students in investigating an electricity problem that may happen in daily life. The environment combined the simulation of scientific instruments and a virtual environment, including gamified elements, such as points and a story for…

  9. WinGraphics: An optimized windowing environment for interactive real-time simulations

    International Nuclear Information System (INIS)

    Verboncoeur, J.P.; Vahedi, V.

    1989-01-01

    We have developed a customized windowing environment, Win Graphics, which provides particle simulation codes with an interactive user interface. The environment supports real-time animation of the simulation, displaying multiple diagnostics as they evolve in time. In addition, keyboard and printer (PostScript and dot matrix) support is provided. This paper describes this environment

  10. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm2)

    International Nuclear Information System (INIS)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J.

    1999-01-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm 2 ). (Author)

  11. Modes of Disintegration of Solid Foods in Simulated Gastric Environment

    Science.gov (United States)

    Kong, Fanbin

    2009-01-01

    A model stomach system was used to investigate disintegration of various foods in simulated gastric environment. Food disintegration modes and typical disintegration profiles are summarized in this paper. Mechanisms contributing to the disintegration kinetics of different foods were investigated as related to acidity, temperature, and enzymatic effect on the texture and changes in microstructure. Food disintegration was dominated by either fragmentation or erosion, depending on the physical forces acting on food and the cohesive force within the food matrix. The internal cohesive forces changed during digestion as a result of water penetration and acidic and enzymatic hydrolysis. When erosion was dominant, the disintegration data (weight retention vs. disintegration time) may be expressed with exponential, sigmoidal, and delayed-sigmoidal profiles. The different profiles are the result of competition among the rates of water absorption, texture softening, and erosion. A linear-exponential equation was used to describe the different disintegration curves with good fit. Acidity and temperature of gastric juice showed a synergistic effect on carrot softening, while pepsin was the key factor in disintegrating high-protein foods. A study of the change of carrot microstructure during digestion indicated that degradation of the pectin and cell wall was responsible for texture softening that contributed to the sigmoidal profile of carrot disintegration. PMID:20401314

  12. Simulated learning environment (SLE) in audiology education: A systematic review.

    Science.gov (United States)

    Dzulkarnain, Ahmad Aidil Arafat; Wan Mhd Pandi, Wan Mahirah; Rahmat, Sarah; Zakaria, Nur 'Azzah

    2015-01-01

    To systematically review the relevant peer-review literature investigating the outcome of simulated learning environment (SLE) training in audiology education. A systematic review research design. Fifteen databases were searched with four studies meeting the inclusion criteria. Three of the four studies revealed positive findings for the use of an SLE (that is, the SLE group showed a higher post-training score compared to the traditional training group or a significantly higher post-training score than the non-training groups). One study revealed negative findings where the traditional training group showed a significantly higher post-training score than the SLE group. In addition, both the studies comparing post- and pre-training scores reported significantly higher post-training scores than the pre-training scores of the participants that underwent SLE training. Overall, this review supports the notions that SLE training is an effective learning tool and can be used for basic clinical training. This conclusion should be treated with caution, considering the limited numbers of studies published in this area and future research should be conducted to cope with the gaps highlighted in this review.

  13. Thermal System Upgrade of the Space Environment Simulation Test Chamber

    Science.gov (United States)

    Desai, Ashok B.

    1997-01-01

    The paper deals with the refurbishing and upgrade of the thermal system for the existing thermal vacuum test facility, the Space Environment Simulator, at NASA's Goddard Space Flight Center. The chamber is the largest such facility at the center. This upgrade is the third phase of the long range upgrade of the chamber that has been underway for last few years. The first phase dealt with its vacuum system, the second phase involved the GHe subsystem. The paper describes the considerations of design philosophy options for the thermal system; approaches taken and methodology applied, in the evaluation of the remaining "life" in the chamber shrouds and related equipment by conducting special tests and studies; feasibility and extent of automation, using computer interfaces and Programmable Logic Controllers in the control system and finally, matching the old components to the new ones into an integrated, highly reliable and cost effective thermal system for the facility. This is a multi-year project just started and the paper deals mainly with the plans and approaches to implement the project successfully within schedule and costs.

  14. FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE

    International Nuclear Information System (INIS)

    1993-01-01

    1 - Description of test facility: BWR/6-218 standard plant. A full size bundle with electrically heated rods is used to simulate the reactor core. A scaling ratio of 1/624 is applied in the design of the system components. Key features of the FIST facility include: (1) Full height test vessel and internals; (2) correctly scaled fluid volume distribution; (3) simulation of ECCS, S/RV, and ADS; (4) level trip capability; (5) heated feedwater supply system, which provides the capability for steady state operation. 2 - Description of test: Test 6IB1 investigates system responses to an intermediate break in the recirculation suction line. BWR system licensing evaluations for various size recirculation break LOCA's indicates that a break size of about 0.2 sq.ft., without LPCS operation, is the highest PCT case for the intermediate break LOCA. Test 6IB1 simulates this event

  15. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  16. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  17. Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

    1996-01-01

    A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

  18. Full immersion simulation: validation of a distributed simulation environment for technical and non-technical skills training in Urology.

    Science.gov (United States)

    Brewin, James; Tang, Jessica; Dasgupta, Prokar; Khan, Muhammad S; Ahmed, Kamran; Bello, Fernando; Kneebone, Roger; Jaye, Peter

    2015-07-01

    To evaluate the face, content and construct validity of the distributed simulation (DS) environment for technical and non-technical skills training in endourology. To evaluate the educational impact of DS for urology training. DS offers a portable, low-cost simulated operating room environment that can be set up in any open space. A prospective mixed methods design using established validation methodology was conducted in this simulated environment with 10 experienced and 10 trainee urologists. All participants performed a simulated prostate resection in the DS environment. Outcome measures included surveys to evaluate the DS, as well as comparative analyses of experienced and trainee urologist's performance using real-time and 'blinded' video analysis and validated performance metrics. Non-parametric statistical methods were used to compare differences between groups. The DS environment demonstrated face, content and construct validity for both non-technical and technical skills. Kirkpatrick level 1 evidence for the educational impact of the DS environment was shown. Further studies are needed to evaluate the effect of simulated operating room training on real operating room performance. This study has shown the validity of the DS environment for non-technical, as well as technical skills training. DS-based simulation appears to be a valuable addition to traditional classroom-based simulation training. © 2014 The Authors BJU International © 2014 BJU International Published by John Wiley & Sons Ltd.

  19. A novel natural environment background model for Monte Carlo simulation and its application in the simulation of anticoincidence measurement.

    Science.gov (United States)

    Li, Sangang; Wang, Lei; Cheng, Yi; Tuo, Xianguo; Liu, Mingzhe; Yao, Fuliang; Leng, Fengqing; Cheng, Yuanyuan; Cai, Ting; Zhou, Yan

    2016-02-01

    This study proposes a novel natural environment background model by modeling brief environment conditions. It uses Geant4 program to simulate decays of (238)U, (232)Th, and (40)K in soil and obtains compositions of different-energy gamma rays in the natural environment background. The simulated gamma spectrum of the natural environment background agrees well with the experimental spectrum, particularly above 250 keV. The model is used in the simulation of anticoincidence measurement, indicating that the natural environment background can be decreased by approximately 88%, and the Compton attenuation factor is 2.22. The simulation of anticoincidence measurement can improve the minimum detectable activity (MDA) of the detection system. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (author)

  1. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

    2017-03-15

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  2. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    International Nuclear Information System (INIS)

    Badea, Aurelian F.; Cacuci, Dan G.

    2017-01-01

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  3. A novel natural environment background model for Monte Carlo simulation and its application in the simulation of anticoincidence measurement

    International Nuclear Information System (INIS)

    Li, Sangang; Wang, Lei; Cheng, Yi; Tuo, Xianguo; Liu, Mingzhe; Yao, Fuliang; Leng, Fengqing; Cheng, Yuanyuan; Cai, Ting; Zhou, Yan

    2016-01-01

    This study proposes a novel natural environment background model by modeling brief environment conditions. It uses Geant4 program to simulate decays of "2"3"8U, "2"3"2Th, and "4"0K in soil and obtains compositions of different-energy gamma rays in the natural environment background. The simulated gamma spectrum of the natural environment background agrees well with the experimental spectrum, particularly above 250 keV. The model is used in the simulation of anticoincidence measurement, indicating that the natural environment background can be decreased by approximately 88%, and the Compton attenuation factor is 2.22. The simulation of anticoincidence measurement can improve the minimum detectable activity (MDA) of the detection system. - Highlights: • This study proposes a novel natural environment background model by simulating decays of "2"3"8U, "2"3"2Th, and "4"0K in soil. • The simulated gamma spectrum of the natural environment background agrees well with the experimental spectrum, particularly above 250 keV. • The proposed environment background model is applied to study the properties of anticoincidence detector.

  4. Simulations of embodied evolving semiosis: Emergent semantics in artificial environments

    Energy Technology Data Exchange (ETDEWEB)

    Rocha, L.M.; Joslyn, C.

    1998-02-01

    As we enter this amazing new world of artificial and virtual systems and environments in the context of human communities, we are interested in the development of systems and environments which have the capacity to grow and evolve their own meanings in the context of this community of interaction. In this paper the authors analyze the necessary conditions to achieve systems and environments with these properties: (1) a coupled interaction between a system and its environment; (2) an environment with sufficient initial richness and structure to allow for; (3) embodied emergent classification of that environment system coupling; and (4) which is subject to pragmatic selection.

  5. High Fidelity Simulation of Littoral Environments: Applications and Coupling of Participating Models

    National Research Council Canada - National Science Library

    Allard, Richard

    2003-01-01

    The High Fidelity Simulation of Littoral Environments (HFSoLE) Challenge Project (C75) encompasses a suite of seven oceanographic models capable of exchanging information in a physically meaningful sense across the littoral environment...

  6. BioNessie - a grid enabled biochemical networks simulation environment

    OpenAIRE

    Liu, X.; Jiang, J.; Ajayi, O.; Gu, X.; Gilbert, D.; Sinnott, R.O.

    2008-01-01

    The simulation of biochemical networks provides insight and understanding about the underlying biochemical processes and pathways used by cells and organisms. BioNessie is a biochemical network simulator which has been developed at the University of Glasgow. This paper describes the simulator and focuses in particular on how it has been extended to benefit from a wide variety of high performance compute resources across the UK through Grid technologies to support larger scale simulations.

