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Sample records for boiling reactor experiment 2

  1. Boiling water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development, and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, Reactor Simulator Development: Workshop Material (2001). Course material for workshops using a WWER-1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21, 2nd edition, WWER-1000 Reactor Simulator: Workshop Material (2005). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator: Workshop Material (2005). This report consists of course material for workshops using a boiling water reactor (BWR) simulator

  2. High Pressure Boiling Water Reactor

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  3. Investigations on coolant boiling in research reactors. 2

    International Nuclear Information System (INIS)

    Subcooled boiling has been investigated systematically at the Rossendorf Research Reactor in the range between boiling onset and boiling crisis. This is of particular interest because in the core the direction of the coolant flow is opposite to the bubble buoyance of the bubbles - in contrast to power reactors. For this reason an experimental fuel assembly equipped with a throttle valve for coolant flow reduction and different detectors was built up and installed in the reactor core. Measurements of thermohydraulic parameters and noise signals from temperature, neutron flux and acoustic sources were subject of the investigations. Besides other results fluctuations of the void fraction induced by a standing wave of the two-phase flow in the coolant channel and the 24-Hz pressure fluctuations of the circulation pumps have been observed. It has been shown that the frequency of the standing wave is determined by the size of the boiling volume in the coolant channel and that this frequency therefore depends on the outlet temperature of the coolant. (author)

  4. Subcooled flow boiling experiments and numerical simulation for a virtual reactor development

    International Nuclear Information System (INIS)

    Subcooled flow boiling experiments and numerical simulations using a Lattice Boltzmann model will be performed at City College of New York as part of the DOE Nuclear HUB project, Consortium for Advanced Simulation of Light Water Reactors (CASL). The experiments being performed include pool boiling from a platinum wire, subcooled flow boiling in a vertical tube and single air bubble injection into a turbulent water stream. Preliminary experiments have been performed to measure the bubble size, shape and motion in an adiabatic experiment involving air bubble injection into water flowing in a vertical annulus, as well as PIV measurements of liquid flow field in a subcooled flow boiling experiment. An advanced thermal Finite Element Lattice Boltzmann Model is being developed to predict the pool and flow boiling experiments. After the validation of the code, improved constitutive relations for subcooled flow boiling will be developed for use in 3-D CFD models. The present work is expected to contribute to the development of a multi-scale, multi-physics model of a PWR in the CASL project. (author)

  5. Industrial application of APOLLO2 to boiling water reactors

    International Nuclear Information System (INIS)

    AREVA NP - a joint's subsidiary of AREVA and Siemens- decided to develop a new calculation scheme based on the multigroup neutron transport code APOLLO2, developed at CEA, for industrial application to Boiling Water Reactors. This scheme is based on the CEA93 library with the XMAS-172 energy mesh and the JEF2.2 evaluation. Microscopic cross-sections are improved by a self-shielding calculation that accounts for 2D geometrical effects and the overlapping of resonances. The flux is calculated with the Method of Characteristics. A best-estimate flux is found with the 172 energy group structure. In the industrial scheme, the computing time and the memory size are reduced by a simplified self-shielding and the calculation of the flux with 26 energy groups. The results are presented for three BWR assemblies. Several BWR operating conditions were simulated. Results are accurate compared to the Monte-Carlo code MCNP. A very good agreement is obtained between the best-estimate and the industrial calculations, also during depletion. These results show the high physical quality of the APOLLO2 code and its capability to calculate accurately BWR assemblies for industrial applications. (authors)

  6. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  7. Validation of HELIOS for calculations of experiments in the Halden Boiling Water Reactor

    International Nuclear Information System (INIS)

    The two dimensional transport code HELIOS has been used since the beginning of 1998 for the neutron physics calculations of experiments in the Halden Boiling Water Reactor (HBWR). Because a lot of measured data from these experiments are depending also on calculated values, it was necessary to validate HELIOS for these calculations. Therefore several experiments were re-calculated and it is shown that there is a good agreement between calculated and measured data. In some cases the effect of the calculated values on the measurements are shown, and this also shows that HELIOS gives reliable results

  8. Boiling water reactor shutdown dose rate experience after on-line NobleChem™

    International Nuclear Information System (INIS)

    All U.S. boiling water reactors (BWRs) inject hydrogen for mitigation of intergranular stress corrosion cracking (IGSCC), depleted zinc oxide (DZO) for control of shutdown dose rates, and most have implemented or plan to implement On-Line NobleChem™ (OLNC). In this process, the injection of a platinum compound that catalyzes the recombination of hydrogen and oxygen at surfaces results in restructuring of oxide films on reactor internals and piping, impacting reactor water Co-60 and shutdown dose rates. Since the first implementation of OLNC in 2005, the experience base has significantly expanded in both U.S. and non-U.S. BWRs. This paper investigates the response of reactor recirculation system (RRS) dose rates after OLNC and their relationship to reactor water chemistry parameters, including Co-60 and zinc, using data from EPRI's BWR Chemistry Monitoring and Assessment database. Results of a recent study evaluating correlations of chemistry parameters, other than Co-60, with RRS dose rates are discussed. Relevant revised guidance in the BWR Water Chemistry Guidelines is also presented. (author)

  9. Local stability tests in Dresden 2 boiling water reactor

    International Nuclear Information System (INIS)

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations

  10. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

  11. Instrumenting a pressure suppression experiment for a MK I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    A scale test facility of a pressure suppression system from a boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy experimental data during a hypothetical loss-of-coolant accident. The instrumentation verified the analysis of the dynamic loading of the pressure suppression system

  12. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  13. Boiling liquid expanding vapour explosion in CO2 small scale experiments

    OpenAIRE

    Bjerketvedt, Dag; Egeberg, Kjersti; Ke, Wei; Gaathaug, Andre Vagner; Vågsæther, Knut; Nilsen, S. H.

    2011-01-01

    Carbon capture and storage systems require handling large volumes of high pressure CO2. Having thorough knowledge of the related hazards is essential, as is knowing how to prevent, detect, control and mitigate accidents. This paper gives a short description of CO2 Boiling Liquid Expanding Vapour Explosions (BLEVEs) and presents results from preliminary, small scale experiments with CO2 BLEVEs. The mechanism of superheated liquid CO2 boiling is not fully understood. Analogies can be made betwe...

  14. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

  15. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  16. U.S. experience with hydrogen water chemistry in boiling water reactors

    International Nuclear Information System (INIS)

    Hydrogen water chemistry in boiling water reactors is currently being adopted by many utilities in the U.S., with eleven units having completed preimplementation test programs, four units operating permanently with hydrogen water chemistry, and six other units in the process of installing permanent equipment. Intergranular stress corrosion cracking protection is required for the recirculation piping system and other regions of the BWR systems. The present paper explores progress in predicting and monitoring hydrogen water chemistry response in these areas. Testing has shown that impurities can play an important role in hydrogen water chemistry. Evaluation of their effects are also performed. Both computer modeling and in plant measurements show that each plant will respond uniquely to feedwater hydrogen addition. Thus, each plant has its own unique hydrogen requirement for recirculation system protecion. Furthermore, the modeling, and plant measurements show that different regions of the BWR respond differently to hydrogen injection. Thus, to insure protection of components other than the recirculation systems may require more (or less) hydrogen demand than indicated by the recirculation system measurements. In addition, impurities such as copper can play a significant role in establishing hydrogen demand. (Nogami, K.)

  17. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  18. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Classification of decommissioning wastes. Addendum 2

    International Nuclear Information System (INIS)

    The radioactive wastes expected to result from decommissioning of the reference boiling water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 18,949 cubic meters of waste from DECON are classified as follows: Class A, 97.5%; Class B, 2.0%; Class C, 0.3%. About 0.2% (47 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods

  19. Fuel performance in the Barsebeck boiling water reactors (Unit 1 and 2)

    International Nuclear Information System (INIS)

    Sydkraft is the largest privately owned utility in Sweden. It serves about 20% of the Swedish population with about 12 TWh of electric power per year, of which 64% is nuclear (1978 figures). The two identical 590 MWE ASEA-ATOM boiling water reactors in Barsebeck have been in operation since 1975 and 1977 respectively. Fission product activity in the primary circuits and in the off-gas systems is extremely low and indicate a near perfect fuel condition. Operating restrictions limiting the effect of pellet cladding interaction have been in use since initial start-up and testing. A few events involving rapid power increases above the preconditioned power level have occurred without causing fuel failures. It is believed that an analysis of power reactor operational transients, which did not cause fuel failures, can be useful to design more adequate and less conservative rules for the operation of nuclear reactor cores

  20. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  1. Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    This paper presents results of ex-vessel boiling experiments performed in the CYBL (cylindrical boiling) facility, a reactor-scale facility which has a tank-within-a-tank design simulating the reactor vessel and the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm2 across the vessel bottom are performed. Boiling outside the reactor vessel is found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that the flooded cavity in a passive PWR like the AP-600 should be capable of cooling the RPV in the central region of the lower head. 9 figs., 1 tab., 14 refs

  2. Analysis of mixed oxide fuel critical experiments with neutronics analysis codes for boiling water reactors

    International Nuclear Information System (INIS)

    Critical experiments of UO2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)

  3. Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

  4. Utilization of the SMART v2.1 monitor to calculate the stability of a boiling water reactor

    International Nuclear Information System (INIS)

    The nuclear reactor stability is very important in the shutdown and start-up of the boiling water reactor, because in these situations, working conditions are close to the unstable zone. For this reason, the Thermohydraulic and Nuclear Engineering Group, together with IBERDROLA, spend several years carrying out a monitor to analyze the stability of these reactors.

  5. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  6. Flow boiling experiment study of OTSG in steady state

    International Nuclear Information System (INIS)

    Generally, Integrated Pressurized Water Reactor uses OTSG. In this paper OTSG's flow boiling laws is analyzed and experimented. More elaborate calculating model of two-phase flow section which is divided into different segments is used for theoretical analysis. Flow boiling experiment is operated in two different conditions. The results of experiment proves the calculation model rational and credible. (authors)

  7. European simplified boiling water reactor (ESBWR) plant

    International Nuclear Information System (INIS)

    This paper covers innovative ideas which made possible the redesign of the US 660-MW Simplified Boiling Water Reactor (SBWR) Reactor Island for a 1,200-MW size reactor while actually reducing the building cost. This was achieved by breaking down the Reactor Island into multiple buildings separating seismic-1 from non-seismic-1 areas, providing for better space utilization, shorter construction schedule, easier maintainability and better postaccident accessibility

  8. PSEPLOT: a controller for plotting data from the Mark I Boiling Water Reactor Pressure Suppression Experiment

    International Nuclear Information System (INIS)

    PSEPLOT is a computer routine that was developed for the Lawrence Livermore Laboratory Octopus computer system to generate several thousand plots of engineering data in a consistent format for referencing and comparison. The time-dependent engineering data were recorded during each of 25 tests of the Mark I Pressure Suppression Experiment (PSE). Although PSEPLOT is restricted to PSE, its concept is applicable to any similar data management task

  9. Effects of Nanofluid for In-Vessel Retention External Reactor Vessel Cooling on Critical Heat Flux using Pool Boiling Experiments

    International Nuclear Information System (INIS)

    In-vessel retention (IVR) is one of the severe accident management (SAM) strategies that are used in some nuclear power plants: AP600, AP1000, Loviisa and APR1400. One way of IVR is the method of external reactor vessel cooling (ERVC). When core melts and deposits on the bottom of reactor vessel, ERVC is starting to flood the reactor cavity to remove the decay heat through the wall of the reactor vessel. This process can improve the plant economics by reducing regulatory requirements. And increased safety margin leads to gain public acceptance. In this system, the heat removal is restricted by thermal limit called by critical heat flux (CHF). Besides, as advanced light water reactors such as South Korea's APR-1400, thermal safety margin is deceased. So, it is essential to get more safety margin. There are some approaches to enhance the ERVC: using the coating on the vessel outer surface, increasing the reactor cavity flood level, streamlining the gap between the vessel and the vessel insulation. Many investigations have been performed to evaluate the coolability of IVR In this paper, we firstly investigated the coating effects in the critical heat flux among the above mentioned approach methods. During the boiling phenomenon, a thin layer was formed on the heater surface in the nanofluid. This coating mechanism is well known theoretically. Nanofluids are colloidal dispersions of nanoparticles in traditional heat transfer fluids. One of the most interesting characteristics of nanofluids is their capability to enhance the critical heat flux (CHF) significantly. Nanofluid is made by typical particle materials. Materials of nanoparticles include metals (e.g., silver, copper, gold), metal oxides (e.g., titania, alumina, silica, zirconia), carbon allotrope (e.g., carbon nanotube, graphite). We selected the grapheneoxide nanofluid which is a kind of carbon allotrope. Graphene-oxide is attractive material with the high thermal conductivity and stable dispersion ability in

  10. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    The fuel assembly of a boiling water reactor contains a number of vertical fuel rods with their lower ends against a bottom tie plate. The rods are positioned by spacers, which are fixed to the canning. The upward motion is reduced by the top plate of a special design. (G.B.)

  11. Self-Sustaining Thorium Boiling Water Reactors

    OpenAIRE

    Ehud Greenspan; Jasmina Vujic; Francesco Ganda; Arias, Francisco J.

    2012-01-01

    A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR) proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorber...

  12. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  13. Experimental investigation into the effects of coolant additives on boiling phenomena in pressurized water reactors

    International Nuclear Information System (INIS)

    This study investigates the effects of coolant additives like boric acid on boiling phenomena in pressurized water reactors under conditions as realistic as possible. The effects covered range from subcooled boiling to critical boiling conditions (CHF). The focus of this project lies on flow boiling with up to 40 bar and 250 °C in order to generate a data basis for a possible extrapolation to reactor conditions. The results of the experiments are used to implement and validate new models into CFD-Codes in context to a nationwide German joint research project with the specific aim of improving CFD boiling-models. (author)

  14. Serious accidents on boiling water reactors (BWR)

    International Nuclear Information System (INIS)

    This short document describes, first, the specificities of boiling water reactors (BWRs) with respect to PWRs in front of the progress of a serious accident, and then, the strategies of accident management: restoration of core cooling, water injection, core flooding, management of hydrogen release, depressurization of the primary coolant circuit, containment spraying, controlled venting, external vessel cooling, erosion of the lower foundation raft by the corium). (J.S.)

  15. Simulator experiments: effects of experience of senior reactor operators and of presence of a shift technical advisor on performance in a boiling water reactor control room

    International Nuclear Information System (INIS)

    This report describes the first experiment in a Nuclear Regulatory Commission-sponsored program of training simulator experiments and field data collection to evaluate the effects of selected performance shaping factors on the performance of nuclear power plant control room operators. The factors investigated were the experience level of the Senior Reactor Operator (SRO) and the presence of a Shift Technical Advisor (STA). Data were collected from 16 two-man crews of licensed operators (one SRO and one RO). The crews were split into high and low SRO-experience groups on the basis of the years of experience of the SROs as SROs. One half (4 of the 8 crews in each group) of the high- and low-SRO experience groups were assisted by an STA or an SRO acting as an STA. The crews responded to four simulated plant casualties which ranged in severity from an uncomplicated turbine trip to an anticipated transient without scram (ATWS). No significant differences in overall performance were found between groups led by high (25 to 114 months licensed as an SRO) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have shorter task performance times than crews led by high experience SROs. Although a tendency for the STA-assisted groups to score higher on four of the five measures was observed, the presence of the STA had no statistically significant effect on overall team performance. The correlation between individual performance, as measured by four of the task performance measures, and experience, measured by months as a licensed operator, was not statistically significant, nor was the correlation between task performance and recency of simulator training. 18 references, 5 figures, 13 tables

  16. Fuel recycling in boiling water reactors

    International Nuclear Information System (INIS)

    The present study confirms the feasibility of inserting mixed-oxid-fuel assemblies (MOX-FA) in boiling-water reactors in conjunction with reactivity-equivalent uranium-fuel assemblies. First, the established calculation methods were extended according to the specific MOX-uranium mutual interaction effects. Then, typical bundle-structures were analysed according to their neutron-physical features. The reactor-simulations show a non-critical behaviour with respect to limiting conditions and reactivity control. The variation of the isotopic composition and the plutonium content with its effects on the physical features was considered. (orig.) With 6 refs., 3 tabs., 29 figs

  17. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  18. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Science.gov (United States)

    2013-06-14

    ... COMMISSION All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos... Licensees operate boiling-water reactors (BWRs) with Mark I and Mark II containment designs. II. The events... Boiling Water Reactors with Mark I and Mark II Containments'' (November 26, 2012). Option 2 in...

  19. Air–water downscaled experiments and three-dimensional two-phase flow simulations of improved steam separator for boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Katono, Kenichi, E-mail: kenichi.katono.kq@hitachi.com [Hitachi, Ltd., Hitachi Research Laboratory, 1-1, Omika-cho 7-chome, Hitachi-shi, Ibaraki-ken 319-1292 (Japan); Ishida, Naoyuki [Hitachi, Ltd., Hitachi Research Laboratory, 1-1, Omika-cho 7-chome, Hitachi-shi, Ibaraki-ken 319-1292 (Japan); Sumikawa, Takashi; Yasuda, Kenichi [Hitachi-GE Nuclear Energy, Ltd., Hitachi Works, 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken 317-0073 (Japan)

    2014-10-15

    Highlights: • We design the improved steam separator for boiling water reactor (BWR). • The improved steam separator comprises the swirler vanes in the first-barrel section. • We evaluate separator performance by an air–water experiments and flow simulation. • The improved separator can decrease the total pressure losses by about 30%. • And the carryover performance is almost the same level as the conventional separator. - Abstract: Reducing the pressure losses in steam-separator systems in boiling water reactor (BWR) plants is useful for reducing the required pump head and enhancing the design margins to ensure core stability. We need to reduce the pressure losses while maintaining the gas–liquid separation performance. In this study, we improve a steam separator with air–water downscaled experiments and two-phase flow simulations. First, we confirm the effectiveness for the separator performance prediction by adjusting the quality and the two-phase centrifugal force between the air–water downscaled experiments and the steam-water mockup tests, and we design the improved steam separator, which moves the swirl-vane section from diffuser section to the first-barrel section. From the air–water downscaled experiments, the improved separator can decrease pressure loss in the swirler more than 50% around the BWR normal operating conditions compared to the conventional separator, and the carryover of the improved separator is almost the same level as the conventional separator. Next, we evaluate the improved steam separator performance under the BWR operating conditions by means of a two-phase flow simulation, and we have the prospects of the improved separator for reducing the total separator pressure losses by about 30% compared to the conventional separator, while maintaining carryover characteristics.

  20. Mitigation performance indicator for boiling water reactors

    International Nuclear Information System (INIS)

    All U.S. boiling water reactors (BWRs) inject hydrogen for mitigation of intergranular stress corrosion cracking (IGSCC), and most currently use or plan to use noble metals technology. The EPRI Boiling Water Reactor Vessels and Internals Project (BWRVIP) developed a Mitigation Performance Indicator (MPI) in 2006 to accurately depict to management the status of mitigation equipment and as a standardized way to show the overall health of reactor vessel internals from a chemistry perspective. It is a 'Needed' requirement in the EPRI BWR Water Chemistry Guidelines that plants have an MPI, and use of the BWRVIP MPI is a 'Good Practice'. The MPI is aligned with inspection relief criteria for reactor piping and internal components for U.S. BWRs. This paper discusses the history of the MPI, from its first use for plants operating with moderate hydrogen water chemistry (HWC-M) or Noble Metal Chemical Application (NMCA) + HWC to its more recent use for plants operating with On-Line NobleChem™ (OLNC) + HWC. Key mitigation parameters are discussed along with the technical bases for the indicators associated with the parameters. (author)

  1. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  2. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A boiling water reactor fuel assembly is described which has vertical fuel rods and guide tubes positioned below the fuel rods and receiving control rod fingers and acting as water pipes, the guide tubes each being formed of a plurality of parts including a part secured to a grid plate positioned in the fuel assembly container, and low parts which fit into holes formed in the bottom of the fuel assembly. There is a flexible connection between the upper and lower parts of the guide tubes to allow for a certain tolerance in the procedure of manufacturing the various parts to allow insertion of the fuel rod bundle into the fuel assembly container

  3. Simulation of Boiling Water Reactor dynamics

    International Nuclear Information System (INIS)

    This master thesis describes a mathematical model of a boiling water reactor and address the dynamic behaviour of the neutron kinetics, boilding dynamics and pressur stability. The simulation have been done using the SIMNON-program. The meaning were that the result from this work possibly would be adjust to supervision methods suitable for application in computer systems. This master thesis in automatic control has been done at the Department of Automatic Control, Lund Institute of Technology. The initiative to the work came from Sydkraft AB. (author)

  4. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A fuel assembly for a boiling water reactor comprises a plurality of fuel rods which constitute four partial bundles and are surrounded by a fuel channel system comprising one partial tube for each partial bundle. Each of the four partial bundles rests on a bottom tie plate and is positioned with respect to the others by means of a common top tie plate which is provided with a lifting loop which is sufficiently strong to be able to lift the four partial bundles simultaneously, a major part of the lifting force being transmitted to said bottom tie plates via a plurality of supporting fuel rods

  5. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    Science.gov (United States)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  6. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity. The feedback reactivity is a function of the operating conditions of the system, and cannot be directly measured. However, power measurements are relatively easy to obtain from the nuclear instrumentation and process computer, and are used in conjunction with a reduced order model to estimate the gain of the thermal-hydraulics feedback using an Extended Kalman Filter. The reduced order model is obtained by estimating the thermal-hydraulic transfer function from the frequency-domain BWR code LAPUR, and the stability properties are evaluated based on the pair of complex conjugate eigenvalues. Because of the recursive nature of the Kalman Filter, an estimate of the decay ratio is generated every sampling time, allowing continuous estimation of the stability parameters. A test platform based on a nuclear-coupled boiling channel is developed to validate the capability of the BWR stability monitoring methodology. The thermal-hydraulics for the boiling channel is modeled and coupled with neutron kinetics to analyze the non-linear dynamics of the closed-loop system. The model uses point kinetics to study core-wide oscillations, and normalized modal kinetics are introduced to study out-of-phase oscillations. The coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate noisy power time series. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions

  7. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    International Nuclear Information System (INIS)

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  8. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  9. Early detection of coolant boiling in research reactors with MTR-type fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozma, R.; Turkcan, E.; Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs.

  10. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    In this paper, a reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University Delft, The Netherlands

  11. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

  12. Detonating gas in boiling water reactors

    International Nuclear Information System (INIS)

    The radiation in the core region of Boiling Water Reactors (BWRs) decomposes a small fraction of the coolant into hydrogen and oxygen, a phenomenon termed radiolysis. The radiolysis gas partitions to the steam during boiling. A 1000 MWe BWR produces around 1.5 tons of steam, containing 25 grams of radiolysis gas, per second. Practically all of the radiolysis gas is carried to the condenser and is taken care of by the condenser evacuation system and the off-gas system. The operation of these systems has been largely trouble-free. Radiolysis gas may also accumulate when stagnant steam condenses in pressurized pipes and components as a result of heat loss. Under certain circumstances a burnable mixture of hydrogen, oxygen and steam may form. Occasionally, the accumulated radiolysis gas has ignited. These incidents typically result in deformation of the components involved, but overpressure bursts have also occurred. Radiolysis gas accumulation in steam systems was largely overlooked by BWR designers (a likely technical reason for this is given in the report) and the problem had to be addressed by utilities. Even though the problem was recognized two decades ago, the counter-measures of today seem not always to be sufficient. Pipe-burst incidents in a German and a Japanese BWR recently attracted attention. Also, damage to a pilot valve in the steam relief system of a Swedish BWR forced a reactor shut-down during 2002. The recent incidents indicate that counter-measures against radiolysis gas accumulation in BWRs should be reviewed, perhaps also improved. The present report provides a short compilation of basic information related to radiolysis gas accumulation in BWRs. It is hoped that the compilation may prove useful to utilities and regulators reviewing the problem

  13. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  14. Design certification program of the simplified boiling water reactor

    International Nuclear Information System (INIS)

    General Electric (GE), the US Department of Energy, the Electric Power Research Institute (EPRI), and utilities are undertaking a cooperative program to enable advanced light water reactor (ALWR) designs to be certified by the US Nuclear Regulatory Commission (NRC). GE is seeking to certify two advanced plants; the Advanced Boiling Water Reactor (ABWR) and the Simplified Boiling Water Reactor (SBWR). Both plants use advanced features that build on proven BWR technology

  15. Dynamic reconstruction and Lyapunov experiments from time series data in boiling water reactors. Application to B.W.R. stability analysis

    International Nuclear Information System (INIS)

    This paper shows how to obtain Lyapunov exponents from time series data on Boiling Water Reactor (BWR) stability. In order to validate the method, these characteristic exponents are compared with the ones obtained directly from the governing equations of the dynamic system. Finally, we present a method for obtaining the stability of the B.W.R. from Lyapunov exponents and describe some other applications related to limit cycles. (Author)

  16. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  17. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  18. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. I. Observation of quenching of downward-facing surfaces

    International Nuclear Information System (INIS)

    For pt.II see ibid., p.89-99 (1997). This paper presents the results of a series of quenching experiments to examine the boiling curve of relative large downward-facing surfaces. The test masses are 61 cm in diameter and the downward-facing surface is either flat or curved. The work is motivated by the need to assess the ex-vessel boiling process for in-vessel core retention. The critical heat flux is found to be approximately 50 W cm-2. The nucleate boiling regime and the critical heat flux regime are found to be characterized by cyclic two-phase flow patterns. (orig.)