  7. Comparison of portable oxygen concentrators in a simulated airplane environment.

    Science.gov (United States)

    Fischer, Rainald; Wanka, Eva R; Einhaeupl, Franziska; Voll, Klaus; Schiffl, Helmut; Lang, Susanne M; Gruss, Martin; Ferrari, Uta

    2013-01-01

    Portable oxygen concentrators (POC) are highly desirable for patients with lung disease traveling by airplane, as these devices allow theoretically much higher travel times if additional batteries can be used. However, it is unclear whether POCs produce enough oxygen in airplanes at cruising altitude, even if complying with aviation regulations. We evaluated five frequently used POCs (XPO2 (Invacare, USA), Freestyle (AirSep C., USA), Evergo (Philipps Healthcare, Germany), Inogen One (Inogen, USA), Eclipse 3 (Sequal, USA)) at an altitude of 2650 m (as simulated airplane environment) in 11 patients with chronic obstructive lung disease (COPD) and compared theses POCs with the standard oxygen system (WS120, EMS Ltd., Germany) used by Lufthansa. Oxygen was delivered by each POC for 30 min to each patient at rest, blood gases were then drawn from the arterialized ear lobe. All POCs were able to deliver enough oxygen to increase the PaO(2) of our subjects by at least 1.40 kPa (10 mmHg). However, to achieve this increase, the two most lightweight POCs (Freestyle and Invacare XPO2) had to be run at their maximum level. This causes a significant reduction of battery life. The three other POCs (EverGo, Inogen One, Eclipse 3) and the WS120 were able to increase the PaO(2) by more than 2.55 kPa (20 mmHg), which provides extra safety for patients with more severe basal hypoxemia. When choosing the right oxygen system for air travel in patients in COPD, not only weight, but also battery life and maximum possible oxygen output must be considered carefully. Copyright © 2012 Elsevier Ltd. All rights reserved.

  8. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  9. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

  10. Simulation training tools for nonlethal weapons using gaming environments

    Science.gov (United States)

    Donne, Alexsana; Eagan, Justin; Tse, Gabriel; Vanderslice, Tom; Woods, Jerry

    2006-05-01

    Modern simulation techniques have a growing role for evaluating new technologies and for developing cost-effective training programs. A mission simulator facilitates the productive exchange of ideas by demonstration of concepts through compellingly realistic computer simulation. Revolutionary advances in 3D simulation technology have made it possible for desktop computers to process strikingly realistic and complex interactions with results depicted in real-time. Computer games now allow for multiple real human players and "artificially intelligent" (AI) simulated robots to play together. Advances in computer processing power have compensated for the inherent intensive calculations required for complex simulation scenarios. The main components of the leading game-engines have been released for user modifications, enabling game enthusiasts and amateur programmers to advance the state-of-the-art in AI and computer simulation technologies. It is now possible to simulate sophisticated and realistic conflict situations in order to evaluate the impact of non-lethal devices as well as conflict resolution procedures using such devices. Simulations can reduce training costs as end users: learn what a device does and doesn't do prior to use, understand responses to the device prior to deployment, determine if the device is appropriate for their situational responses, and train with new devices and techniques before purchasing hardware. This paper will present the status of SARA's mission simulation development activities, based on the Half-Life gameengine, for the purpose of evaluating the latest non-lethal weapon devices, and for developing training tools for such devices.

  11. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

    1992-01-01

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  12. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

  13. Development of BWR computerized operator support system for emergency conditions

    International Nuclear Information System (INIS)

    Murata, F.

    1984-01-01

    A BWR computerized operator support system (COSS) for emergency conditions has been under development for three years. The conceptual design of the system has been settled and some of the subsystems are in the detailed design or manufacturing stage. The principal functions are technical specification monitoring, diagnosis, guidance during emergency conditions, predictive simulation and safety monitoring. Before a reactor trip, alternative operational guidance for anomalous events is provided by utilization of the CTT (cause consequence tree) and FPS (failure propagation simulator). After the trip, operational guidance is based on event-oriented and symptom-oriented methods in association with the safety function monitor. The technical specification monitor controls the readiness monitor and performs surveillance tests of safety systems to maintain plant operational reliability and to ensure correct performance when initiated. The predictive simulator gives the future trends of significant plant parameters. These subsystems are expected to assist the operational personnel. The feasibility of the COSS functions is confirmed separately by off-line simulation. The paper considers the conceptual design, the functions of the subsystems and the off-line simulation results. Each subsystem has shown that useful information to operational personnel is provided. Henceforth these functions will be integrated into a single system and the feasibility will be thoroughly evaluated using a plant simulator which is being separately developed to verify the COSS. (author)

  14. BWR Mark I pressure suppression study: bench mark experiments

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1977-01-01

    Computer simulations representative of the wetwell of Mark I BWR's have predicted pressures and related phenomena. However, calculational predictions for purposes of engineering decision will be possible only if the code can be verified, i.e., shown to compute in accord with measured values. Described in the report is a set of single downcomer spherical flask bench mark experiments designed to produce quantitative data to validate various air-water dynamic computations; the experiments were performed since relevant bench mark data were not available from outside sources. Secondary purposes of the study were to provide a test bed for the instrumentation and post-experiment data processing techniques to be used in the Laboratory's reactor safety research program and to provide additional masurements for the air-water scaling study

  15. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  16. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

    1986-01-01

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  17. BWR mechanics and materials technology update

    International Nuclear Information System (INIS)

    Kiss, E.

    1983-01-01

    This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration. (orig.)

  18. Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

    International Nuclear Information System (INIS)

    Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

    2004-01-01

    Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

  19. Pressurized water reactor simulation in the training environment

    International Nuclear Information System (INIS)

    Wills, A.G.

    1990-01-01

    The paper gives a brief history of PWR Simulation within the DNST and an outline of the training courses leading to the requirement for the Display Array Simulation System. Focus is then placed upon the flexible use of real time simulation in the teaching of plant dynamics by the use of model generated data. The use of interactive consoles and a large scale colour graphic display has led to the success of the Display Array Simulation System within the DNST. Realisation of the potential of the system has led to many other proposed uses for the installed system and the paper concludes by discussing some of these. (orig./DG)

  20. Initial Development of a Quadcopter Simulation Environment for Auralization

    Science.gov (United States)

    Christian, Andrew; Lawrence, Joseph

    2016-01-01

    This paper describes a recently created computer simulation of quadcopter flight dynamics for the NASA DELIVER project. The goal of this effort is to produce a simulation that includes a number of physical effects that are not usually found in other dynamics simulations (e.g., those used for flight controller development). These effects will be shown to have a significant impact on the fidelity of auralizations - entirely synthetic time-domain predictions of sound - based on this simulation when compared to a recording. High-fidelity auralizations are an important precursor to human subject tests that seek to understand the impact of vehicle configurations on noise and annoyance.

  1. Simulation Assessment Validation Environment (SAVE). Software User’s Manual

    Science.gov (United States)

    2000-09-01

    TOOL USAGE Figure 2-14: Factory Simulation Tool Usage This tool directly emulates real-world system behaviors that are associated with each resource...manufacturing simulation tools and Computer Aided Design (CAD) tools. The Factory/Schedule Simulation tool is used to simulate real-world system behaviors ...char Name>_pfchar Part Usage(Produced)/Part/Feature/Char <partuseName>_<partName>_<matName>_<char Name>_cmat PartUsage( Comsumed )/Part/Material

  2. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm{sup 2}); Comportamiento a la fractura de un acero inoxidable AISI 304 sensibilizado en condiciones de reactor BWR (288 grados Centigrados y 80 Kg/cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm{sup 2}). (Author)

  3. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  4. Delivering high performance BWR fuel reliably

    Energy Technology Data Exchange (ETDEWEB)

    Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

    1998-07-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  5. PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

    OpenAIRE

    MELARA SAN ROMÁN, JOSÉ

    2016-01-01

    [EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

  6. Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

    International Nuclear Information System (INIS)

    Pruitt, D.W.

    1992-01-01

    This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

  7. Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

    International Nuclear Information System (INIS)

    Ivanov, A.; Sanchez, V.; Hoogenboom, J. E.

    2012-01-01

    As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

  8. Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, A.; Sanchez, V. [Karlsruhe Inst. of Technology, Inst. for Neutron Physics and Reactor Technology, Herman-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Hoogenboom, J. E. [Delft Univ. of Technology, Faculty of Applied Sciences, Mekelweg 15, 2629 JB Delft (Netherlands)

    2012-07-01

    As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

  9. Propagation of cracks by stress corrosion in conditions of BWR type reactor

    International Nuclear Information System (INIS)

    Merino C, F.J.; Fuentes C, P.

    2004-01-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  10. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  11. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  12. Economic analysis of hydride fueled BWR

    International Nuclear Information System (INIS)

    Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

    2009-01-01

    The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

  13. Investigation of BWR stability in Forsmark 2

    International Nuclear Information System (INIS)

    Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

    1988-01-01

    A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  14. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  15. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  16. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  17. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1982-01-01

    A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  18. Launch Environment Water Flow Simulations Using Smoothed Particle Hydrodynamics

    Science.gov (United States)

    Vu, Bruce T.; Berg, Jared J.; Harris, Michael F.; Crespo, Alejandro C.

    2015-01-01

    This paper describes the use of Smoothed Particle Hydrodynamics (SPH) to simulate the water flow from the rainbird nozzle system used in the sound suppression system during pad abort and nominal launch. The simulations help determine if water from rainbird nozzles will impinge on the rocket nozzles and other sensitive ground support elements.

  19. Sensitiaztion of austenitic stainless steels and its significance as regards stress-corrosion cracking of BWR pipe systems

    International Nuclear Information System (INIS)

    Roberts, W.; Otterberg, R.

    1984-05-01

    A critical literature evaluation dealing with sensitization of austenitic stainless steels and its importance in the context of intergranular stress-corrosion cracking (IGSCC) in high-temperature, oxygenated water is presented. The factors influencing the degree of sensitization are discussed, principally for type-304 stainless steels, both as regards sensitization arising as a result of isothermal holding within the critical temperature range and weld sensitization. The phenomenon of low-temperature sensitization is described and its potential significance under BWR operating conditions speculated upon. The principal features of and mechanisms controlling IGSCC of sensitized 304 steels in BWR-type environments are reviewed and some thoughts are given to the relevance of laboratory SCC testing in predicting the occurrence of cracking in actual BWR systems. Finally various countermeasures against IGSCC in existing and projected reactors are presented and discussed. (Author)

  20. BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

    International Nuclear Information System (INIS)

    Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

  1. Analysis of results of AZTRAN and AZKIND codes for a BWR; Analisis de resultados de los codigos AZTRAN y AZKIND para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M., E-mail: gbo729@yahoo.com.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  2. Evaluation and development the routing protocol of a fully functional simulation environment for VANETs

    Science.gov (United States)

    Ali, Azhar Tareq; Warip, Mohd Nazri Mohd; Yaakob, Naimah; Abduljabbar, Waleed Khalid; Atta, Abdu Mohammed Ali

    2017-11-01

    Vehicular Ad-hoc Networks (VANETs) is an area of wireless technologies that is attracting a great deal of interest. There are still several areas of VANETS, such as security and routing protocols, medium access control, that lack large amounts of research. There is also a lack of freely available simulators that can quickly and accurately simulate VANETs. The main goal of this paper is to develop a freely available VANETS simulator and to evaluate popular mobile ad-hoc network routing protocols in several VANETS scenarios. The VANETS simulator consisted of a network simulator, traffic (mobility simulator) and used a client-server application to keep the two simulators in sync. The VANETS simulator also models buildings to create a more realistic wireless network environment. Ad-Hoc Distance Vector routing (AODV), Dynamic Source Routing (DSR) and Dynamic MANET On-demand (DYMO) were initially simulated in a city, country, and highway environment to provide an overall evaluation.