  19. Study of neutron noise physical model for reactor coolant boiling

    International Nuclear Information System (INIS)

    The neutron noise method has been used to monitoring reactor coolant boiling. Wach-Kosaly model has been used to interpret the neutron noise induced by coolant boiling. The equation based on the model is got and used for calculation. The physical variable with the relation of bubble's velocity is got from the calculated result (autopower spectral density)

  20. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  1. Experience with primary water cleaning and waste water treatment plant in nuclear power stations with pressurised and boiling water reactors

    International Nuclear Information System (INIS)

    Powder resin alluvial filtration using structured filter layers permits constantly improving adaption of the water treatment technology to even the most demanding problem situations - particularly in the field of primary water and waste water treatment in nuclear power stations. From experience in operation the authors show the advantages of this technique compared to other techniques, which can be deduced from theoretical concepts, taking into account the various target figures decisive in operating nuclear power stations. (orig.)

  2. SWR 1000: the Boiling Water Reactor of the future

    International Nuclear Information System (INIS)

    Siemens Power Generation Group (KWU) is currently developing - on behalf of and in close cooperation with the German nuclear utilities and with support from various European partners - Germany's next generation of boiling water reactor. This innovative design concept marks a new era in the successful tradition of boiling water reactor technology and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared lo large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. In addition, a state-of-the-art materials concept featuring erosion-resistant materials and low-cobalt alloys as well as cobalt-free substitute materials ensures a low cumulative dose for operating and maintenance personnel and also minimizes radioactive waste. (author)

  3. SWR 1000: The new boiling water reactor power plant concept

    International Nuclear Information System (INIS)

    Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

  4. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  5. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.)

  6. Resolution of US regulatory issues involving boiling water reactor stability

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) and the Boiling Water Reactor Owners Group (BWROG) have been reexamining BWR instability characteristics and consequences since the March 1988 instability event at LaSalle Unit 2. The NRC and BWROG concluded that existing reactor protection systems do not prevent violation of the critical power ratio (CPR) safety limits caused by large asymmetric oscillations. The studies are also examining the need to modify the automatic and operator actions previously developed for response to an anticipated transient without scram (ATWS) event because of oscillation effects not fully considered in previous studies. This paper presents the current status of these studies and an assessment of actions needed to resolve the issue. (author)

  7. Factor analysis of flow instability between heating reactor and boiling reactor

    International Nuclear Information System (INIS)

    The affections on instability of sub-cooled boiling, condensation, flashing, and steam-space pressure equilibrium, have been analyzed on the base of experimental results in natural circulation system with low pressure and low steam quality, and compared with those of boiling reactor. The analysis shows: (1) Sub-cooled boiling, condensation and flashing play an important role on the flow instabilities in a natural circulation system, and have connection with lots of instabilities, which is different to the forced circulation system; (2) The size of steam-space is important to the system stability. In order to improve it, the size of steam-space should be small to the best of its ability, if the system can support the large pressure

  8. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    The performance of the boiling detection system has been tested on boiling signals coming from the research reactor HOR during experiments with the NIOBE boiling setup. Several detection methods utilizing frequency domain analysis have been tested both on- and off-line. Results of these methods indicate that boiling detection is possible in real-time even in the incipient stage of the boiling. Both DC and AC components of the in-core and ex-core neutron detector signals can be used for boiling detection; these two components provide complementary information. Advanced signal analysis application to the DC signals may give information about the dynamic changes of the reactor, provided that the changes of the signal exceed the inherent noise of the measured channel. At the same time, AC signal analysis will characterize the changes even in the inherent signal fluctuation level. Boiling experiments of HOR and the methods implemented for signal analysis validates the techniques used for these experiments. (orig./HP)

  9. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Tuerkcan, E. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kozma, R. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Nabeshima, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1993-01-01

    The performance of the boiling detection system has been tested on boiling signals coming from the research reactor HOR during experiments with the NIOBE boiling setup. Several detection methods utilizing frequency domain analysis have been tested both on- and off-line. Results of these methods indicate that boiling detection is possible in real-time even in the incipient stage of the boiling. Both DC and AC components of the in-core and ex-core neutron detector signals can be used for boiling detection; these two components provide complementary information. Advanced signal analysis application to the DC signals may give information about the dynamic changes of the reactor, provided that the changes of the signal exceed the inherent noise of the measured channel. At the same time, AC signal analysis will characterize the changes even in the inherent signal fluctuation level. Boiling experiments of HOR and the methods implemented for signal analysis validates the techniques used for these experiments. (orig./HP)

  10. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  11. Uncommon water chemistry observations in modern day boiling water reactors

    International Nuclear Information System (INIS)

    Numerous technologies have been developed to mitigate intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) materials that include hydrogen water chemistry (HWC), noble metal chemical application (NMCA) and on-line NMCA (OLNC). These are matured technologies with extensive plant operating experiences, HWC – 32 years, NMCA – 18 years and OLNC – 9 years. Over the past three decades, numerous water chemistry data, dose rate data and IGSCC mitigation data relating to these technologies have been published and presented at many international conferences. However, there are many valuable and critical water chemistry and dose rate data that have gone unnoticed and unreported. The purpose of this paper is to highlight some of the uncommon water chemistry and dose rate experiences that reveal valuable information on the performance and durability of NMCA and OLNC technologies. Data will be presented, that have hitherto been unseen in public domain, from the lead OLNC plant in Switzerland giving reasons for some of the uncommon or overlooked water chemistry observations. They include, decreasing reactor water platinum concentration with each successive OLNC application, lack of increase in reactor water activation products in later applications, gradual disappearance of main steam line radiation (MSLR) monitor response decrease, Curium and Au-199 release during OLNC applications, rapid increase in reactor water clean-up conductivity, and Iodine, Mo-99 and Tc-99m spiking when hydrogen is interrupted and brought back to service, and main steam and reactor water conductivity spiking when clean-up beds or condensate demineralizers are changed. All these observations give valuable information on the success of OLNC applications and also signal the presence of sufficient noble metal on in-reactor surfaces from the long term durability and effectiveness stand point. Some of these observations can be used as secondary parameters, if and when a primary

  12. Stability monitoring of the Dodewaard boiling-water reactor

    International Nuclear Information System (INIS)

    Methods for measuring the stability of a boiling-water are discussed. The results of experiments performed on the Dodewaard reactor (The Netherlands) are reported. Research on this reactor is of interest as it is cooled by natural circulation, a cooling principle that is also being considered for new reactor design. The stability of the Dodewaard reactor was studied both with deterministic methods (control-rod steps and pressure-valve movements) and by noise analysis. The latter method can be applied during normal operation and avoids any intentional system disturbance. Reactorkinetic stability, thermal-hydraulic stability and total-plant stability were investigated separately. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced; it was tested thorougly. It can be derived on-line from the noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations was calculated in order to assure a proper stability surveillance. A novel technique is presented, which enables the variations of the in-core coolant velocity to be determined by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies were performed on the fuel time constant, a parameter of great importance to the reactor-kinetic stability. It is shown that the effective value of this constant can be much smaller than the value commonly agreed on (author). 71 figs.; 73 figs,; 21 tabs

  13. Modelling of subcooled boiling in ATHLET and application in water cooled research reactors

    International Nuclear Information System (INIS)

    A model is implemented to describe the thermodynamic nonequilibrium effects in subcooled boiling regime. The aim is to simulate void distribution and thermodynamic instability, which is practicularly pronounced in research reactors due to high power densities and low system pressures, and to include the influence of the steam formed in this boiling regime on the neutron balance. The model developed considers the competing effects of vaporization and condensation during subcooled boiling. It describes the rate of bubble generation on superheated surfaces and the subsequent condensation of steam in the subcooled liquid. The installed model is validated by postcalculations of two extensive series of experiments. The extended and verified program is used to simulate the Juelich research reactor FRJ-2. For this purpose, a full-scale simulation model of the entire plant is developed ensuring, in particular, a precise reproduction of the geometry and the arrangement of the annular fuel element cooling channels. The modelled reactor plant is first used to simulate normal reactor operation. The resulting steady-state temperature and pressure distributions assuming a thermal power of 23 MW show good agreement with real operating data. Safety investigations are conducted to examine plant behaviour under design-basis accident conditions. This includes failure of all three main coolent pumps with proper and delayed reactor scram. In both cases, the simulation shows that the fuel elements are not endangered in any phase of the transient, although in the event of a delayed scram initial signs of parallel channel instability due to steam formation in the central fuel element are to be observed which, however, only prevails for a short period of 30 ms. (orig./HP)

  14. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

    During the next decade a rise in the energy demand is expected worldwide, and this will in particular call for electricity generation capacity. A number of old generating plants, both nuclear and other plants, will probably have to be shut down for aging reasons, and their replacement will enhance the need for new generating capacity. The ABB Atom considers this situation to be met with a 'cautious evolution'. The offerings will largely be based on 'evolutions' of the successful light water reactor BWR 75. The new, evolutionary plant design of ABB Atom is the BWR 90. It can be designed, licensed and constructed in accordance with any safety regulations now in force or envisaged in the Western world. Emphasis has been, and will be, placed on features that facilitates licensing, shortens construction time and keeps electricity generation costs favourable. ABB also continues to develop a design of the 'passive' type, such as the 'passive' PIUS system, for possible deployment in the future. These efforts are more long-term activities, since development, verification and licensing of distinctly 'new' reactor concepts will have an extensive lead time. This paper presents the BWR 90 and its current status. The design is based on that of its forerunner, the BWR 75 standard design, taking into account the experiences gained from design and engineering, construction, commissioning, and operation of BWR 75 plants, the needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings. (author) 4 figs

  15. Results of boil-off experiment QUENCH-11

    International Nuclear Information System (INIS)

    The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor (LWR). The typical QUENCH test bundle consists of 21 fuel rod simulators with a total length of approximately 2.5 m. Boil-off experiment QUENCH-11 was performed at Forschungszentrum Karlsruhe (Karlsruhe Research Center) on 8 December, 2005 as the second of two experiments in the frame of the EC-supported LACOMERA program. The experiment focused on studying bundle behavior during boil-off and subsequent quenching at a small water injection rate. It was proposed by INRNE Sofia (Bulgarian Academy of Sciences) and defined together with the Karlsruhe Research Center. The analytical support by using the SCDAP/RELAP5 mod 3.2.irs and ASTEC codes for the preparation of the entire test was essential in conducting the test. A steady boil-off and a corresponding top-down uncovery of the test bundle was achieved by applying power from an electric auxiliary heater at the bundle bottom in addition to the electric bundle power. An additional outer heating system compensated heat losses that would lead to a reduction in boiling off the covered part of the bundle. When the water level had fallen to -70 mm elevation water was injected into the lower plenum at a rate of ca. 1 g/s enabling a nearly stable water level and an extension of the boil-off phase. Quenching of the bundle was performed at a maximum measured bundle temperature of 2040 K with a rather low mass flow rate of water, i.e. 18 (17+1) g/s, compared to the standard water injection rate of approx. 50 g/s. The conditions led to an enhanced cladding and shroud oxidation, quite similar to standard conditions of forced-convection steam flow. In the upper part the test bundle was significantly degraded by oxidation and melt formation. The total generation of hydrogen measured by the mass spectrometer was 141 g, of which 132 g, i.e. more than 90 % of the total, was produced during

  16. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  17. Boiling water reactor stability analysis in the time domain

    International Nuclear Information System (INIS)

    Boiling water nuclear reactors may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate due to the tight coupling of flow to power, especially under gravity-driven circulation. In order to predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model has been developed for a typical boiling water reactor. Using this tool it has been demonstrated that density waxes may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases have been analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. From that study it has been concluded that two-phase friction controls the extent of the oscillation and that the existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from case to case. It has also been determined that higher dimensional nuclear feedback models reduce the extent of the oscillation. It has also been confirmed from a nonlinear dynamic standpoint that the birth of this oscillation may be described as a Hopf Bifurcation

  18. Radial nodalization effects on BWR [boiling water reactor] stability calculations

    International Nuclear Information System (INIS)

    Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using ''representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs

  19. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  20. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas...

  1. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  2. Self-Sustaining Thorium Boiling Water Reactors

    International Nuclear Information System (INIS)

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  3. Two-dimensional simulation of the downcomer boiling experiment using the CUPID code

    International Nuclear Information System (INIS)

    For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. We simulated the DOBO (Downcomer Boiling) experiment in two-dimensions using the CUPID code to evaluate its two-phase flow models and verify its applicability to the downcomer boiling analysis. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change and a circulation of liquid accelerated by bubbles. The two-phase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling. (author)

  4. On the possibility of sodium boiling detection in the BN-600 reactor by neutron noises

    International Nuclear Information System (INIS)

    The possibility of early diagnostics of sodium leakage and its boiling in the BN-600 reactor fuel assembley on the basis of neutron flux or reactivity noise analysis is studied. So determine the nature of integral and local neutron flux changes under fuel assembley blockage calculated and experimental data, obtained at the BOR-60 reactor, are analysed. Calculations are carried out using the NF-6 program complex. The reaction of local neutron flux monitors, made of rhodium, was determined during the experiments besides reactivity change Δk/k measuring. It is ascertained, that the effect of a fuel assembly complete devastation depending ion its location in the zone changes within the range from -10 Δk/k up to 2x10-5 Δk/k. The amplitude of signal pulsations of the neutron flux monitor, installed on the turning plus bottom plate, is 0.25-0,51%, taking into account its dynamic characteristics. It means, that using one of such monitors it is possible point of the reactor core. So register a weaker boiling a number local monitors will be needed

  5. Interim report on the result of the sodium boiling detection benchmark test using BOR-60 reactor noise data

    International Nuclear Information System (INIS)

    The present paper deals with the second stage of investigations of acoustic signals from a boiling experiment performed on the KNS I loop at KfK Karlsruhe and first results of analysis of data from a series of boiling experiments carried out in the BOR 60 reactor in the USSR. Signals have been analysed in frequency as well as in time domain. Signal characteristics successfully used to detect the boiling process have been found in time domain. A proposal for in-service boiling monitoring by acoustic means is briefly described. (author). 1 ref., 8 figs, 1 tab

  6. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  7. Fundamentals of boiling water reactor systems

    International Nuclear Information System (INIS)

    The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK)

  8. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  9. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  10. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition, numerica

  11. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    International Nuclear Information System (INIS)

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended

  12. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    Energy Technology Data Exchange (ETDEWEB)

    Aksan, S.N.; Stierli, F.; Analytis, G.T. [Paul Scherrer Inst. (PSI), Villigen (Switzerland). Lab. for Thermal-Hydraulics

    1992-03-01

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended.

  13. Overview of activities for the reduction of dose rates in Swiss boiling water reactors

    International Nuclear Information System (INIS)

    Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

  14. GE simplified boiling water reactor stability analysis in time domain

    Science.gov (United States)

    Lu, Shanlai

    1997-12-01

    General Electric Simplified Boiling Water Reactor (SBWR) was designed as a next generation light water reactor. It uses natural circulation to remove the heat from the reactor core. Because of this unique in-vessel circulation feature, SBWR is expected to exhibit different stability behaviors. The main emphasis of this thesis is to study the SBWR stability behavior in the time domain. The best-estimate BWR accident/transient analysis computer code, TRAC-BF1, is employed to analyze the SBWR stability behavior. A detailed TRAC-BF1 SBWR model has been developed, which has the capability to model the in-vessel natural circulation and the reactor core kinetics. The model is used to simulate three slow depressurization processes. The simulation results show that the reactor is stable under low pressure and nominal downcomer water level conditions. However, when the downcomer water level is raised to about 19.2 m above the bottom of the reactor vessel, an unstable power oscillation is observed. The identified power oscillation is further analyzed using TRAC-BF1 1-D kinetics and the new TRAC-BF1 3-D kinetics code developed in this thesis. The effects of different time step sizes and vessel model nodalizations are examined. It is found that the power oscillation is in-phase and has a frequency of 0.3 HZ. In order to further explore the physical instabilty initiation mechanisms, a simplified dynamic model consisting of six simple differential equations is developed. The simplified model is able to predict the dominant physical phenomenon identified by the TRAC-BF1 analysis. The results indicate that the system instability is possibly caused by the steam separator hydro-static head oscillation under the high water level condition. In order to explore the higher order spacial effect of power oscillation, a 3-D reactor core kinetics code is coupled with the TRAC-BF1 computer code in the PVM parallel processing environment. A new coupling scheme and a multiple time step marching

  15. Boiling in the presence of boron compounds in light water reactors

    International Nuclear Information System (INIS)

    The scope of the thesis on boiling in the presence of boron compounds in light water reactors was to study the effects of the boron compound addition on the heat removal from the fuel elements. For an effective cooling of the fuel elements in case of boiling processes a high heat transfer coefficient is of importance. Up to now experimental studies were not performed under reactor specific conditions, for instance with respect to the geometry of the flow conditions, high temperature and pressure levels were not represented. Therefore the experiments in the frame of the thesis were using reactor specific parameters. The test facility SECA (study into the effects of coolant additives) was designed and constructed. The experiments simulated the conditions of normal PWR operation, accidental PWR and accidental BWR conditions.

  16. Flow boiling in microgap channels experiment, visualization and analysis

    CERN Document Server

    Alam, Tamanna; Jin, Li-Wen

    2013-01-01

    Flow Boiling in Microgap Channels: Experiment, Visualization and Analysis presents an up-to-date summary of the details of the confined to unconfined flow boiling transition criteria, flow boiling heat transfer and pressure drop characteristics, instability characteristics, two phase flow pattern and flow regime map and the parametric study of microgap dimension. Advantages of flow boiling in microgaps over microchannels are also highlighted. The objective of this Brief is to obtain a better fundamental understanding of the flow boiling processes, compare the performance between microgap and c

  17. Calculation of limit cycle amplitudes in commercial boiling water reactors

    International Nuclear Information System (INIS)

    This paper describes an investigation of the dynamic behavior of a boiling water reactor (BWR) in the nonlinear region corresponding to linearly unstable conditions. A nonlinear model of a typical BWR was developed. The equations underlying this model represent a one-dimensional void reactivity feedback, point kinetics with a single delayed neutron group, fuel behavior, and recirculation loop dynamics (described by a single-node integral momentum equation)

  18. Core surveillance of boiling-water reactors

    International Nuclear Information System (INIS)

    Methods suitable for a calculational procedure which determines the three-dimensional power distribution in boilingwater reactors on the basis of in-core detector readings are described. A two- dimensional equation based on diffusion theory is set up, and a method for incorporating detector readings in the solution of this equation is presented. A three-dimensional calculational method based on nodal theory is developed. Calculations are carried out using this method, and the results are compared with a three-dimensional nodal theory calculation . Finally, parameters affecting the detector readings are examined. (author)

  19. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices

  20. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  1. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  2. Evaluation of damages of airplane crash in European Advanced Boiling Water Reactor (EU-ABWR)

    International Nuclear Information System (INIS)

    European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash (APC), severe accident mitigation systems, N+2 principle in safety systems and a large output of 1600 MWe. Thanks to above mentioned features, EU-ABWR's design objectives and principles are consistent with safety requirements in an European market. In this paper, evaluation of damages induced by APC has been summarized. (author)

  3. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 2: A survey of the accuracy of the Studsvik of America CMS codes

    International Nuclear Information System (INIS)

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. Thus, before performing any kind of calculation with MOx fuels, it is necessary to be able to establish the reliability and the accuracy of these Core Management System (CMS) codes. This report presents a quantitative analysis of the models used in the package. A qualitative presentation is realized in a coming report

  4. Boils

    Science.gov (United States)

    ... the boil is very bad or comes back. Antibacterial soaps and creams cannot help much once a boil ... following may help prevent the spread of infection: Antibacterial soaps Antiseptic (germ-killing) washes Keeping clean (such as ...

  5. Invited talk on ageing management of boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    A nuclear power plant is built with a certain design life but by managing the operation of the plant with a well designed in-service inspection, repair and replacement programme of the equipment as required we will be able to extend the operation of the plant well beyond it's design life. This is also economically a paying proposition in view of the astronomical cost of construction of a new plant of equivalent capacity. In view of this, there is a growing trend the world over to study the ageing phenomena, especially in respect of nuclear power plant equipment and system which will contribute towards the continued operation of the nuclear power plants beyond their economic life which is fixed mainly to amortize the investments over a period. Tarapur Atomic Power Station (TAPS) which consists of 2 nos. of Boiling Water Reactor (BWRs) with the presently rated capacity of 160 MWe each has been operating for the past 24 years and is completing its 25th year of service by the year 1994 which was considered as its economic life and the plant depreciation as well as fuel supply agreement were based on this period of 25 years. I will be discussing about the available residual life which is much more than the above (25 years) and the studies we have undertaken in respect of the assessment of this residual life. (author). 2 tabs., 6 figs

  6. Boiling Experiment Facility for Heat Transfer Studies in Microgravity

    Science.gov (United States)

    Delombard, Richard; McQuillen, John; Chao, David

    2008-01-01

    Pool boiling in microgravity is an area of both scientific and practical interest. By conducting tests in microgravity, it is possible to assess the effect of buoyancy on the overall boiling process and assess the relative magnitude of effects with regards to other "forces" and phenomena such as Marangoni forces, liquid momentum forces, and microlayer evaporation. The Boiling eXperiment Facility is now being built for the Microgravity Science Glovebox that will use normal perfluorohexane as a test fluid to extend the range of test conditions to include longer test durations and less liquid subcooling. Two experiments, the Microheater Array Boiling Experiment and the Nucleate Pool Boiling eXperiment will use the Boiling eXperiment Facility. The objectives of these studies are to determine the differences in local boiling heat transfer mechanisms in microgravity and normal gravity from nucleate boiling, through critical heat flux and into the transition boiling regime and to examine the bubble nucleation, growth, departure and coalescence processes. Custom-designed heaters will be utilized to achieve these objectives.

  7. Electrochemical sensors for application to boiling water reactors

    International Nuclear Information System (INIS)

    An effective measure in combating the intergranular cracking of stainless steel in Boiling Water Reactors (BWRs) is the control of the electrochemical corrosion potential (ECP). It has been found that when the ECP of austenitic stainless steel alloys susceptible to cracking is decreased below -0.230 V(SHE) cracks will not initiate. Similarly, the decrease in potential decreases crack growth rate of existing cracks. Decrease of the potential to acceptable values is accomplished by addition of hydrogen to the reactor feedwater. The amount of hydrogen required is determined by the ECP measured at high temperature either in-situ or from a water sample delivered to an external monitoring station. Both reference and metal sensor electrodes are required to determine the ECP. A multiplicity of reference electrodes are used to verify the validity of the measurements. The reference electrodes, Ag/AgCl, the yttria-stabilized ZrO2 sensor and the platinum electrode are designed for either remote, high radiation environments or accessible monitoring installation at plant operating temperatures. In the former application the support structure for the electrochemical sensor is fabricated from ceramics, usually sapphire, and ceramic-to-metal brazes are used for seals. Metal-to-metal seals are welds. For accessible installations high temperature elastomeric seals are used as long as some periodic maintenance is possible. Just as the reference electrodes are designed for remote or accessible installation, the metal sensor electrodes, principally stainless steel, can be manufactured with ceramic-to-metal brazes or elastomeric seals. The complete electrochemical package, with data acquisition system, is then used by plant personnel to control the feedwater H2 injection rate for environmental crack mitigation

  8. Electrochemical sensors for application to boiling water reactors

    International Nuclear Information System (INIS)

    An effective measure in combating the intergranular cracking of stainless steel in Boiling Water Reactors (BWRs) is the control of the electrochemical corrosion potential (ECP). It has been found that when the ECP of austenitic stainless steel alloys susceptible to cracking is decreased below -0.230 V(SHE) cracks will not initiate. Similarly, the decrease in potential to acceptable values is accomplished by addition of hydrogen to the reactor feedwater. The amount of hydrogen required is determined by the ECP measured at high temperature either in-situ or from a water sample delivered to an external monitoring station. Both reference and metal sensor electrodes are required to determine the ECP. A multiplicity of reference electrodes are used to verify the validity of the measurements. The reference electrodes, Ag/AgCl, the yttria-stabilized ZrO2 sensor and the platinum electrode are designed for either remote, high radiation environments or accessible monitoring installation at plant operating temperatures. In the former application the support structure for the electrochemical sensor is fabricated from ceramics, usually sapphire, and ceramic-to-metal brazes are used for seals. Metal-to-metal seals are welds. For accessible installations high temperature elastomeric seals are used as long as some periodic maintenance is possible. Just as the reference electrodes are designed for remote or accessible installation, the metal sensor electrodes, principally stainless steel, can be manufactured with ceramic-to-metal brazes or elastomeric seals. The complete electro-chemical package, with data acquisition system, is then used by plant personnel to control the feedwater H2 injection rate for environmental crack mitigation

  9. Nuclear reactor noise investigations on boiling effects in a simulated MTR-type fuel assembly. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Kozma, R.