  3. Discrete event simulation in an artificial intelligence environment: Some examples

    International Nuclear Information System (INIS)

    Roberts, D.J.; Farish, T.

    1991-01-01

    Several Los Alamos National Laboratory (LANL) object-oriented discrete-event simulation efforts have been completed during the past three years. One of these systems has been put into production and has a growing customer base. Another (started two years earlier than the first project) was completed but has not yet been used. This paper will describe these simulation projects. Factors which were pertinent to the success of the one project, and to the failure of the second project will be discussed (success will be measured as the extent to which the simulation model was used as originally intended). 5 figs

  4. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo a baja presion (LPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Membrillo G, O. E.; Chavez M, C., E-mail: garzo1012@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present work presents the modeling and simulation of the aspersion system to low pressure of reactor of the nuclear power plant of Laguna Verde using the nuclear code RELAP/SCDAP. The objective of the emergency systems inside a nuclear reactor is the cooling of the core, nor caring the performance of any other emergency system in the case of an accident design base for coolant loss. To obtain a simulation of the system is necessary to have a model based on their main components, pipes, pumps, valves, etc. This article describes the model for the simulation of the main line and the test line for the HPCS. At the moment we have the simulation of the reactor vessel and their systems associated to the nuclear power plant of Laguna Verde, this work will allow to associate the emergency system model LPCS to the vessel model. The simulation of the vessel and the emergency systems will allow knowing the behavior of the reactor in the stage of the coolant loos, giving the possibility to analyze diverse scenarios. The general model will provide an auxiliary tool for the training in classroom and at distance in the operation of nuclear power plants. (Author)

  5. Simulating Nonmodel-Fitting Responses in a CAT Environment. ACT Research Report Series 98-10.

    Science.gov (United States)

    Yi, Qing; Nering, Michael L.

    This study developed a model to simulate nonmodel-fitting responses in a computerized adaptive testing (CAT) environment, and to examine the effectiveness of the model. The underlying idea was to simulate examinees' test behaviors realistically. This study simulated a situation in which examinees are exposed to or are coached on test items before…

  6. Simulations of depleted CMOS sensors for high-radiation environments

    CERN Document Server

    Liu, J.; Bhat, S.; Breugnon, P.; Caicedo, I.; Chen, Z.; Degerli, Y.; Godiot-Basolo, S.; Guilloux, F.; Hemperek, T.; Hirono, T.; Hügging, F.; Krüger, H.; Moustakas, K.; Pangaud, P.; Rozanov, A.; Rymaszewski, P.; Schwemling, P.; Wang, M.; Wang, T.; Wermes, N.; Zhang, L.

    2017-01-01

    After the Phase II upgrade for the Large Hadron Collider (LHC), the increased luminosity requests a new upgraded Inner Tracker (ITk) for the ATLAS experiment. As a possible option for the ATLAS ITk, a new pixel detector based on High Voltage/High Resistivity CMOS (HV/HR CMOS) technology is under study. Meanwhile, a new CMOS pixel sensor is also under development for the tracker of Circular Electron Position Collider (CEPC). In order to explore the sensor electric properties, such as the breakdown voltage and charge collection efficiency, 2D/3D Technology Computer Aided Design (TCAD) simulations have been performed carefully for the above mentioned both of prototypes. In this paper, the guard-ring simulation for a HV/HR CMOS sensor developed for the ATLAS ITk and the charge collection efficiency simulation for a CMOS sensor explored for the CEPC tracker will be discussed in details. Some comparisons between the simulations and the latest measurements will also be addressed.

  7. Valuation of power oscillations in a BWR after control rod banks withdrawal events

    International Nuclear Information System (INIS)

    Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

    2009-01-01

    The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

  8. Comparative study of the effectiveness of three learning environments: Hyper-realistic virtual simulations, traditional schematic simulations and traditional laboratory

    Directory of Open Access Journals (Sweden)

    Maria Isabel Suero

    2011-10-01

    Full Text Available This study compared the educational effects of computer simulations developed in a hyper-realistic virtual environment with the educational effects of either traditional schematic simulations or a traditional optics laboratory. The virtual environment was constructed on the basis of Java applets complemented with a photorealistic visual output. This new virtual environment concept, which we call hyper-realistic, transcends basic schematic simulation; it provides the user with a more realistic perception of a physical phenomenon being simulated. We compared the learning achievements of three equivalent, homogeneous groups of undergraduates—an experimental group who used only the hyper-realistic virtual laboratory, a first control group who used a schematic simulation, and a second control group who used the traditional laboratory. The three groups received the same theoretical preparation and carried out equivalent practicals in their respective learning environments. The topic chosen for the experiment was optical aberrations. An analysis of variance applied to the data of the study demonstrated a statistically significant difference (p value <0.05 between the three groups. The learning achievements attained by the group using the hyper-realistic virtual environment were 6.1 percentage points higher than those for the group using the traditional schematic simulations and 9.5 percentage points higher than those for the group using the traditional laboratory.

  9. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, D.; Chavez M, C., E-mail: danmirnyi@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  10. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    Behrooz, A.

    2008-01-01

    for a large reactor could reach 20 gigabytes) that it is not possible to load into RAM memory of an operating system with 32 bit architecture. A special procedure has been developed within the MATLAB environment to remove this memory limitation, and to invert such large matrices and finally obtain the reactor transfer functions that enable the study of system stability. Various applications of the present frequency-domain code to a typical BWR fuel assembly, a BWR core, and to a chemical reactor showed a good agreement with reference results. (author)

  11. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  12. Experimental and numerical investigations of BWR fuel bundle inlet flow

    International Nuclear Information System (INIS)

    Hoashi, E; Morooka, S; Ishitori, T; Komita, H; Endo, T; Honda, H; Yamamoto, T; Kato, T; Kawamura, S

    2009-01-01

    We have been studying the mechanism of the flow pattern near the fuel bundle inlet of BWR using both flow visualization test and computational fluid dynamics (CFD) simulation. In the visualization test, both single- and multi-bundle test sections were used. The former test section includes only a corner orifice facing two support beams and the latter simulates 16 bundles surrounded by four beams. An observation window is set on the side of the walls imitating the support beams upstream of the orifices in both test sections. In the CFD simulation, as well as the visualization test, the single-bundle model is composed of one bundle with a corner orifice and the multi-bundle model is a 1/4 cut of the test section that includes 4 bundles with the following four orifices: a corner orifice facing the corner of the two neighboring support beams, a center orifice at the opposite side from the corner orifice, and two side orifices. Twin-vortices were observed just upstream of the corner orifice in the multi-bundle test as well as the single-bundle test. A single-vortex and a vortex filament were observed at the side orifice inlet and no vortex was observed at the center orifice. These flow patterns were also predicted in the CFD simulation using Reynolds Stress Model as a turbulent model and the results were in good agreement with the test results mentioned above. (author)

  13. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  14. GPE-BWR and the containment venting and filtering issue

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

    The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

  15. Multiscale simulation of molecular processes in cellular environments.

    Science.gov (United States)

    Chiricotto, Mara; Sterpone, Fabio; Derreumaux, Philippe; Melchionna, Simone

    2016-11-13

    We describe the recent advances in studying biological systems via multiscale simulations. Our scheme is based on a coarse-grained representation of the macromolecules and a mesoscopic description of the solvent. The dual technique handles particles, the aqueous solvent and their mutual exchange of forces resulting in a stable and accurate methodology allowing biosystems of unprecedented size to be simulated.This article is part of the themed issue 'Multiscale modelling at the physics-chemistry-biology interface'. © 2016 The Author(s).

  16. Cooperative visualization and simulation in a supercomputer environment

    International Nuclear Information System (INIS)

    Ruehle, R.; Lang, U.; Wierse, A.

    1993-01-01

    The article takes a closer look on the requirements being imposed by the idea to integrate all the components into a homogeneous software environment. To this end several methods for the distribtuion of applications in dependence of certain problem types are discussed. The currently available methods at the University of Stuttgart Computer Center for the distribution of applications are further explained. Finally the aims and characteristics of a European sponsored project, called PAGEIN, are explained, which fits perfectly into the line of developments at RUS. The aim of the project is to experiment with future cooperative working modes of aerospace scientists in a high speed distributed supercomputing environment. Project results will have an impact on the development of real future scientific application environments. (orig./DG)

  17. Development of a BWR loading pattern design system based on modified genetic algorithms and knowledge

    International Nuclear Information System (INIS)

    Martin-del-Campo, Cecilia; Francois, Juan Luis; Avendano, Linda; Gonzalez, Mario

    2004-01-01

    An optimization system based on Genetic Algorithms (GAs), in combination with expert knowledge coded in heuristics rules, was developed for the design of optimized boiling water reactor (BWR) fuel loading patterns. The system was coded in a computer program named Loading Pattern Optimization System based on Genetic Algorithms, in which the optimization code uses GAs to select candidate solutions, and the core simulator code CM-PRESTO to evaluate them. A multi-objective function was built to maximize the cycle energy length while satisfying power and reactivity constraints used as BWR design parameters. Heuristic rules were applied to satisfy standard fuel management recommendations as the Control Cell Core and Low Leakage loading strategies, and octant symmetry. To test the system performance, an optimized cycle was designed and compared against an actual operating cycle of Laguna Verde Nuclear Power Plant, Unit I

  18. Experimental study on reduced moderation BWR with Advanced Recycle System (BARS)

    International Nuclear Information System (INIS)

    Hiraiwa, K.; Yoshioka, K.; Yamamoto, Y.; Akiba, M.; Yamaoka, M.; Abe, N.; Mimatsu, J.

    2004-01-01

    Experimental study has been done for reduced-moderation spectrum boiling water reactor named BARS (BWR with Advanced Recycle System). The critical assembly experiment for triangular tight uranium lattice has been done in TOSHIBA critical assembly (NCA). Experimental method based on modified conversion ratio was adopted to evaluate the void reactivity effect. Void fraction was simulated by formed polystyrene in this experiment. The measured void coefficient for tight uranium lattice agreed with calculation. The thermal hydraulic test study has been done to study the coolability of BARS lattice. Visual test and high-pressure thermal hydraulic test have been done as the thermal hydraulic test. Visual test has indicated the flow behavior for BARS lattice is same as that of current BWR. The high-pressure thermal hydraulic test has indicated the applicability of modified Arai's correlation to the BARS lattice. (authors)

  19. Development of membrane moisture separator for BWR off-gas system

    International Nuclear Information System (INIS)

    Ogata, H.; Kawamura, S.; Kumasaka, M.; Nishikubo, M.