    1992-05-04

    Boiling experiments at HOR have a relatively long history that began more than a decade ago. Following some introductory out-of-pile experiments, a boiling setup was operating at HOR between 1984 and 1986. Based on the experience of that setup, a new boiling experiment, NIOBE (Noise Investigations On Boiling Effects), has been designed. The NIOBE setup has been in operation since the end of 1986. The present work contains results of experiments performed between 1988 and 1991, with the exception of Chapter 5, in which experiments with the predecessor of NIOBE are treated. The thesis is based on results of experiments performed at the NIOBE loop located in a research reactor (HOR).

  10. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

  11. Automatic thickness measuring system of zirconium and zircaloy-2 layers of zirconium liner cladding tubes for boiling water reactor

    International Nuclear Information System (INIS)

    An automatic measuring system using ultrasonic method and electromagnetic method has been developed to measure the thickness of zirconium and zircaloy-2 layers. The sophisticated mechanism and the unique signal processing for suppression of several types of error enable high accurate measurement. The standard deviation of the liner thickness measurement is 2.2 μm and that of mother layer measurement is 3.0 μm. This system is very useful to assure the thickness of each layer and to produce high quality zirconium liner cladding tubes. (author)

  12. Recycling heterogeneous americium targets in a boiling water reactor

    International Nuclear Information System (INIS)

    One of the limiting contributors to the heat load constraint for a long term spent fuel repository is the decay of americium-241. A possible option to reduce the heat load produced by Am-241 is to eliminate it via transmutation in a light water reactor thermal neutron environment, in particular, by taking advantage of the large thermal fission cross section of Am-242 and Am-242m. In this study we employ lattice loading optimization techniques to define the loadings and arrangements of fuel pins with blended americium and uranium oxide in boiling water reactor bundles, specifically, by defining the incineration of pre-loaded americium as an objective function to maximize americium transmutation. Subsequently, the viability of these optimized lattices is tested by assembling them into bundles with Am-spiked fuel pins and by loading these bundles into realistic three-dimensional BWR core-wide simulations that model multiple reload cycles and observe standard operational constraints. These simulations are possible via our collaboration with the Westinghouse Electric Co. which facilitates the use of industrial-caliber design tools such as the PHOENIX-4/POLCA-7 sequence and the Core Master 2 GUI work environment for fuel management. The resulting analysis confirms the ability to axially uniformly eliminating roughly 90% of the pre-loaded inventory of recycled Am-241 in BWR bundles with heterogeneous target pins. This high level of incineration was achieved within three to four 18-month operational cycles, which is equivalent to a typical in-core residence time of a BWR bundle.

  13. Proceedings of the International Workshop on Boiling Water Reactor Stability

    International Nuclear Information System (INIS)

    General design criteria for nuclear power plants in every OECD country require that the reactor core and associated coolant, control, and protection systems be designed so that power oscillations which can result in conditions exceeding acceptable fuel design limits are not possible, or they can be reliably and readily detected and suppressed. In practice, this means that reactor cores should be stable with regard to perturbations from their normal operating state, so that expected variations to the operating parameters do not induce undamped power oscillations. These power oscillations can take a variety of forms, from very local power peaks which can cause no damage, or only slight damage to only a few fuel rods, to large core-wide oscillations where entire segments of the core can become neutronically uncoupled, with wide power swings. Ever since the fast boiling water reactors began operating, over 30 years ago, it has been recognized that their operation under certain conditions of power and flow could cause power and flow oscillations. Considerable research was performed at that time to better understand the principal operating parameters which contribute to the initiation of these oscillations, and guidelines were developed to avoid plant operation under the conditions which were the most unstable. Experiments in the the first Special Power Excursion Reactor Test (SPERT-1) program produced spontaneous power oscillations, and investigations in an out-of-pile loop were necessary to demonstrate that the immediate cause of the oscillations was a power-to-reactivity feedback. Further investigations indicated that the instabilities were limited to certain areas on the operating map. These regions could not be absolutely defined, but there was sufficient understanding of them that they could be generally avoided, with only minor examples of instability events. More recently, though, several reactor events, and especially one that occurred at the La Salle Nuclear

  14. Channel-type nuclear reactor with a boiling coolant

    International Nuclear Information System (INIS)

    The invention is aimed at increasing the channel-type reactor safety, in particular, RBMK-type reactors, during accidents resulting in the coolant circulation discontinuation. The reactor core is assembled of vertial technological channels connected in parallel between distributing group collectors and drum-separator. Each technological channel contains a high pressure tube, a fuel assembly with fuel elements and a storage vessel located above the fuel assembly which is filled with water at saturation temperature in the normal operation regime. After dehydration of channels in the course of accident the boiling water from storage vessel is ejected into them. So the device described allows one to reduce the fuel element can temperature in the course of accidents connected with the coolant circulation discontinuation and so to increase the plant safety level

  15. Boiling water reactor operator training and qualification in Japan

    International Nuclear Information System (INIS)

    Nuclear power plant operators in Japan are individuals employed by each electric power company. A recruit goes through his company's training; afterwards, he is given a qualification rating and is assigned to practical duty. The only formal qualification authorized by the Japanese government is the full-fledged shift supervisor. Other classifications such as assistant shift supervisor, shift foreman, reactor operator, and subreactor operator are all designated and appointed by each company's in-house regulations. As a part of the training system, power companies that require the use of a full-scope simulator in their training programs utilize the boiling water reactor (BWR) and pressurized water reactor operator training centers. Both were set up independently of the power companies. A synopsis of the BWR Operator Training Center Corp. (BTC ) and its training systems, features, performance evaluation, curriculum improvement, and related items is presented

  16. Circulational characteristics of a natural circulation circuit of a weakly boiling reactor large-scale model

    International Nuclear Information System (INIS)

    A large-scale model for determining circulational characteristics of a natural circulation circuit of weakly boiling (core outlet steam content below 4%) tank tpype water cooled reactors is described. The model consists of 61 elecrtroheated fuel elements 14 mm in-diameter and 3 m height. Outlet pressure can vary within 1.7-5.0 MPa inlet water subcooling is 20-90 deg C, weight outlet balance steam content from-9 to 3.2 %. Results of the experiments performed for checking the algorithms developed for thermohydraulic calculation of steady-state characteristics of the investigated circuit are given. It is concluded that for one-phase coolant estimated and experimental values for pressure head and hydraulic resistance agree well with

  17. Digital control application for the advanced boiling water reactor

    International Nuclear Information System (INIS)

    The Advanced Boiling Water Reactor (ABWR) is a 1300 MWe class Nuclear Power Plant whose design studies and demonstration tests are being performed by the three manufacturers, General Electric, Toshiba and Hitachi, under requirement specifications from the Tokyo Electric Power Company. The goals are to apply new technology to the BWR in order to achieve enhanced operational efficiencies, improved safety measures and cost reductions. In the plant instrumentation and control areas, traditional analog control equipment and wire cables will be replaced by distributed digital microprocessor based control units communicating with each other and the control room over fiber optic multiplexed data buses

  18. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  19. On-site staffing requirements for a simplified boiling water reactor (SBWR)

    International Nuclear Information System (INIS)

    In 1992 the total generating costs were estimated by EPRI for a baseload, nth-of-a-kind advanced reactor with the following cost distribution: capital cost 62%, operation and maintenance (O and M) cost 20%, fuel cost 16%, and decommissioning cost 2%. Thus the O and M cost is a significant component of the total cost of electricity, second only to the capital cost. The O and M cost in turn can be split into: cost for on-site staff, maintenance materials, supplies and expenses, off-site technical support, regulatory fees, insurance premiums and administration. The costs for on-site staff is about 30% of the total O and M cost. In 1992, the US Council for Energy Awareness (USCEA) estimated the on-site staffing for a typical 600 MWe advanced reactor to be about 330 with 25 (full time equivalent, FTE) contractors. This estimate was reevaluated by EPRI, and the staffing was modified based on a reengineering of the organizational structure that eliminated unnecessary layers of vertical management. As a result of this review, the on-site staffing was decreased to 259 with 25 (FTE) contractors, for a total of 284 people. The Dodewaard power plant (GKN) in The Netherlands is a 60 MWe facility with a natural circulating reactor. Since the 600 MWe Simplified Boiling Water Reactor (SBWR), an advanced reactor, also utilizes a natural circulating reactor with other passive safety features it was desired to extrapolate the GKN staffing to the SBWR. Also, some of the European O and M practices that utilize fewer skilled labor are reflected. This paper provides the results of the comparison between the EPRI recommendations and the staffing based on GKN experience

  20. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  1. Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

    International Nuclear Information System (INIS)

    The Halden boiling heavy-water reactor, designed and built by the Norwegian Institutt for Atomenergi, has since June 1958 been operated as an international project. On its second charge the reactor was operated at power levels up to 25 MW and most of the time at a pressure of 28.5 kg/cm2. During the period from July 1964 to December 1966 the plant availability was close to 64% including shutdowns because of test fuel failures and loading/unloading of fuel. Disregarding such stops, the availability was close to 90%. The average burnup of the core is about 6200 MWd/t UO2 : the most highly exposed elements have reached 10000 MWd/t UO2. The transition temperature of the reactor tank has been followed closely. The results of the surveillance programme and the implication on the reactor operation are discussed. The reactor is located in a cave in a rock. Some experiences with such a containment are given. To locate failed test-fuel elements a fuel failure location system has been installed. A fission gas collection system has saved valuable reactor time during clean-up of the reactor system following test fuel failures. Apart from one incident with two of the control stations, the plant control and instrumentation systems have functioned satisfactorily. Two incidents with losses of 150 and 200 kg of heavy water have occurred. However, after improved methods for leakage detection had been developed, the losses have been kept better than 50 g/h . Since April 1962 the isotopic purity of the heavy water (14 t) has decreased from 99.75 to 99.62%. The tritium concentration is now slightly above 700 μC/cm3. This activity level has not created any serious operational or maintenance problems. An extensive series of water chemistry experiments has been performed to study the influence of various operating parameters on radiolytic gas formation. The main results of these experiments will be reported. Different materials such as mild steel, ferritic steel and aluminium have been

  2. Improvements in a prototype boiling water reactor: Laguna Verde, Mexico

    International Nuclear Information System (INIS)

    Laguna Verde is the first nuclear power station in Mexico. It has two GE Boiling Water Reactors which will produce 654 MWe each via Mitsubishi turbine generators. At this moment we are ready to load fuel on Unit 1 and 50% complete on Unit 2 beginning electromechanical installation. The project has required 3,600 million dollars including interest rate, over 1,100 full time engineers and about 3,800 direct labour workers and additionally QA, engineering, construction, start-up and operations prepared using approximately 4,400 procedures to perform their activities. Furthermore, 54 industry branches in Mexico have been qualified by quality assurance and they have been providing equipment, components and sub components for the project. Constructing Unit 2 has given us the opportunity to realize the benefits of standardization. Once ''people'' become familiar with a design concept, a BWR-5 with a Mark II containment in this case, the engineering, construction and testing process improves drastically. As of this date, the average savings in man-hours required to build Unit 2 is 40.59% versus the amount needed for Unit 1. We are not making any dramatic change in the design concept of Unit 1, what we are changing in Unit 2 are our working methods and improving when it is appropriate. For instance, large bore piping, HVAC ducts and cable trays are remaining as they are in Unit 1; however, small bore piping, conduit and tubing will be routed in a different manner to reduce as much as possible the number of supports. Supports in Unit 2 will be multidisciplinary since many interferences in Unit 1 were due to an excessive number of supports which were installed on a per discipline basis. We have not achieved that point yet, but in general in control systems, instrumentation and computers there is plenty of room for improvements, by using fiber optics, multiplexers, etc. We will certainly try it. The message is, a developing country does not have the luxury of changing its

  3. A stochastic study of noise in boiling reactors

    International Nuclear Information System (INIS)

    A stochastic point model is considered of a boiling reactor, involving the population of neutrons, the population of delayed neutron precursors, fuel temperature, and the number of bubbles in the coolant as random variables. Whereas the first two variables are related to capture, fission and delayed neutron processes, the other two take into account heat transfer between fuel and coolant and changes in coolant density. The variations in the fuel temperature and the coolant density generate reactivity feedbacks which affects the neutron power spectral density; analysis of the shape of this spectral density is expected to give information on the value of reactor parameters such as, for example, the void coefficient. (U.K.)

  4. Operational margin monitoring system for boiling water reactor power plants

    International Nuclear Information System (INIS)

    This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

  5. Improvements in boiling water reactor designs and safety

    International Nuclear Information System (INIS)

    The advanced boiling water reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are discussed in this paper. They include: design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 last stage buckets; and advanced radwaste technology

  6. Boils

    Science.gov (United States)

    ... or recurrent boils, which are usually due to Staph infections. The bacteria are picked up somewhere and then ... version of boils is folliculitis . This is an infection of hair follicles, usually with Staph bacteria. These often itch more than hurt. The ...

  7. The benefits of international cooperation via the Boiling Water Reactor Owners' Group (BWROG)

    International Nuclear Information System (INIS)

    The Boiling Water Reactors Owners' Group (BWROG) is an industry organization that was created in an effort to support common resolution of technical issues, address regulatory concerns, promote sharing of information and lessons learned among members, as well as to promote safety, minimize cost, and provide the proper forum for its members to address various specific issues. The BWROG is set up with an Executive Committee, responsible for overall organization performance, a General Committee responsible for day to day issues and operations, as well as numerous Technical Committees. BWROG represents almost 70 reactors worldwide and thousands years of operating experience

  8. Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

    International Nuclear Information System (INIS)

    The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was 'filtered', meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network

  9. Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, B.G. [GSM Power Systems AB, Nykoeping (Sweden)

    1998-05-01

    The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was `filtered`, meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network 11 refs, 46 figs

  10. Passive gamma analysis of the boiling-water-reactor assemblies

    Science.gov (United States)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) 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. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  11. Modelling of void formation in the subcooled boiling regime in the athlet code to simulate flow instability for research reactors

    International Nuclear Information System (INIS)

    The ATHLET thermohydraulic code was developed at the Gesellschaft fuer Anlagen und Reaktorsicherheit (Grs - Society for Plant and Reactor Safety) to analyse leaks and transients for power reactors. In order to extend the code's range of application to the safety analysis of research reactors, a model was implemented permitting a description of the thermodynamic nonequilibrium effects in the subcooled boiling regime. The aim of the extension is, on the one hand, to cover the thermohydraulic instability which is particularly characteristic of research reactors due to their high power densities and low system pressures. and, on the other hand, to provide a consideration of the influence of the steam formed in this boiling regime on the neutron balance. It describes the bubble production rate at the superheated heating surfaces as well as the subsequent condensation of the bubbles in the subcooled core flow. The installed model is validated by the recalculation of two extensive series of experiments. In the first series, the McMaster experiments on axial void distribution in the subcooled boiling regime are recalculated. The recalculation shows that the extended program is capable of calculating the axial void distribution in the subcooled boiling regime with good agreement with the data. The second series deals with KFA experiments on thermohydraulic instability (Flow Excursion) in the subcooled boiling regime comprising a broad parameter range of heat flow density, inlet temperature and channel width. Recalculation of this experimental series shows that the program extension ensures simulation of thermohydraulic instability (Author)

  12. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    International Nuclear Information System (INIS)

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are 2-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown. (orig.)

  13. Simulation of boiling flow experiments close to CHF with the NEPTUNE-CFD code

    International Nuclear Information System (INIS)

    A three-dimensional two-fluid code NEPTUNECFD has been validated against the ASU (Arizona State University) [1] and DEBORA [2, 3] boiling flow experiments. Nucleate boiling processes in the subcooled flow boiling regime have been studied on ASU experiments. Within this scope a new wall function is implemented in the NEPTUNECFD V1.0.6 code to improve the prediction of flow parameters in the boiling boundary layer. The capability of the code to predict boiling flow regime close to critical heat flux (CHF) conditions has been assessed on selected DEBORA experiments. It was shown that the code is able to predict wall temperature excursion and a sharp void fraction increase near the heated wall, which are characteristic phenomena for CHF conditions. (author)

  14. Neutronic challenges of advanced boiling water reactor designs

    International Nuclear Information System (INIS)

    The advancement of Boiling Water Reactor technology has been under investigation at the Center for Advance Nuclear Energy Systems at MIT. The advanced concepts under study provide economic incentives through enabling further power uprates (i.e. increasing vessel power density) or better fuel cycle uranium utilization. The challenges in modeling of three advanced concepts with focus on neutronics are presented. First, the Helical Cruciform Fuel rod has been used in some Russian reactors, and studied at MIT for uprating the power in LWRs through increased heat transfer area per unit core volume. The HCF design requires high fidelity 3D tools to assess its reactor physics behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher power density over existing BWRs, while using current licensing tools and existing technology. Its larger assembly size requires stronger coupling between neutronics and thermal hydraulics compared to the current practice. Third is the reduced moderation BWRs, which had been proposed in Japan to enable breeding and burning of fuel as an alternative to sodium fast reactors. Such technology suffers from stronger sensitivity of its neutronics to the void fraction than the traditional BWRs, thus requiring exact modeling of the core conditions such as bypass voiding, to correctly characterize its performance. (author)

  15. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor... modifying or adding EP requirements in Section 50.47, Section 50.54, and Appendix E of 10 CFR part 50 (76...

  16. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has been... revised 10 CFR 73.55 through the issuance of a final rule on March 27, 2009 (74 FR 13926). Section...

  17. NO2 Catalytic Decomposition - from Laboratory Experiment to Industry Reactor

    Czech Academy of Sciences Publication Activity Database

    Obalová, L.; Jirátová, Květa; Kovanda, F.

    Kraków : Wydawnictwo Uniwersitetu Jagiellonskiego, 2011, s. 97. ISBN 978-83-233-3249-7. [International Symposium on Nitrogen Oxides Emission Abatement NOEA 2011. Zakopane (PL), 04.09.2011-07.09.2011] R&D Projects: GA TA ČR TA01020336 Institutional research plan: CEZ:AV0Z40720504 Keywords : catalytic decomposition of CO2 * reactor * kinetic data Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  18. Future Reactor Experiments

    CERN Document Server

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measurement techniques have been explored. A proposed experiment JUNO, with a 20 kton liquid scintillator detector of $3%/$$\\sqrt{E(MeV)}$ energy resolution, $\\sim$ 53 km far from reactors of $\\sim$ 36 GW total thermal power, can reach to a sensitivity of $\\Delta\\chi^{2}>16$ considering the spread of reactor cores and uncertainties of the detector response. Three of mixing parameters are expected to be measured to better than 1% precision. There are multiple detector options for JUNO under investigation. The technical challenges...

  19. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling

  20. Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design

    International Nuclear Information System (INIS)

    Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation

  1. Investigation on the applicability of Piety's on-line PSD-pattern recognition algorithm to boiling detection by neutron-noise at a swimming-pool reactor

    International Nuclear Information System (INIS)

    The neutron noise signal of an initiation-of-boiling experiment performed at the SAPHIR reactor has been analyzed by the PSD-pattern recognition algorithm of Piety (1977); the results indicate that the onset of boiling can be detected by this method. Improved confidence statements for the statistical decision discriminants are given. (Auth.)

  2. Liquid-cooled nuclear reactor, especially a boiling water reactor

    International Nuclear Information System (INIS)

    A nuclear reactor with a special arrangement of fuel rods in the core is designed. Each fuel element has its shaft which is made of sheets, has the same cross section as the fuel element and protrudes at least the length of the control rod above the reactor core. Made of a zirconium alloy in the core area and of stainless steel above it, the shaft is equipped with channels for sliding the rods in and out and serves to spatially secure the position of the rods. Coolant flow is provided by the chimney effect. The shaft can conveniently enclose the control rod drive. It can also serve to bear the water separator. Moreover, it can constitute a part of the casing which surrounds the fuel rods and keeps the fuel in an intimate contact with the coolant; the other part of this casing is constituted by inserted sheets which can conveniently have the shape of angles. The walls of neighboring shafts form a compartment accommodating a neutron absorber plate. (M.D.). 11 figs

  3. Dynamic simulation of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    For the application of modern control theory, specifically optimal control, to the boiling water reactor, it is necessary to have a linear model that is validated. The nonlinear model of the BWR derived on the basis of physical laws and empirical relations is linearized around an operating point and the model if verified against experimental results by simulating various tests such as the pressure transient test, change in power to recirculating pump etc. The transport delay occurring in the model is approximated by various representations and the results are compared with the exact delay representation. Validation such as discussed in the paper forms the basis for devising appropriate control strategies in the presence of disturbances. (author)

  4. Analytical simulation of boiling water reactor pressure suppression pool swell

    International Nuclear Information System (INIS)

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement

  5. Analytical simulation of boiling water reactor pressure suppression pool swell

    Energy Technology Data Exchange (ETDEWEB)

    Widener, S.K.

    1986-01-01

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement.

  6. Multi-dimensional nodal analysis of boiling water reactor stability

    International Nuclear Information System (INIS)

    A computer program, NUFREQ-3D, was developed for boiling water reactor stability analysis. The code, which incorporates sophisticated thermal-hydraulic model coupled with a space dependent nodal neutronic model, is able to evaluate the system stabilities in terms of state variables such as inlet flow rate, power density, and system pressure. The detailed full 3-D representation was developed for more accurate stability analysis by using the sparse matrix techniques and by a channel grouping procedure. Results of modeling a representative operating BWR system show that spatial coupling has a significant effect on the prediction of stability margins. Comparisons of calculated transfer functions with the measured data also reveal that the code generally predict well the trends of system transfer functions

  7. Boiling water reactor off-gas systems evaluation

    International Nuclear Information System (INIS)

    An evaluation of the off-gas systems for all 25 operating Boiling Water Reactors (BWR) was made to determine the adequacy of their design and operating procedures to reduce the probability of off-gas detonations. The results of the evaluations are that, of the 25 operable units, 13 meet all the acceptance criteria. The other 12 units do not have the features needed to meet the criteria, but have been judged to have, or are committed to provide, features which give reasonable assurance that the potential for external off-gas detonations is minimized. The 12 units which did not originally meet the criteria are aware of the potential hazards associated with off-gas detonations and have agreed to take action to minimize the probability of future detonations

  8. CFD simulation of DEBORA boiling experiments

    Science.gov (United States)

    Rzehak, Roland; Krepper, Eckhard

    2012-08-01

    In this work we investigate the present capabilities of computational fluid dynamics for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. This kind of modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant non-dimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12) as the working fluid. This facilitated measurements of radial profiles for gas volume fraction, gas velocity, bubble size and liquid temperature as well as axial profiles of wall temperature. After reviewing the theoretical and experimental basis of correlations used in the ANSYS CFX model used for the calculations, we give a careful assessment of the necessary recalibrations to describe the DEBORA tests. The basic CFX model is validated by a detailed comparison to the experimental data for two selected test cases. Simulations with a single set of calibrated parameters are found to give reasonable quantitative agreement with the data for several tests within a certain range of conditions and reproduce the observed tendencies correctly. Several model refinements are then presented each of which is designed to improve one of the remaining deviations between simulation and measurements. Specifically we consider a homogeneous MUSIG model for the bubble size, modified bubble forces, a wall function for turbulent boiling flow and a partial slip boundary condition for the liquid phase. Finally, needs for further model developments are identified and promising directions discussed.

  9. Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal. 2. The second experiment

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Shinohara, Yoshikuni; Inoue, Kimihiko

    1995-01-01

    In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the second reactor noise experiment using pseudo random binary sequences (PRBS) was performed on August 30, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 50% of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author).