    2001-01-01

    In BWR plant off-gas treatment systems, dehumidifiers are used to maintain noble gas adsorption efficiency in the first half of the charcoal hold-up units. From the perspective of simplifying and reducing the cost of such a dehumidification system, Japanese BWR utilities and plant fabricators have been developing a dehumidification system employing moisture separation membrane of the type already proven in fields such as medical instrumentation and precision measuring apparatus. The first part of this development involved laboratory testing to simulate the conditions found in an actual off-gas system, the results of which demonstrated satisfactory results in terms of moisture separation capability and membrane durability, and suggested favorable prospects for application in actual off-gas systems. Further, in-plant testing to verify moisture separation capability and membrane durability in the presence of actual gases is currently underway, with results so far suggesting that the system is capable of obtaining good moisture separation capability. (author)

  20. Analysis of void reactivity measurements in full MOX BWR physics experiments

    International Nuclear Information System (INIS)

    Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

    2008-01-01

    In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

  1. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    International Nuclear Information System (INIS)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

    2013-01-01

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  2. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de inyeccion de agua de refrigeracion a baja presion (LPCI) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C., E-mail: renedelgado2015@hotmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

  3. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    International Nuclear Information System (INIS)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

  4. Analysis of radiological consequences in a typical BWR with a mark-II containment

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro

    2003-01-01

    INS/NUPEC in Japan has been carrying out the Level 3 PSA program. In the program, the MACCS2 code has been extensively applied to analyze radiological consequences for typical BWR and PWR plants in Japan. The present study deals with analysis of effects of the AMs, which were implemented by industries, on radiological consequence for a typical BWR with a Mark-II containment. In the present study, source terms and their frequencies of source terms were used based on results of Level 2 PSA taking into account AM countermeasures. Radiological consequences were presented with dose risks (Sv/ry), which were multiplied doses (Sv) by containment damage frequencies (/ry), and timing of radionuclides release to the environment. The results of the present study indicated that the dose risks became negligible in most cases taking AM countermeasures and evacuations. (author)

  5. Simulation of maize growth under conservation farming in tropical environments.

    NARCIS (Netherlands)

    Stroosnijder, L.; Kiepe, P.

    1998-01-01

    This book is written for students and researchers with a keen interest in the quantification of the field soil water balance in tropical environments and the effect of conservation farming on crop production. Part 1 deals with the potential production, i.e. crop growth under ample supply of water

  6. Improved climate risk simulations for rice in arid environments

    NARCIS (Netherlands)

    Oort, van P.A.J.; Vries, de M.; Yoshida, H.; Saito, K.

    2015-01-01

    We integrated recent research on cardinal temperatures for phenology and early leaf growth, spikelet formation, early morning flowering, transpirational cooling, and heat- and cold-induced sterility into an existing to crop growth model ORYZA2000. We compared for an arid environment observed

  7. Simulation of indoor environment in low energy housing

    DEFF Research Database (Denmark)

    Vagiannis, Georgios; Knudsen, Henrik N.; Toftum, Jørn

    2012-01-01

    was selected and sensitivity analyses were conducted for the importance of occupancy, ventilation, window opening, and heat recovery efficiency. In particular occupancy and venting played significant roles for the indoor environment and energy consumption. It was also shown that with passive measures, but also...

  8. Prevention of organic iodide formation in BWR's

    International Nuclear Information System (INIS)

    Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

    1996-01-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR's as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs

  9. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

    2005-01-01

    Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

  10. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Alessandro Petruzzi; Shin Chin; Kostadin Ivanov; Asok Ray; Fan-Bill Cheung

    2005-01-01

    Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

  11. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    Shack, W.J.; Kassner, T.F.; Maiya, P.S.; Park, J.Y.; Ruther, W.; Kuzay, T.; Rybicki, E.F.; Stonesifer, R.B.

    1988-01-01

    Piping in light-water-reactor power systems has been affected by several types of environmental degradation. This paper presents results from studies of (1) stress corrosion crack growth in fracture mechanics specimens of modified Type 347 SS and Type 304/308L SS weld overlay material, (2) heat-to-heat variations in stress corrosion cracking (SCC) of Types 316NG and 347 SS, (3) SCC of sensitized Type 304 SS in water with cupric ion or organic acid impurities, (4) electrochemical potential (ECP) measurements under gamma irradiation, (5) SCC of ferritic steels, (6) strain-controlled fatigue of Type 316NG SS in air at ambient temperature, and (7) through-wall residual stress measurements and finite-element calculation of residual stresses in weldments treated by a mechanical stress improvement process (MSIP). Fracture-mechanics crack-growth-rate tests on Type 316NG SS have shown that transgranular cracking can occur even in high purity environments, whereas no crack growth was observed in Type 347 SS even in impurity environments. In tests on weld overlay specimens, no cracks penetrated into the overlay even in impurity environments. Instead, the cracks branched when they approached the overlay, and then grew parallel to interface. In SCC tests on sensitized Type 304 SS, cupric ions at concentrations greater than ∼1 ppm were found to be deleterious, whereas organic acids at this concentration were not detrimental. Tests on several ferritic steels indicate a strong correlation between the sulfur content of the steels and susceptibility to SCC. External gamma radiation fields produced a large positive shift in the ECP of Type 304 SS at low dissolved-oxygen concentrations (<5 ppb), whereas in the absence of an external gamma field there was no difference in the ECP values of irradiated and nonirradiated material. Fatigue data for Type 316NG SS are consistent with the ASME code mean curve at high strains, but fall below the curve at low strains. Calculations of the

  12. A simplified spatial model for BWR stability

    International Nuclear Information System (INIS)

    Berman, Y.; Lederer, Y.; Meron, E.

    2012-01-01

    A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

  13. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  14. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  15. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  16. Computer simulation of defect behavior under fusion irradiation environments

    International Nuclear Information System (INIS)

    Muroga, T.; Ishino, S.

    1983-01-01

    To simulate defect behavior under irradiation, three kinds of cascade-annealing calculations have been carried out in alpha-iron using the codes MARLOWE, DAIQUIRI and their modifications. They are (1) cascade-annealing calculation with different masses of projectile, (2) defect drifting near dislocations after cascade production and (3) cascade-overlap calculation. The defect survival ratio is found to increase as decreasing mass of the projectile both after athermal close-pair recombination and after thermal annealing. It is shown that at moderate temperatures vacancy clustering is enhanced near dislocations. Cascade-overlap is found to decrease the defect survivability. In addition, the role of helium in vacancy clustering has been calculated in aluminium lattices and its effect is found to depend strongly on temperature, interstitials and the mobility of small clusters. These results correspond well to the experimental data and will be helpful for correlating between fusion and simulation irradiations. (orig.)

  17. 3D hybrid simulation of the Titan's plasma environment

    Science.gov (United States)

    Lipatov, Alexander; Sittler, Edward, Jr.; Hartle, Richard

    2007-11-01

    Titan plays an important role as a simulation laboratory for multiscale kinetic plasma processes which are key processes in space and laboratory plasmas. A development of multiscale combined numerical methods allows us to use more realistic plasma models at Titan. In this report, we describe a Particle-Ion--Fluid-Ion--Fluid--Electron method of kinetic ion-neutral simulation code. This method takes into account charge-exchange and photoionization processes. The model of atmosphere of Titan was based on a paper by Sittler, Hartle, Vinas et al., [2005]. The background ions H^+, O^+ and pickup ions H2^+, CH4^+ and N2^+ are described in a kinetic approximation, where the electrons are approximated as a fluid. In this report we study the coupling between background ions and pickup ions on the multiple space scales determined by the ion gyroradiis. The first results of such a simulation of the dynamics of ions near Titan are discussed in this report and compared with recent measurements made by the Cassini Plasma Spectrometer (CAPS, [Hartle, Sittler et al., 2006]). E C Sittler Jr., R E Hartle, A F Vinas, R E Johnson, H T Smith and I Mueller-Wodarg, J. Geophys. Res., 110, A09302, 2005.R E Hartle, E C Sittler, F M Neubauer, R E Johnson, et al., Planet. Space Sci., 54, 1211, 2006.

  18. A COMPUTATIONAL WORKBENCH ENVIRONMENT FOR VIRTUAL POWER PLANT SIMULATION

    International Nuclear Information System (INIS)

    Mike Bockelie; Dave Swensen; Martin Denison

    2002-01-01

    This is the fifth Quarterly Technical Report for DOE Cooperative Agreement No: DE-FC26-00NT41047. The goal of the project is to develop and demonstrate a computational workbench for simulating the performance of Vision 21 Power Plant Systems. Within the last quarter, our efforts have become focused on developing an improved workbench for simulating a gasifier based Vision 21 energyplex. To provide for interoperability of models developed under Vision 21 and other DOE programs, discussions have been held with DOE and other organizations developing plant simulator tools to review the possibility of establishing a common software interface or protocol to use when developing component models. A component model that employs the CCA protocol has successfully been interfaced to our CCA enabled workbench. To investigate the software protocol issue, DOE has selected a gasifier based Vision 21 energyplex configuration for use in testing and evaluating the impacts of different software interface methods. A Memo of Understanding with the Cooperative Research Centre for Coal in Sustainable Development (CCSD) in Australia has been completed that will enable collaborative research efforts on gasification issues. Preliminary results have been obtained for a CFD model of a pilot scale, entrained flow gasifier. A paper was presented at the Vision 21 Program Review Meeting at NETL (Morgantown) that summarized our accomplishments for Year One and plans for Year Two and Year Three

  19. Applied environmetrics. Simulation applied to the physical environment

    Energy Technology Data Exchange (ETDEWEB)

    Beer, T

    1988-02-01

    Environmetrics is the application of quantitative methods to all aspects of the social and natural environment. This includes forecasting, mathematical modelling, data analysis, and statistics. Applied Environmetrics as a discipline involves the analysis of environmental data through the use of packaged, or specially designed computer software. Two case studies of recent implementations of applied environmetrics within the Australian mining industry are dealt with. 3 figs., 5 refs.

  20. High performance computing network for cloud environment using simulators

    OpenAIRE

    Singh, N. Ajith; Hemalatha, M.

    2012-01-01

    Cloud computing is the next generation computing. Adopting the cloud computing is like signing up new form of a website. The GUI which controls the cloud computing make is directly control the hardware resource and your application. The difficulty part in cloud computing is to deploy in real environment. Its' difficult to know the exact cost and it's requirement until and unless we buy the service not only that whether it will support the existing application which is available on traditional...

  1. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

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

  2. Training and learning for crisis management using a virtual simulation/gaming environment

    NARCIS (Netherlands)

    Walker, W.E.; Giddings, J.; Armstrong, S.