  10. Expected reactivity effect of fuel channel coolant boiling in the Darlington NGS A reactor core

    International Nuclear Information System (INIS)

    We have developed a formalism for estimating the expected reactivity due to channel boiling in any reactor designed to have some quality in the channel. In applying this formalism to the Darlington NGS A equilibrium core, we calculate a value of 0.024 ± 0.003 mk at 100% power operation. In Darlington, the channel feeders are individually sized so that the coolant in each channel has some boiling on reaching the entrance to the reactor outlet header. (Hereafter called the 'ROH quality'). The design is such that when each channel is at its nominal time-averaged 100 percent power, the quality at the ROH should be just under 2%. The day-to-day variation of each channel's power around its time-averaged value (i.e., 'ripple') results in a corresponding variation in the quality and consequently in the reactivity due to boiling. Traditionally, fuel management codes such as SORO, FMDP, RFSP and OHRFSP use fuel properties generated by a lattice code such as POWDERPUFS or LATREP. These fuel properties are functions of fuel irradiation only, with all other core-varying input parameters to the lattice code held constant at core-averaged values. Recently, some work has gone into developing a Pt. Lepreau version of RFSP in which the fuel properties are functions of fuel temperature and coolant density as well as of fuel irradiation. This paper reports the results of a study which was undertaken to quantify the expected variation in core reactivity due to this day-to-day variation in channel power and channel boiling. It could then be determined whether the reactivity effect of this boiling is sufficient to justify the explicit representation of the fuel properties as a function of coolant density

  11. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au)

  12. BWR [boiling water reactor] shutdown margin model in SIMULATE-3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) technical specifications require that the reactor be kept subcritical (by some prescribed margin) when at room temperature rodded conditions with any one control rod fully withdrawn. The design of an acceptable core loading pattern may require hundreds or thousands of neutronic calculations in order to predict the shutdown margin for each control rod. Direct, full-core, three-dimensional calculations with the SIMULATE-3 two-group advanced nodal code require 3 to 6 CPU min (on a SUN-4 workstation) for each statepoint/control rod that is computed. Such computing and manpower requirements may be burdensome, particularly during the early core design process. These requirements have been significantly reduced by the development of a fast, accurate shutdown margin model in SIMULATE-3. The SIMULATE-3 shutdown margin model achieves a high degree of accuracy and speed without using axial collapsing approximations inherent in many models. The mean difference between SIMULATE-3 one-group and two-group calculations is approximately - 12 pcm with a standard deviation of 35 pcm. The SIMULATE-3 shutdown margin model requires a factor of ∼15 less CPU time than is required for stacked independent two-group SIMULATE-3 calculations

  13. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  14. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

  15. Experience gained in the erection of the Philippsburg-2 reactor

    International Nuclear Information System (INIS)

    The article is a comprehensive summary with tables and figures of contract award, chronology of the various licences and construction permits, the licencing procedure under atomic law, consequences of technological advances made during construction, and measures securing compliance with time limits. The chronology of projects showed intervals of up to 3 years, most of which were caused by the courts putting a freeze on the plant. Owing to different expertises, there were many differences of opinion that needed settling. The present state of affairs and the situation concerning the licencing procedure has now come to a point where the course of the project can be expected to be on schedule. After the first criticality in October 1984 and performance tests until spring 1985, the Philippsburg-2 reactor will be an essential source of energy supply. (orig.)

  16. Radiation levels at boiling water reactors of a commercial nuclear power plant fleet

    International Nuclear Information System (INIS)

    Radiation field control at a Boiling Water Reactor (BWR) is a complex process that requires the application of both theoretical knowledge and practical experience in order to achieve low radiation fields. Older BWRs were usually designed with cobalt containing components, such as Stellite™ materials in valves, control rod blades, turbine blades and others, that contribute to high radiation fields due to the activation of cobalt to Co-60. Newer BWRs are designed with improvements in these areas; however, only the newest BWRs have been designed using low cobalt source term materials for all components in streams that enter the reactor. Control and minimization of the cobalt source term (material that can be activated to Co-60 in the reactor) will ensure that as low as reasonably achievable (ALARA) dose rates are achieved during power operation and during refueling outages. (author)

  17. Experience with Kamini reactor

    International Nuclear Information System (INIS)

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  18. Modelling of void formation in the subcooled boiling regime in the ATHLET code to simulate flow instability for research reactors

    International Nuclear Information System (INIS)

    The ATHLET thermohydraulic code was developed at the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Society for Plant and Reactor Safety) to analyse leaks and transients for power reactors. In order to extend the code's range of application to the safety analysis of research reactors, a model was implemented permitting a description of the thermodynamic non-equilibrium effects in the subcooled boiling regime. The aim of the extension is, on one hand, to cover the thermohydraulic instability which is particularly characteristic of research reactors owing to their high power densities and low system pressures and, on the other hand, to provide a consideration of the influence of the steam formed in this boiling regime on the neutron balance. The model developed takes into consideration the competing evaporation and condensation effects in the subcooled boiling regime. It describes the bubble production rate at the superheated heating surfaces as well as the subsequent condensation of the bubbles in the subcooled core flow. The installed model is validated by the recalculation of two extensive series of experiments. In the first series the McMaster experiments on axial void distribution in the subcooled boiling regime are recalculated. The recalculation shows that the extended programme is capable of calculating the axial void distribution in the subcooled boiling regime with good agreement with the data. The second series deals with KFA experiments on thermohydraulic instability (flow excursion) in the subcooled boiling regime, comprising a broad parameter range of heat flow density, inlet temperature and channel width. Recalculation of this experimental series shows that the programme extension ensures simulation of thermohydraulic instability. (orig.)

  19. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  20. High conversion pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–233U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–233U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm3, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore, some means of

  1. Potential effects of ex-vessel molten core debris interactions on boiling water reactor containment integrity

    International Nuclear Information System (INIS)

    There is a steadily increasing awareness of the highly plant-specific nature of reactor safety issues. This awareness is reflected in the increasing number of research programs focused on problems limited to specific reactor or containment types. This report is limited to NRC-sponsored research on accident phenomena that may affect the integrity of boiling water reactor containment systems arising out of ex-vessel interactions of molten core debris in the reactor cavity. Some safety issues that are generic to all types of BWRs are discussed, these include: (1) effects of concrete composition, (2) dispersive effect of structures below the reactor vessel, (3) influence of unoxidized zirconium metal in the debris pool, (4) the influence of water in the reactor cavity on debris coolability and magnitude of the radiological source term, and (5) the nature of high-temperature condensed-phase chemistry and fission-product aerosol generation. Certain ex-vessel core-debris phenomena which may threaten the integrity of specific BWR containment designs include the following: (1) integrity of the BWR MARK-I steel pressure boundary, (2) potential for penetration of the MARK-II drywell floor and/or supression-pool bypass, and (3) possible failure of the MARK-III reactor support system due to thermal ablation of the reactor pedestal. Some recent experimental results derived from NRC-sponsored programs are also presented

  2. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  3. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  4. Experiments with the initial erections of the Angra-2 reactor

    International Nuclear Information System (INIS)

    The main erection services for Angra 2 will be contracted 11/95; the actual services will start in 02/96. FURNAS and NUCLEN decided to anticipate some services related to the main erection; the present paper presents the situation of these initial erection services, as well as the experience gained so far. (author)

  5. Nucleate boiling pressure drop in an annulus: Book 2

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ''power tilt'' or asymmetric heating of the inner and outer annulus walls. The test facility uses H2O rather than D2O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux

  6. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  7. Sub-channel analysis by RELAP5 system code of boil-off experiment (Test 5002) with NEPTUN facility

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, A. [Pennsylvania State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, Pennsylvania (United States)]. E-mail: axp46@psu.edu; Bousbia Salah, A.; D' Auria, F. [Univ. of Pisa, Dipartimento di Ingegneria Meccanica, Nucleare d della Produzione, Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; f.dauria@ing.unipi.it

    2004-07-01

    This paper presents the results of RELAP5/Mod3.2 system thermalhydraulic code using the sub-channel analysis approach in predicting the NEPTUN separate effect boil off experiments. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the NEPTUN low pressure test N{sup o}5002 has been considered. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory and demonstrate, as well, the reasonable success of the 'sub-channel analysis' approach adopted in the present context for a system thermalhydraulic code. (author)

  8. Multi-cycle boiling water reactor fuel cycle optimization

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, K.; Maldonado, G.I. [University of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  9. Status of sodium boiling noise detection programme at reactor research centre, India

    International Nuclear Information System (INIS)

    Acoustic detection of sodium boiling is a promising technique to monitor subassembly fault in a last reactor. This paper summarises the programme for developing this detection system and describes the design of a high temperature transducer for boiling detection. It is appreciated that the background noise from primary pumps can interfere with this detection. Noise measurements were therefore carried out during water testing of the primary pump of the Fast Breeder Test Reactor. Some preliminary results of these measurements are presented

  10. Statics and dynamics of a natural circulation cooled boiling water reactor. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Stekelenburg, A.J.C.

    1994-02-21

    Due to the strong interconnection of the various processes in the reactor vessel of a natural circulation cooled boiling water reactor (BWR), explaining the physics of both the statics and the dynamics of the Dodewaard reactor is not an easy task. In this thesis, the physics is studied through a combined experimental and theoretical investigation. The experiments are analyzed further with the use of the model, and the results of the model calculations provide ideas for new experiments. For an experimental study of the reactor behavior, measurement tools are required. Many relevant process variables are supplied by the power plant's data-logger, but a direct method for measuring the circulation flow rate is not available. Reactor behavior can be studied theoreticallly with the use of a complex computer code, based on a multi-node model. In this way, reliable results are obtained. In many cases, however, such a code is not easy to use, and the calculations require much computer time. Calculations based on a simple model have a lower reliability, but, as the model is clearer, provide more insight into the physics of the system. For this reason, a simple theoretical dynamical model for the main physical processes of the Dodewaard natural circulation cooled BWR is presented in the thesis.

  11. Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor

    Science.gov (United States)

    Cheng, Ling

    The thesis evaluates the safety systems in the next generation boiling water reactor by analyzing the main steam line break loss of coolant accident performed in the Purdue university multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform the safety analysis for SBWR. The comparison also validates the three-level scaling methodology applied to the design of the PUMA facility. The PUMA suppression pool mixing and condensation test data have been studied to give the detailed understanding on this important local phenomenon. A simple one dimensional integral model, which can reasonably simulate the mixing process inside suppression pool have been developed and the comparison between the model prediction and the experiment data demonstrates the model can be utilized for analyzing the suppression pool mixing process.

  12. Experiments on microgravity boiling heat transfer by using transparent heaters

    Energy Technology Data Exchange (ETDEWEB)

    Ohta, H. [Kyushu Univ., Fukuoka (Japan). Dept. of Energy and Mech. Eng.

    1997-11-01

    To clarify the relation between the liquid-vapor behavior and the heat transfer characteristics in the boiling phenomena, the structures of transparent heaters were developed for both flow boiling experiments and were applied to the microgravity environment realized by the parabolic flight of aircraft. In the flow boiling experiment, a transparent heated tube makes the heating, the observation of liquid-vapor behavior and the measurement of heat transfer data simultaneously possible. The heat transfer coefficient in the annular flow regime at moderate quality has distinct dependence on gravity provided that the mass velocity is not so high, while no noticeable gravity effect is seen at high quality and in the bubbly flow regime. The measured gravity effect was directly related to the behavior of annular liquid film observed through the transparent tube wall. In the pool boiling experiment, a structure of transparent heating surface realizes both the observation of the macrolayer or microlayer behavior from underneath and the measurements of local surface temperatures and the layer thickness. It was clarified in the microgravity experiments that no vapor stem exists but tiny bubbles are observed in the macrolayer underneath a large coalesced bubble at high heat flux. The heat flux evaluated by the heat conduction across the layer assumes less than 30% of the total to be transferred. The evaporation of the microlayers underneath primary bubbles just after the generation dominates the heat transfer in the microgravity, not only in the isolated bubble region but also in the coalesced bubble region. (orig.) 14 refs.

  13. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  14. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    International Nuclear Information System (INIS)

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

  15. Void Reactivity Effects in the Second Charge of the Halden Boiling Water Reactor

    International Nuclear Information System (INIS)

    The reactivity effect of voids caused by boiling inside the coolant channels in the second fuel charge of the Halden Boiling Heavy Water Reactor has been measured both in void-simulated zero-power experiments and under actual power conditions. The void-simulated experiments consisted of measuring the reactivity effect of introducing void columns inside thin-walled tubes to various depths. The tubes were placed at different positions between die stringers in a single 7-rod cluster element practically identical with the normal second-charge fuel elements. This experiment enables an investigation of the reactivity dependence upon void fraction, and also the reactivity dependence of steam-bubble position in the coolant channel. The experiment was carried out in the Norwegian zero-power facility NORA, with a core consisting of 36 second-charge elements and with a lattice geometry identical to the one in HBWR. The temperature dependence of the void effect was investigated in a zero-power experiment with the 100 fuel-element core of HBWR. In a single fuel element the water level inside the coolant channel was depressed to various depths, and the reactivity effect of this perturbation was measured at different temperatures in the temperature interval 50°C-220°C. The power void reactivity has been measured in HBWR as a function of nuclear power at different moderator temperatures between 150°C and 230°C at powers up to about 16 MW at the highest temperature. The power-void reactivity coefficient is an important quantity in determining the dynamic behaviour of a boiling- water reactor. The theoretical determination of this quantity is, however, complicated by the fact that knowledge about the void distribution in the core is required. The detailed power-void distribution is not easily amenable to experimental determination, and accordingly the void-simulated experiments represent a better case for testing the reactor physics calculation of void effects. Preliminary

  16. Multivariable autoregressive model of the dynamics of a boiling water reactor

    International Nuclear Information System (INIS)

    An autoregressive (AR) model with pseudo-random binary sequence (PRBS) test signals was applied to the dynamics of the Japan Power Demonstration Reactor, a boiling water reactor (BWR). The decision of the order of the AR model was based on the Akaike criterion. Multi-input test signals of the PRBS were applied to the steam-flow control valve and the forced circulation pump speed control terminal. Seventeen variables including the instrumented fuel assemblies were observed. The AR model identification facilitated building the BWR dynamics model as a multivariable system. The experiment indicated that the BWR dynamics with rather intensive nonwhite noise interference was effectively represented by the AR model, which was compared with a linear theoretical dynamics model. The results suggested that the identified AR model plays an important role in verifying, modifying, and improving the theeoretical dynamics model

  17. Experiments on HFE-7100 pool boiling at atmospheric pressure in horizontal narrow spaces

    Energy Technology Data Exchange (ETDEWEB)

    Guglielmini, G.; Misale, M.; Priarone, A. [Universita degli Studi di Genova (Italy). DIPTEM - Sezione di Termoenergetica e Condizionamento Ambientale

    2009-07-01

    Experiments were performed to examine the pool boiling heat transfer and critical heat flux on a smooth copper circular surface, confined by a face-to-face parallel unheated surface, by changing the gap between the surfaces and the unheated surface diameter. Pool boiling data at atmospheric pressure were obtained for saturated HFE-7100. The gap values investigated, between the boiling surface and the adiabatic one, were s 0.5, 1.0, 2.0, 3.5 mm. To confine the boiling surface, two different Plexiglas plates were used: the former characterised by a diameter D = 60 mm, large as the overall test section support, the latter characterised by a diameter D = 30 mm, large to cover only copper boiling surface (d = 30 mm). For each configuration, boiling curves were obtained up to the thermal crisis. For both different types of confinement, it was observed that the boiling curves match at low wall superheat, except for s = 0.5 mm, 1 mm. However, at high wall superheat, a drastic reduction in heat transfer as well as CHF appears decreasing the channel width s; for all gap sizes, this reduction is less pronounced for the smaller confinement wall (D = 30 mm). Instead, at low wall superheat for gap of 0.5 and 1.0 mm, the heat transfer coefficient is higher for diameter disc of 60 mm. CHF data were also compared with a literature correlation (Misale and al., 2009). (author)

  18. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  19. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    International Nuclear Information System (INIS)

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied

  20. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  1. State-of-the-art and Prospects for Development of Innovative Simplified Boiling Water Reactor VK-300

    International Nuclear Information System (INIS)

    At present RDIPE is close to completion of the detailed design of innovative simplified boiling water reactor VK-300. A nuclear power plant equipped with VK-300 reactor facility is intended for small- and medium-size power systems as well as for electricity and heat co-generation. The innovative reactor facility VK-300 has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype reactor VK-50 in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR (Siemens). Thus, the reactor pressure vessel, fuel elements and moisture separators developed for WWER-1000 reactor facility were taken for VK-300. The presentation will be focused on the most important design features of VK-300. More attention will be given to the specific features of the reactor design related to: original and efficient scheme of coolant circulation and separation, top placement of CPS drive mechanisms, unique system for reactor core emergency cooling. Reactor passive safety features will be given a special emphasis. The economical indices of a power unit with VK-300 reactor will be presented and an analysis will be done to illustrate how a small to medium power reactor can be economically competitive with large sized plants. The prospects for developing in Russia the nuclear power units with VK-300 reactor facility will be analyzed. (author)

  2. Boiling water reactor stability analysis by stochastic transfer function identification

    International Nuclear Information System (INIS)

    The univariate and the bivariate ARMA models are proposed as the stochastic transfer function models for the identification of BWR systems. This technique has been developed as a new method for on-line system identification, optimum control, and malfunction monitoring of nuclear power plants. The relationships between the stochastic transfer function model and the differential equation model are derived. The estimation algorithms are developed through the related covariance functions and Green's function by the least squares method. It has been shown that the stochastic models can also be used for fitting the stochastic data which are contaminated with sinusoidal waves. Both the univariate and the bivariate modeling are applied in the BWR system identification and stability analysis. The univariate modeling is applied to decompose the pressure dynamics from the neutron data. From both of the normal operation data and the perturbation experiment data, the reactor dynamics are consistently estimated. The dynamics of the reactor core are estimated as a second order mode with a natural frequency of 0.4 Hz and a damping ratio of 0.1. The univariate modeling is also applied to monitor the local performance of the coolant channel in the reactor. The transfer functions between system's variables are obtained by use of bivariate modeling. The obtained transfer functions are closely related to the stability analysis of thermal-hydraulics in the reactor. The transition of the system dynamics from normal operation to the perturbation experiment are observed

  3. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-04-30

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

  4. Modernized IBR-2 reactor and first experiments at its neutron beams

    International Nuclear Information System (INIS)

    IBR-2 reactor is the main basic facility at JINR dedicated to condensed matter research. The IBR-2 operates as a fast pulsed reactor. Its main distinctive property, which makes it differ from other nuclear reactors, is the mechanical modulation of the reactivity by means of a movable reflector. Producing a record neutron flux of 1016 n/cm2/s in the pulse, the IBR-2 reactor is also an economical and relatively inexpensive facility. The IBR-2 reactor is mainly used for investigations in the fields of condensed matter physics (solids and liquids), biology, chemistry, earth and materials science. Operating experience has shown that it is a very effective neutron source; in most areas of application it compares well with the best neutron sources based on proton accelerators. At present, this experience is of special importance in connection with the increasing interest in long-pulsed neutron sources. IBR-2 operated successfully from 1984 until 2006. On December 18, 2006 reactor was shut down for modernization. Main directions of reactor modernization include: 1. A compact reactor core. 2. Lower speed of rotation of the main movable reflector, counter rotation of rotors, use of a nickel alloy as a reflector material. 3. Use fuel pellets configuration that will allow increasing the depth of fuel burn up to 9%. 4. New design of safety system which improves its parameters. 5. Creation of easily replaceable moderators, their optimization for each neutron beam. Development of the cryogenic moderators with palletized moderator material. Dismantling of the old reactor parts and installation of the new equipment was completed in 2010. Brief history of the work will be outlined in the report. In February 2011 loading of fresh fuel to the reactor core was completed and physical start-up has begun. After successful realization of this stage in June, the power start-up program was fulfilled resulting in increase of the mean reactor power to the design value of 2 MW (peak power of

  5. Two compartment water rod for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matzner, B.; Wolters, B.A.

    1993-07-27

    In a fuel bundle for boiling water nuclear reactor, said fuel bundle is described including a matrix of upstanding fuel rods for undergoing nuclear reaction and generating steam, a lower tie plate for supporting the matrix of fuel rods and admitting liquid water moderator to the fuel bundle from the lower portion of said fuel bundle, an upper tie plate for fastening to at least some of the fuel rods and permitting the outflow of liquid and vapor water moderator from the upper portion of said fuel bundle, a channel surrounding said upper and lower tie plates and said fuel rods therebetween for confining moderator flow between said tie plates and around said fuel rods, and a plurality of vertically spaced apart fuel rod spacers, each said spacer surrounding each said fuel rod at the particular elevation of said spacer for maintaining said fuel rods in side-by-side relation, and a water rod for installation to said fuel bundle for supplying liquid moderator to the upper two phase region of said fuel bundle, the improvement to said water rod comprising: said water rod having a first upper compartment, and a second lower compartment, said upper compartment isolated from said lower compartment; said first upper compartment defining an open, upwardly exposed end for receiving and maintaining water in said upper water rod compartment during the power generating operation of said fuel bundle will fill with liquid by gravity flow from above; means communicated to the bottom portion of said lower compartment for receiving water from said lower portion of said fuel bundle; and, means communicated to the upper portion of said lower compartment for discharging water to the interior of said fuel bundle below the upper most spacer of said fuel bundle whereby discharge to said fuel bundle occurs in said upper two phase region of said bundle.

  6. Two compartment water rod for boiling water reactors

    International Nuclear Information System (INIS)

    In a fuel bundle for boiling water nuclear reactor, said fuel bundle is described including a matrix of upstanding fuel rods for undergoing nuclear reaction and generating steam, a lower tie plate for supporting the matrix of fuel rods and admitting liquid water moderator to the fuel bundle from the lower portion of said fuel bundle, an upper tie plate for fastening to at least some of the fuel rods and permitting the outflow of liquid and vapor water moderator from the upper portion of said fuel bundle, a channel surrounding said upper and lower tie plates and said fuel rods therebetween for confining moderator flow between said tie plates and around said fuel rods, and a plurality of vertically spaced apart fuel rod spacers, each said spacer surrounding each said fuel rod at the particular elevation of said spacer for maintaining said fuel rods in side-by-side relation, and a water rod for installation to said fuel bundle for supplying liquid moderator to the upper two phase region of said fuel bundle, the improvement to said water rod comprising: said water rod having a first upper compartment, and a second lower compartment, said upper compartment isolated from said lower compartment; said first upper compartment defining an open, upwardly exposed end for receiving and maintaining water in said upper water rod compartment during the power generating operation of said fuel bundle will fill with liquid by gravity flow from above; means communicated to the bottom portion of said lower compartment for receiving water from said lower portion of said fuel bundle; and, means communicated to the upper portion of said lower compartment for discharging water to the interior of said fuel bundle below the upper most spacer of said fuel bundle whereby discharge to said fuel bundle occurs in said upper two phase region of said bundle

  7. Non linear dynamics of boiling water reactor dynamical system

    International Nuclear Information System (INIS)

    The fifth order phenomenological model of March-Leuba for boiling water reactors include the point reactor kinetics equations for neutron balance and effective delayed neutron precursor groups with one node representation of the heat transfer process and channel thermal hydraulics. This nonlinear mathematical model consists five coupled nonlinear ordinary differential equations. The reactivity feedback (void coefficient of reactivity as well as the fuel temperature coefficient of reactivity), heat transfer process and momentum balance are major reasons for the appearance of nonlinearity in this dynamical system. The linear stability of a dynamical system with the existence of nonlinearity cannot predict a true picture of the stability characteristics of dynamical system; hence nonlinear stability analyses become an essential part to predict the global stable region on the stability map. The linear stable region is analyzed by the eigenvalues. In this stable region all the eigenvalues have negative real parts, but when pair of one of the complex eigenvalues passes transversely through imaginary axis, the dynamical system loses or gain its stability via a Hopf bifurcation and limit cycles emerges from the tip. The study of eigenvalues can predict a few bifurcations. The first Lyapunov coefficient and normal form coefficients can be used for the detection of other bifurcations in the systems. Stable or unstable limit cycles excite from these Hopf points. These limits cycles gains or loses their stability via limit point bifurcation of cycles, period doubling bifurcation of cycles and Neimark-Sacker bifurcation of cycles when one of the parameters of the nuclear dynamical system is varied. The stability of these limit cycles can be studied by Floquet theory and Lyapunov coefficient, but the bifurcations of limit cycles can be investigated only by critical Floquet multiplier which is basically the eigenvalue of the monodromy matrices. The cascade of period doubling

  8. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  9. Modeling of two-phase flow in boiling water reactor using phase-weighted ensemble average method

    International Nuclear Information System (INIS)

    Investigations into boiling, the generation of vapor and the prediction of its behavior are important in the stability of boiling water reactors. The present models are limited to simplifications made to draw governing equations or lack of closure framework of the constitutive relations. The commercial codes fall into this category as well. Consequently, researchers cannot simply find the comprehensive updated relations before simplification in order to simplify them for their own works. This study offers a state of the art, phase-weighted, ensemble-averaged, two-phase flow, two-fluid model for the simulation of two-phase flow with heat and mass transfer. This approach is then used for modeling the bulk boiling (thermal-hydraulic modeling) in boiling water reactors. The resultant approach is based on using the energy balance