    2011-01-01

    Recent advances in computers, networking, and telecommunications offer new opportunities for using simulation and gaming as methodological tools for improving crisis management. It has become easy to develop virtual environments to support games, to have players at distributed workstations

  3. An Open-Source Simulation Environment for Model-Based Engineering, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed work is a new spacecraft simulation environment for model-based engineering of flight algorithms and software. The goal is to provide a much faster way...

  4. The Use of Computer Simulation to Compare Student performance in Traditional versus Distance Learning Environments

    Directory of Open Access Journals (Sweden)

    Retta Guy

    2015-06-01

    Full Text Available Simulations have been shown to be an effective tool in traditional learning environments; however, as distance learning grows in popularity, the need to examine simulation effectiveness in this environment has become paramount. A casual-comparative design was chosen for this study to determine whether students using a computer-based instructional simulation in hybrid and fully online environments learned better than traditional classroom learners. The study spans a period of 6 years beginning fall 2008 through spring 2014. The population studied was 281 undergraduate business students self-enrolled in a 200-level microcomputer application course. The overall results support previous studies in that computer simulations are most effective when used as a supplement to face-to-face lectures and in hybrid environments.

  5. A Virtual Simulation Environment for Lunar Rover: Framework and Key Technologies

    Directory of Open Access Journals (Sweden)

    Yan-chun Yang

    2008-06-01

    Full Text Available Lunar rover development involves a large amount of validation works in realistic operational conditions, including its mechanical subsystem and on-board software. Real tests require equipped rover platform and a realistic terrain. It is very time consuming and high cost. To improve the development efficiency, a rover simulation environment called RSVE that affords real time capabilities with high fidelity has been developed. It uses fractional Brown motion (fBm technique and statistical properties to generate lunar surface. Thus, various terrain models for simulation can be generated through changing several parameters. To simulate lunar rover evolving on natural and unstructured surface with high realism, the whole dynamics of the multi-body systems and complex interactions with soft ground is integrated in this environment. An example for path planning algorithm and controlling algorithm testing in this environment is tested. This simulation environment runs on PC or Silicon Graphics.

  6. A Virtual Simulation Environment for Lunar Rover: Framework and Key Technologies

    Directory of Open Access Journals (Sweden)

    Yan-chun Yang

    2008-11-01

    Full Text Available Lunar rover development involves a large amount of validation works in realistic operational conditions, including its mechanical subsystem and on-board software. Real tests require equipped rover platform and a realistic terrain. It is very time consuming and high cost. To improve the development efficiency, a rover simulation environment called RSVE that affords real time capabilities with high fidelity has been developed. It uses fractional Brown motion (fBm technique and statistical properties to generate lunar surface. Thus, various terrain models for simulation can be generated through changing several parameters. To simulate lunar rover evolving on natural and unstructured surface with high realism, the whole dynamics of the multi-body systems and complex interactions with soft ground is integrated in this environment. An example for path planning algorithm and controlling algorithm testing in this environment is tested. This simulation environment runs on PC or Silicon Graphics.

  7. Tutoring electronic troubleshooting in a simulated maintenance work environment

    Science.gov (United States)

    Gott, Sherrie P.

    1987-01-01

    A series of intelligent tutoring systems, or intelligent maintenance simulators, is being developed based on expert and novice problem solving data. A graded series of authentic troubleshooting problems provides the curriculum, and adaptive instructional treatments foster active learning in trainees who engage in extensive fault isolation practice and thus in conditionalizing what they know. A proof of concept training study involving human tutoring was conducted as a precursor to the computer tutors to assess this integrated, problem based approach to task analysis and instruction. Statistically significant improvements in apprentice technicians' troubleshooting efficiency were achieved after approximately six hours of training.

  8. Lithium-ion Battery Electrothermal Model, Parameter Estimation, and Simulation Environment

    Directory of Open Access Journals (Sweden)

    Simone Orcioni

    2017-03-01

    Full Text Available The market for lithium-ion batteries is growing exponentially. The performance of battery cells is growing due to improving production technology, but market request is growing even more rapidly. Modeling and characterization of single cells and an efficient simulation environment is fundamental for the development of an efficient battery management system. The present work is devoted to defining a novel lumped electrothermal circuit of a single battery cell, the extraction procedure of the parameters of the single cell from experiments, and a simulation environment in SystemC-WMS for the simulation of a battery pack. The electrothermal model of the cell was validated against experimental measurements obtained in a climatic chamber. The model is then used to simulate a 48-cell battery, allowing statistical variations among parameters. The different behaviors of the cells in terms of state of charge, current, voltage, or heat flow rate can be observed in the results of the simulation environment.

  9. MathModelica - An Extensible Modeling and Simulation Environment with Integrated Graphics and Literate Programming

    OpenAIRE

    Fritzson, Peter; Gunnarsson, Johan; Jirstrand, Mats

    2002-01-01

    MathModelica is an integrated interactive development environment for advanced system modeling and simulation. The environment integrates Modelica-based modeling and simulation with graphic design, advanced scripting facilities, integration of program code, test cases, graphics, documentation, mathematical type setting, and symbolic formula manipulation provided via Mathematica. The user interface consists of a graphical Model Editor and Notebooks. The Model Editor is a graphical user interfa...

  10. Fatigue cracking of alloy 600 in simulated steam generator crevice environment

    International Nuclear Information System (INIS)

    Ogundele, G.; Lepik, O.

    1998-01-01

    Investigations were carried out to generate fatigue life (S-N) and near-threshold fatigue crack propagation (da/dN) data to determine the environmental influence on fatigue behavior for Alloy 600 in air, deionized water and in simulated Bruce Nuclear Generating Station 'A' crevice environments under appropriate loading conditions. In the low cycle fatigue regime, the simulated crevice environment did not affect the fatigue life of Alloy 600 under the applied loading conditions. The near-threshold fatigue crack growth rates of Alloy 600 in the simulated crevice environment were significantly lower compared to either pure water or air environments and is believed to be the result of higher crack closure in the crevice environment. (author)

  11. Simulation of worst-case operating conditions for integrated circuits operating in a total dose environment

    International Nuclear Information System (INIS)

    Bhuva, B.L.

    1987-01-01

    Degradations in the circuit performance created by the radiation exposure of integrated circuits are so unique and abnormal that thorough simulation and testing of VLSI circuits is almost impossible, and new ways to estimate the operating performance in a radiation environment must be developed. The principal goal of this work was the development of simulation techniques for radiation effects on semiconductor devices. The mixed-mode simulation approach proved to be the most promising. The switch-level approach is used to identify the failure mechanisms and critical subcircuits responsible for operational failure along with worst-case operating conditions during and after irradiation. For precise simulations of critical subcircuits, SPICE is used. The identification of failure mechanisms enables the circuit designer to improve the circuit's performance and failure-exposure level. Identification of worst-case operating conditions during and after irradiation reduces the complexity of testing VLSI circuits for radiation environments. The results of test circuits for failure simulations using a conventional simulator and the new simulator showed significant time savings using the new simulator. The savings in simulation time proved to be circuit topology-dependent. However, for large circuits, the simulation time proved to be orders of magnitude smaller than simulation time for conventional simulators

  12. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  13. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

    Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

    2010-01-01

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  14. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  15. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  16. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  17. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  18. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  19. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  20. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  1. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions

  2. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  3. Retention Capability of Local Backfill Materials 1-Simulated Disposal Environment

    International Nuclear Information System (INIS)

    Ghattas, N.K.; Eskander, S.B.; El-Adham, K.A.; Mahmoud, N.S.

    2001-01-01

    In Egypt, a shallow ground disposal facility was the chosen option for the disposal of low and and intermediate radioactive wastes. The impact of the waste disposal facility on the environment depends on the nature of the barriers, which intend to limit and control contaminant migration. Owing to their physical, chemical and mechanical characteristics. Local soil materials were studied to illustrate the role of the back fill as part of an optimized safety multi-barrier system, which can provide the required level of protection of the environment and meet economic and regulatory requirements. A theoretical model was proposed to calculate the transport phenomena through the backfill materials. The credibility and validity of the proposed model was checked by the experimental results obtained from a three-arms arrangement system. The obtained data for the distribution coefficient (K d ) and the apparent diffusion coefficient (D a ) were in good agreement with those previously obtained in the literatures. Taking in consideration the prevailing initial conditions, the data calculated by the theoretical model applied show a reasonable agreement with the results obtained from experimental work. Prediction of radioactive cesium migration through the backfill materials using the proposed model was performed as a function of distance. The results obtained show that after 100 years, a fraction not exceeding 1E-9 of the original activity could be detected at 1m distance away from the waste material

  4. Dose rate reduction method for NMCA applied BWR plants

    International Nuclear Information System (INIS)

    Nagase, Makoto; Aizawa, Motohiro; Ito, Tsuyoshi; Hosokawa, Hideyuki; Varela, Juan; Caine, Thomas

    2012-09-01

    BRAC (BWR Radiation Assessment and Control) dose rate is used as an indicator of the incorporation of activated corrosion by products into BWR recirculation piping, which is known to be a significant contributor to dose rate received by workers during refueling outages. In order to reduce radiation exposure of the workers during the outage, it is desirable to keep BRAC dose rates as low as possible. After HWC was adopted to reduce IGSCC, a BRAC dose rate increase was observed in many plants. As a countermeasure to these rapid dose rate increases under HWC conditions, Zn injection was widely adopted in United States and Europe resulting in a reduction of BRAC dose rates. However, BRAC dose rates in several plants remain high, prompting the industry to continue to investigate methods to achieve further reductions. In recent years a large portion of the BWR fleet has adopted NMCA (NobleChem TM ) to enhance the hydrogen injection effect to suppress SCC. After NMCA, especially OLNC (On-Line NobleChem TM ), BRAC dose rates were observed to decrease. In some OLNC applied BWR plants this reduction was observed year after year to reach a new reduced equilibrium level. This dose rate reduction trends suggest the potential dose reduction might be obtained by the combination of Pt and Zn injection. So, laboratory experiments and in-plant tests were carried out to evaluate the effect of Pt and Zn on Co-60 deposition behaviour. Firstly, laboratory experiments were conducted to study the effect of noble metal deposition on Co deposition on stainless steel surfaces. Polished type 316 stainless steel coupons were prepared and some of them were OLNC treated in the test loop before the Co deposition test. Water chemistry conditions to simulate HWC were as follows: Dissolved oxygen, hydrogen and hydrogen peroxide were below 5 ppb, 100 ppb and 0 ppb (no addition), respectively. Zn was injected to target a concentration of 5 ppb. The test was conducted up to 1500 hours at 553 K. Test

  5. Performance of iron–chromium–aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Heuser, Brent J., E-mail: bheuser@illinois.edu [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Mandapaka, Kiran K.; Was, Gary S. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI 48109 (United States)

    2016-03-15

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe–Zr is addressed with the FeCrAl-YSZ system. - Graphical abstract: Weight gain normalized to total sample surface area versus time during 700 °C steam exposure for FeCrAl samples with different composition (A) and Fe/Cr/Al:62/4/34 (B). In both cases, the responses of uncoated Zry2 (Zry2-13A and Zry2-19A) are shown for comparison. This uncoated Zry2 response shows the expected pre-transition quasi-cubic kinetic behavior and eventual breakaway (linear) kinetics. Highlights: • FeCrAl coatings deposited on Zy2 have been tested with respect to oxidation in high-temperature steam. • FeCrAl compositions promoting alumina formation inhibited oxidation of Zy2 and delay weight gain. • Autoclave testing to 20 days of coated Zy2 in a simulated BWR environment demonstrates minimal weight gain and no film degradation. • The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

  6. CFD predictions of standby liquid control system mixing in lower plenum of a BWR

    International Nuclear Information System (INIS)

    Boyd, Christopher; Skarda, Raymond

    2014-01-01

    Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

  7. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  8. Computer simulation of population dynamics inside the urban environment

    Science.gov (United States)

    Andreev, A. S.; Inovenkov, I. N.; Echkina, E. Yu.; Nefedov, V. V.; Ponomarenko, L. S.; Tikhomirov, V. V.

    2017-12-01

    In this paper using a mathematical model of the so-called “space-dynamic” approach we investigate the problem of development and temporal dynamics of different urban population groups. For simplicity we consider an interaction of only two population groups inside a single urban area with axial symmetry. This problem can be described qualitatively by a system of two non-stationary nonlinear differential equations of the diffusion type with boundary conditions of the third type. The results of numerical simulations show that with a suitable choice of the diffusion coefficients and interaction functions between different population groups we can receive different scenarios of population dynamics: from complete displacement of one population group by another (originally more “aggressive”) to the “peaceful” situation of co-existence of them together.

  9. Computer simulations of polymers in a confined environment

    International Nuclear Information System (INIS)

    Sikorski, Andrzej; Romiszowski, Piotr

    2007-01-01

    A coarse-grained model of star-branched polymers confined in a slit formed by two parallel impenetrable surfaces, which were attractive for polymer segments, was developed and studied. The model chains were regular stars consisting of f = 3 branches of equal length. The flexible chains were constructed of united atoms (segments) and were restricted to vertices of a simple cubic lattice. Good solvent conditions were modelled and, thus, the macromolecules interacted only with the excluded volume. The properties of the model chains were determined by means of Monte Carlo simulations with a sampling algorithm based on the local changes of conformation of the chains. It appeared that the strongly adsorbed chains located in slits of appropriate width could swap between both confining surfaces. The influence of the chain length, width of the slit and the temperature on the frequency of such jumps was studied. The mechanism of the chain motion is also discussed

  10. Behavioral finance and games: simulations in the academic environment

    Directory of Open Access Journals (Sweden)

    Eliana Marcia Martins Fittipaldi Torga

    2017-12-01

    Full Text Available ABSTRACT The contribution from this study lies in its reflection on the factors that influence market efficiency, which requires a multidisciplinary view to analyze the intervening factors that impact results of the financial system. It also contributes by reflecting on the need for new approaches for training professionals who will go on to work in financial and related areas and preparing them by using different financial analysis techniques; by reflecting on the fact that analytical practices are influenced by social, cognitive, and emotional aspects, enabling the students to be better prepared to act in the financial market; by presenting various technical possibilities and providing more comprehensive knowledge to choose the one that best suits the object of analysis and their preferences; and by reflecting on different ways of perceiving investment opportunities and risk, which can be expanded on in other studies on the segmentation of clients according to their preferences in the investor market. The aim of this study was to analyze how social and psychological aspects influenced the decisions involved in simulated trading operations. The relevance lies in its discussion of the philosophical and epistemological position in finance, which suffers from a vision that only focuses on the rationality of means and does not explain the anomalies verified in the financial market. The study originated from the application of a company game simulating the work of stock market trading desk operators, applied in the Stock Market Operations course and using fundamental, technical, and graphical techniques. The population was intentional and made up of undergraduate and graduate students from one of the four best Brazilian federal universities. The data analysis was performed by analyzing the content of the questionnaires applied and the journal entries made during participant observation.

  11. Analysis of results of AZTRAN and AZKIND codes for a BWR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L.; Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M.

    2016-09-01

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  12. D-VASim: An Interactive Virtual Laboratory Environment for the Simulation and Analysis of Genetic Circuits

    DEFF Research Database (Denmark)

    Baig, Hasan; Madsen, Jan

    2016-01-01

    runtime. The runtime interaction gives the user a feeling of being in the lab performing a real world experiment. In this work, we present a user-friendly software tool named D-VASim (Dynamic Virtual Analyzer and Simulator), which provides a virtual laboratory environment to simulate and analyze...

  13. Using Blackboard Wiki Pages as a Shared Space for Simulating the Professional Translation Work Environment

    Science.gov (United States)

    Vine, Juliet

    2015-01-01

    The Work-Integrated Simulation for Translators module is part of a three year undergraduate degree in translation. The semester long module aims to simulate several aspects of the translation process using the Blackboard virtual learning environment's Wikis as the interface for completing translation tasks. For each translation task, one of the…

  14. Optimizing NEURON Simulation Environment Using Remote Memory Access with Recursive Doubling on Distributed Memory Systems

    OpenAIRE

    Shehzad, Danish; Bozkuş, Zeki

    2016-01-01

    Increase in complexity of neuronal network models escalated the efforts to make NEURON simulation environment efficient. The computational neuroscientists divided the equations into subnets amongst multiple processors for achieving better hardware performance. On parallel machines for neuronal networks, interprocessor spikes exchange consumes large section of overall simulation time. In NEURON for communication between processors Message Passing Interface (MPI) is used. MPI_Allgather collecti...

  15. Optimization of axial enrichment and gadolinia distributions for BWR fuel under control rod programming, (2)

    International Nuclear Information System (INIS)

    Hida, Kazuki; Yoshioka, Ritsuo

    1992-01-01

    A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance. This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs. (author)

  16. Prediction of droplet deposition around BWR fuel spacer by FEM flow analysis

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shinichi

    1997-01-01

    The critical power of the BWR fuel assembly has been remarkably increased. That increase mainly depends on the improvement of the spacer which keeps fixed gaps between fuel rods. So far, these improvements have been carried out on the basis of what developers consider to be appropriate and the results of mockup tests of the BWR fuel assembly. However, continued reliance on these approaches for the development of a higher performance fuel assembly will prove time-consuming and costly. Therefore, it is hoped that the spacer effects for the critical power can be investigated by computer simulation, and it is significantly important to develop the critical power prediction method. Direct calculation of the two-phase flow in a BWR fuel channel s still difficult. Accordingly, a new method for predicting the critical power was proposed. Our method consists of CFD (computer fluid dynamics) code based on the single-phase flow analysis method and the subchannel analysis code. To verify our method, the critical power predictions for various spacer geometries were performed. The predicted results of the critical power were compared with the experimental data. The result of the comparison showed a good agreement and the applicability of our method for various spacer geometries. (author)

  17. Intelligent manufacturing through participation : a participative simulation environment for integral manufacturing enterprise renewal

    NARCIS (Netherlands)

    Eijnatten, F.M. van

    2002-01-01

    This book deals with a 'Participative Simulation environment for Intelligent Manufacturing' (PSIM). PSIM is a software environment for use in assembly operations and it is developed and pilot-demonstrated in five companies: Volvo (Sweden), Finland Post, Fiat (Italy), Yamatake (Japan), Ford (USA).

  18. Application of computational fluid dynamics in building performance simulation for the outdoor environment: an overview

    NARCIS (Netherlands)

    Blocken, B.J.E.; Stathopoulos, T.; Carmeliet, J.; Hensen, J.L.M.

    2011-01-01

    This paper provides an overview of the application of CFD in building performance simulation for the outdoor environment, focused on four topics: (1) pedestrian wind environment around buildings, (2) wind-driven rain on building facades, (3) convective heat transfer coefficients at exterior building

  19. The Potential of Simulated Environments in Teacher Education: Current and Future Possibilities

    Science.gov (United States)

    Dieker, Lisa A.; Rodriguez, Jacqueline A.; Lignugaris/Kraft, Benjamin; Hynes, Michael C.; Hughes, Charles E.

    2014-01-01

    The future of virtual environments is evident in many fields but is just emerging in the field of teacher education. In this article, the authors provide a summary of the evolution of simulation in the field of teacher education and three factors that need to be considered as these environments further develop. The authors provide a specific…

  20. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  1. Methods and results of a PSA level 2 for a German BWR of the 900 MWe class

    International Nuclear Information System (INIS)

    Loffler, H.; Sonnenkalb, M.

    2006-01-01

    On behalf of the federal Ministry for Environment, Nature Conservation and Reactor Safety (BMU) GRS has performed a PSA level 2 for a BWR type 69 NPP of the 900 MWe class, equipped with a N 2 inerted steel containment and a pressure suppression system. Integral deterministic accident analyses have been performed with the computer code MELCOR 1.8.5. Additional analyses have been done for those events and phenomena which are not or not sufficiently covered by MELCOR. The probabilistic event tree analysis begins with the core damage states received from PSA level 1, and it ends with the definition of release categories and the determination of their frequencies. Uncertainties about the frequency of core damage states and about events during the accident progression are taken into account by means of Monte Carlo simulations. If there is a core damage state there is a high probability (>50 %) for a very high and rapid release of radionuclides into the environment. This high conditional probability is due to the very low probability to retain a partly destroyed core inside the reactor pressure vessel (RPV) and because the containment almost certainly fails at the bottom of the control rod drives room after melt release from the failed RPV. (authors)

  2. Learner-Centered Instruction (LCI): Volume IV, The Simulated Maintenance Task Environment (SMTE): A Job Specific Simulator.

    Science.gov (United States)

    Rifkin, Kenneth I.; And Others

    The purpose of the simulated maintenance task environment is to provide a means for training and job performance testing of the flight line weapon control systems mechanic/technician for the F-111A aircraft. It provides practice in flight line equipment checkout, troubleshooting, and removal and replacement of line replaceable units in the…

  3. Two-loop feed water control system in BWR plants

    International Nuclear Information System (INIS)

    Omori, Takashi; Watanabe, Takao; Hirose, Masao.

    1982-01-01

    In the process of the start-up and shutdown of BWR plants, the operation of changing over feed pumps corresponding to plant output is performed. Therefore, it is necessary to develop the automatic changeover system for feed pumps, which minimizes the variation of water level in reactors and is easy to operate. The three-element control system with the water level in reactors, the flow rate of main steam and the flow rate of feed water as the input is mainly applied, but long time is required for the changeover of feed pumps. The two-loop feed control system can control simultaneously two pumps being changed over, therefore it is suitable to the automatic changeover control system for feed pumps. Also it is excellent for the control of the recirculating valves of feed pumps. The control characteristics of the two-loop feed water control system against the external disturbance which causes the variation of water level in reactors were examined. The results of analysis by simulation are reported. The features of the two-loop feed water control system, the method of simulation and the evaluation of the two-loop feed water control system are described. Its connection with a digital feed water recirculation control system is expected. (Kako, I.)

  4. Real-Time and High-Fidelity Simulation Environment for Autonomous Ground Vehicle Dynamics

    Science.gov (United States)

    Cameron, Jonathan; Myint, Steven; Kuo, Calvin; Jain, Abhi; Grip, Havard; Jayakumar, Paramsothy; Overholt, Jim

    2013-01-01

    This paper reports on a collaborative project between U.S. Army TARDEC and Jet Propulsion Laboratory (JPL) to develop a unmanned ground vehicle (UGV) simulation model using the ROAMS vehicle modeling framework. Besides modeling the physical suspension of the vehicle, the sensing and navigation of the HMMWV vehicle are simulated. Using models of urban and off-road environments, the HMMWV simulation was tested in several ways, including navigation in an urban environment with obstacle avoidance and the performance of a lane change maneuver.

  5. A comparison of the accuracy of intraoral scanners using an intraoral environment simulator.

    Science.gov (United States)

    Park, Hye-Nan; Lim, Young-Jun; Yi, Won-Jin; Han, Jung-Suk; Lee, Seung-Pyo

    2018-02-01

    The aim of this study was to design an intraoral environment simulator and to assess the accuracy of two intraoral scanners using the simulator. A box-shaped intraoral environment simulator was designed to simulate two specific intraoral environments. The cast was scanned 10 times by Identica Blue (MEDIT, Seoul, South Korea), TRIOS (3Shape, Copenhagen, Denmark), and CS3500 (Carestream Dental, Georgia, USA) scanners in the two simulated groups. The distances between the left and right canines (D3), first molars (D6), second molars (D7), and the left canine and left second molar (D37) were measured. The distance data were analyzed by the Kruskal-Wallis test. The differences in intraoral environments were not statistically significant ( P >.05). Between intraoral scanners, statistically significant differences ( P Kruskal-Wallis test with regard to D3 and D6. No difference due to the intraoral environment was revealed. The simulator will contribute to the higher accuracy of intraoral scanners in the future.

  6. GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

    International Nuclear Information System (INIS)

    Hansen, P.N.

    2004-01-01

    This article introduces a GOTHIC version 7.1 model of the Secondary Containment Reactor Building Post LOCA drawdown analysis for a BWR. GOTHIC is an EPRI sponsored thermal hydraulic code. This analysis is required by the Utility to demonstrate an ability to restore and maintain the Secondary Containment Reactor Building negative pressure condition. The technical and regulatory issues associated with this modeling are presented. The analysis includes the affect of wind, elevation and thermal impacts on pressure conditions. The model includes a multiple volume representation which includes the spent fuel pool. In addition, heat sources and sinks are modeled as one dimensional heat conductors. The leakage into the building is modeled to include both laminar as well as turbulent behavior as established by actual plant test data. The GOTHIC code provides components to model heat exchangers used to provide fuel pool cooling as well as area cooling via air coolers. The results of the evaluation are used to demonstrate the time that the Reactor Building is at a pressure that exceeds external conditions. This time period is established with the GOTHIC model based on the worst case pressure conditions on the building. For this time period the Utility must assume the primary containment leakage goes directly to the environment. Once the building pressure is restored below outside conditions the release to the environment can be credited as a filtered release

  7. Repeated Induction of Inattentional Blindness in a Simulated Aviation Environment

    Science.gov (United States)

    Kennedy, Kellie D.; Stephens, Chad L.; Williams, Ralph A.; Schutte, Paul C.

    2017-01-01

    The study reported herein is a subset of a larger investigation on the role of automation in the context of the flight deck and used a fixed-based, human-in-the-loop simulator. This paper explored the relationship between automation and inattentional blindness (IB) occurrences in a repeated induction paradigm using two types of runway incursions. The critical stimuli for both runway incursions were directly relevant to primary task performance. Sixty non-pilot participants performed the final five minutes of a landing scenario twice in one of three automation conditions: full automation (FA), partial automation (PA), and no automation (NA). The first induction resulted in a 70 percent (42 of 60) detection failure rate with those in the PA condition significantly more likely to detect the incursion compared to the FA condition or the NA condition. The second induction yielded a 50 percent detection failure rate. Although detection improved (detection failure rates declined) in all conditions, those in the FA condition demonstrated the greatest improvement with doubled detection rates. The detection behavior in the first trial did not preclude a failed detection in the second induction. Group membership (IB vs. Detection) in the FA condition showed a greater improvement than those in the NA condition and rated the Mental Demand and Effort subscales of the NASA-TLX (NASA Task Load Index) significantly higher for Time 2 compared Time 1. Participants in the FA condition used the experience of IB exposure to improve task performance whereas those in the NA condition did not, indicating the availability and reallocation of attentional resources in the FA condition. These findings support the role of engagement in operational attention detriment and the consideration of attentional failure causation to determine appropriate mitigation strategies.

  8. Reliability Verification of DBE Environment Simulation Test Facility by using Statistics Method

    International Nuclear Information System (INIS)

    Jang, Kyung Nam; Kim, Jong Soeg; Jeong, Sun Chul; Kyung Heum

    2011-01-01

    In the nuclear power plant, all the safety-related equipment including cables under the harsh environment should perform the equipment qualification (EQ) according to the IEEE std 323. There are three types of qualification methods including type testing, operating experience and analysis. In order to environmentally qualify the safety-related equipment using type testing method, not analysis or operation experience method, the representative sample of equipment, including interfaces, should be subjected to a series of tests. Among these tests, Design Basis Events (DBE) environment simulating test is the most important test. DBE simulation test is performed in DBE simulation test chamber according to the postulated DBE conditions including specified high-energy line break (HELB), loss of coolant accident (LOCA), main steam line break (MSLB) and etc, after thermal and radiation aging. Because most DBE conditions have 100% humidity condition, in order to trace temperature and pressure of DBE condition, high temperature steam should be used. During DBE simulation test, if high temperature steam under high pressure inject to the DBE test chamber, the temperature and pressure in test chamber rapidly increase over the target temperature. Therefore, the temperature and pressure in test chamber continue fluctuating during the DBE simulation test to meet target temperature and pressure. We should ensure fairness and accuracy of test result by confirming the performance of DBE environment simulation test facility. In this paper, in order to verify reliability of DBE environment simulation test facility, statistics method is used

  9. Virtual X-ray imaging techniques in an immersive casting simulation environment

    International Nuclear Information System (INIS)

    Li, Ning; Kim, Sung-Hee; Suh, Ji-Hyun; Cho, Sang-Hyun; Choi, Jung-Gil; Kim, Myoung-Hee

    2007-01-01

    A computer code was developed to simulate radiograph of complex casting products in a CAVE TM -like environment. The simulation is based on the deterministic algorithms and ray tracing techniques. The aim of this study is to examine CAD/CAE/CAM models at the design stage, to optimize the design and inspect predicted defective regions with fast speed, good accuracy and small numerical expense. The present work discusses the algorithms for the radiography simulation of CAD/CAM model and proposes algorithmic solutions adapted from ray-box intersection algorithm and octree data structure specifically for radiographic simulation of CAE model. The stereoscopic visualization of full-size of product in the immersive casting simulation environment as well as the virtual X-ray images of castings provides an effective tool for design and evaluation of foundry processes by engineers and metallurgists

  10. Simulation of sustainability aspects within the industrial environment and their implication on the simulation technique

    OpenAIRE

    Rabe, M.; Jäkel, F.-W.; Weinaug, H.

    2010-01-01

    Simulation is a broadly excepted analytic instrument and planning tool. Today, industrial simulation is mainly applied for engineering and physical purposes and covers a short time horizon compared to intergenerational justice. In parallel, sustainability is gaining more importance for the industrial planning because themes like global warming, child labour, and compliance with social and environmental standards have to be taken into account. Sustainability is characterized by comprehensively...

  11. MaGate Simulator: A Simulation Environment for a Decentralized Grid Scheduler

    Science.gov (United States)

    Huang, Ye; Brocco, Amos; Courant, Michele; Hirsbrunner, Beat; Kuonen, Pierre

    This paper presents a simulator for of a decentralized modular grid scheduler named MaGate. MaGate’s design emphasizes scheduler interoperability by providing intelligent scheduling serving the grid community as a whole. Each MaGate scheduler instance is able to deal with dynamic scheduling conditions, with continuously arriving grid jobs. Received jobs are either allocated on local resources, or delegated to other MaGates for remote execution. The proposed MaGate simulator is based on GridSim toolkit and Alea simulator, and abstracts the features and behaviors of complex fundamental grid elements, such as grid jobs, grid resources, and grid users. Simulation of scheduling tasks is supported by a grid network overlay simulator executing distributed ant-based swarm intelligence algorithms to provide services such as group communication and resource discovery. For evaluation, a comparison of behaviors of different collaborative policies among a community of MaGates is provided. Results support the use of the proposed approach as a functional ready grid scheduler simulator.

  12. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

  13. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  14. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  15. BWR plant advanced central control panel PODIA

    International Nuclear Information System (INIS)

    Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

    1983-01-01

    BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

  16. Evaluation of internal flooding in a BWR

    International Nuclear Information System (INIS)

    Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

    1985-01-01

    Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event

  17. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  18. A BWR Safety and Operability Improvements

    International Nuclear Information System (INIS)

    Sawyer, Craig D.

    1993-01-01

    The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

  19. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  20. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J.M.; Blazquez, J.B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical struct-ure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

  1. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Kato, Shigeru

    1996-01-01

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  2. Facility of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsuji

    1998-01-01

    A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

  3. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  4. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  5. BWR startup and shutdown activity transport control

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

    2010-07-01

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

  6. Generation of large scale urban environments to support advanced sensor and seeker simulation

    Science.gov (United States)

    Giuliani, Joseph; Hershey, Daniel; McKeown, David, Jr.; Willis, Carla; Van, Tan

    2009-05-01

    One of the key aspects for the design of a next generation weapon system is the need to operate in cluttered and complex urban environments. Simulation systems rely on accurate representation of these environments and require automated software tools to construct the underlying 3D geometry and associated spectral and material properties that are then formatted for various objective seeker simulation systems. Under an Air Force Small Business Innovative Research (SBIR) contract, we have developed an automated process to generate 3D urban environments with user defined properties. These environments can be composed from a wide variety of source materials, including vector source data, pre-existing 3D models, and digital elevation models, and rapidly organized into a geo-specific visual simulation database. This intermediate representation can be easily inspected in the visible spectrum for content and organization and interactively queried for accuracy. Once the database contains the required contents, it can then be exported into specific synthetic scene generation runtime formats, preserving the relationship between geometry and material properties. To date an exporter for the Irma simulation system developed and maintained by AFRL/Eglin has been created and a second exporter to Real Time Composite Hardbody and Missile Plume (CHAMP) simulation system for real-time use is currently being developed. This process supports significantly more complex target environments than previous approaches to database generation. In this paper we describe the capabilities for content creation for advanced seeker processing algorithms simulation and sensor stimulation, including the overall database compilation process and sample databases produced and exported for the Irma runtime system. We also discuss the addition of object dynamics and viewer dynamics within the visual simulation into the Irma runtime environment.

  7. Analysis of the effects of simulated synergistic LEO environment on solar panels

    Science.gov (United States)

    Allegri, G.; Corradi, S.; Marchetti, M.; Scaglione, S.

    2007-02-01

    The effects due to the LEO environment exposure of a solar array primary structure are here presented and discussed in detail. The synergistic damaging components featuring LEO environment are high vacuum, thermal cycling, neutral gas, ultraviolet (UV) radiation and cold plasma. The synergistic effects due to these environmental elements are simulated by "on ground" tests, performed in the Space Environment Simulator (SAS) at the University of Rome "La Sapienza"; numerical simulations are performed by the Space Environment Information System (SPENVIS), developed by the European Space Agency (ESA). A "safe life" design for a solar array primary structure is developed, taking into consideration the combined damaging action of the LEO environment components; therefore results from both numerical and experimental simulations are coupled within the framework of a standard finite element method (FEM) based design. The expected durability of the solar array primary structure, made of laminated sandwich composite, is evaluated assuming that the loads exerted on the structure itself are essentially dependent on thermo-elastic stresses. The optical degradation of surface materials and the stiffness and strength degradation of structural elements are taken into account to assess the global structural durability of the solar array under characteristic operative conditions in LEO environment.

  8. Interactive Learning Environment: Web-based Virtual Hydrological Simulation System using Augmented and Immersive Reality

    Science.gov (United States)

    Demir, I.

    2014-12-01

    Recent developments in internet technologies make it possible to manage and visualize large data on the web. Novel visualization techniques and interactive user interfaces allow users to create realistic environments, and interact with data to gain insight from simulations and environmental observations. The hydrological simulation system is a web-based 3D interactive learning environment for teaching hydrological processes and concepts. The simulation systems provides a visually striking platform with realistic terrain information, and water simulation. Students can create or load predefined scenarios, control environmental parameters, and evaluate environmental mitigation alternatives. The web-based simulation system provides an environment for students to learn about the hydrological processes (e.g. flooding and flood damage), and effects of development and human activity in the floodplain. The system utilizes latest web technologies and graphics processing unit (GPU) for water simulation and object collisions on the terrain. Users can access the system in three visualization modes including virtual reality, augmented reality, and immersive reality using heads-up display. The system provides various scenarios customized to fit the age and education level of various users. This presentation provides an overview of the web-based flood simulation system, and demonstrates the capabilities of the system for various visualization and interaction modes.

  9. BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    Anstine, L.D.

    1983-05-01

    This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

  10. Physics-based statistical model and simulation method of RF propagation in urban environments

    Science.gov (United States)

    Pao, Hsueh-Yuan; Dvorak, Steven L.

    2010-09-14

    A physics-based statistical model and simulation/modeling method and system of electromagnetic wave propagation (wireless communication) in urban environments. In particular, the model is a computationally efficient close-formed parametric model of RF propagation in an urban environment which is extracted from a physics-based statistical wireless channel simulation method and system. The simulation divides the complex urban environment into a network of interconnected urban canyon waveguides which can be analyzed individually; calculates spectral coefficients of modal fields in the waveguides excited by the propagation using a database of statistical impedance boundary conditions which incorporates the complexity of building walls in the propagation model; determines statistical parameters of the calculated modal fields; and determines a parametric propagation model based on the statistical parameters of the calculated modal fields from which predictions of communications capability may be made.

  11. Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

    International Nuclear Information System (INIS)

    Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

    2004-01-01

    The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

  12. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  13. Development of power change maneuvering method for BWR

    International Nuclear Information System (INIS)

    Fukuzaki, Takaharu; Yamada, Naoyuki; Kiguchi, Takashi; Sakurai, Mikio.

    1985-01-01

    A power change maneuvering method for BWR has been proposed to generate an optimal power control maneuver, which realizes the power change operation closest to a power change demand pattern under operating constraints. The method searches for the maneuver as an optimization problem, where the variables are thermal power levels sampled from the demand pattern, the performance index is defined to express the power mismatch between demand and feasible patterns, and the constraints are limit lines on the thermal power-core flow rate map and limits on keeping fuel integrity. The usable feasible direction method is utilized as the optimization algorithm, with newly developed techniques for initial value generation and step length determination, which apply one-dimensional search and inverse-interpolation methods, respectively, to realize the effective search of the optimal solution. Simulation results show that a typical computing time is about 5 min by a general purpose computer and the method has been verified to be practical even for on-line use. (author)

  14. BWR regional instability model and verification on ringhals-1 test

    International Nuclear Information System (INIS)

    Hotta, Akitoshi; Suzawa, Yojiro

    1996-01-01

    Regional instability is known as one type of the coupled neutronic-thermohydraulic phenomena of boiling water reactors (BWRs), where the thermohydraulic density wave propagation mechanism is predominant. Historically, it has been simulated by the three-dimensional time domain code in spite of its significant computing time. On the other hand, there have been proposals to apply the frequency domain models in regional instability considering the subcriticality of the higher neutronic mode. However, their application still remains in corewide instability mainly because of the lack of more detailed methodological and empirical studies. In this study, the current version of the frequency domain model was extended and verified based on actual core regional instability measurement data. The mathematical model LAPUR, the well-known frequency domain stability code, was reviewed from the standpoint of pure thermohydraulics and neutronic-thermohydraulic interaction mechanisms. Based on the ex-core loop test data, the original LAPUR mixed friction and local pressure loss model was modified, taking into account the different dynamic behavior of these two pressure-loss mechanisms. The perturbation term of the two-phase friction multiplier, which is the sum of the derivative of void fraction and subcool enthalpy, was adjusted theoretically. The adequacy of the instability evaluation system was verified based on the Ringhals unit 1 test data, which were supplied to participants of the Organization for Economic Cooperation and Development/Nuclear Energy Agency BWR Stability Benchmark Project

  15. Using numeric simulation in an online e-learning environment to teach functional physiological contexts.

    Science.gov (United States)

    Christ, Andreas; Thews, Oliver

    2016-04-01

    Mathematical models are suitable to simulate complex biological processes by a set of non-linear differential equations. These simulation models can be used as an e-learning tool in medical education. However, in many cases these mathematical systems have to be treated numerically which is computationally intensive. The aim of the study was to develop a system for numerical simulation to be used in an online e-learning environment. In the software system the simulation is located on the server as a CGI application. The user (student) selects the boundary conditions for the simulation (e.g., properties of a simulated patient) on the browser. With these parameters the simulation on the server is started and the simulation result is re-transferred to the browser. With this system two examples of e-learning units were realized. The first one uses a multi-compartment model of the glucose-insulin control loop for the simulation of the plasma glucose level after a simulated meal or during diabetes (including treatment by subcutaneous insulin application). The second one simulates the ion transport leading to the resting and action potential in nerves. The student can vary parameters systematically to explore the biological behavior of the system. The described system is able to simulate complex biological processes and offers the possibility to use these models in an online e-learning environment. As far as the underlying principles can be described mathematically, this type of system can be applied to a broad spectrum of biomedical or natural scientific topics. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

  16. Interpretation of the results of the CORA-33 dry core BWR test

    International Nuclear Information System (INIS)

    Ott, L.J.; Hagen, S.

    1993-01-01

    All BWR degraded core experiments performed prior to CORA-33 were conducted under ''wet'' core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ''dry'' core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ''dry'' core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ''dry'' BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ''dry'' core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed

  17. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  18. Simulation-based computation of dose to humans in radiological environments

    Energy Technology Data Exchange (ETDEWEB)

    Breazeal, N.L. [Sandia National Labs., Livermore, CA (United States); Davis, K.R.; Watson, R.A. [Sandia National Labs., Albuquerque, NM (United States); Vickers, D.S. [Brigham Young Univ., Provo, UT (United States). Dept. of Electrical and Computer Engineering; Ford, M.S. [Battelle Pantex, Amarillo, TX (United States). Dept. of Radiation Safety

    1996-03-01

    The Radiological Environment Modeling System (REMS) quantifies dose to humans working in radiological environments using the IGRIP (Interactive Graphical Robot Instruction Program) and Deneb/ERGO simulation software. These commercially available products are augmented with custom C code to provide radiation exposure information to, and collect radiation dose information from, workcell simulations. Through the use of any radiation transport code or measured data, a radiation exposure input database may be formulated. User-specified IGRIP simulations utilize these databases to compute and accumulate dose to programmable human models operating around radiation sources. Timing, distances, shielding, and human activity may be modeled accurately in the simulations. The accumulated dose is recorded in output files, and the user is able to process and view this output. The entire REMS capability can be operated from a single graphical user interface.

  19. Simulation-based computation of dose to humans in radiological environments

    International Nuclear Information System (INIS)

    Breazeal, N.L.; Davis, K.R.; Watson, R.A.; Vickers, D.S.; Ford, M.S.

    1996-03-01

    The Radiological Environment Modeling System (REMS) quantifies dose to humans working in radiological environments using the IGRIP (Interactive Graphical Robot Instruction Program) and Deneb/ERGO simulation software. These commercially available products are augmented with custom C code to provide radiation exposure information to, and collect radiation dose information from, workcell simulations. Through the use of any radiation transport code or measured data, a radiation exposure input database may be formulated. User-specified IGRIP simulations utilize these databases to compute and accumulate dose to programmable human models operating around radiation sources. Timing, distances, shielding, and human activity may be modeled accurately in the simulations. The accumulated dose is recorded in output files, and the user is able to process and view this output. The entire REMS capability can be operated from a single graphical user interface

  20. Multiple wavelength spectral system simulating background light noise environment in satellite laser communications

    Science.gov (United States)

    Lu, Wei; Sun, Jianfeng; Hou, Peipei; Xu, Qian; Xi, Yueli; Zhou, Yu; Zhu, Funan; Liu, Liren

    2017-08-01

    Performance of satellite laser communications between GEO and LEO satellites can be influenced by background light noise appeared in the field of view due to sunlight or planets and some comets. Such influences should be studied on the ground testing platform before the space application. In this paper, we introduce a simulator that can simulate the real case of background light noise in space environment during the data talking via laser beam between two lonely satellites. This simulator can not only simulate the effect of multi-wavelength spectrum, but also the effects of adjustable angles of field-of-view, large range of adjustable optical power and adjustable deflection speeds of light noise in space environment. We integrate these functions into a device with small and compact size for easily mobile use. Software control function is also achieved via personal computer to adjust these functions arbitrarily. Keywords: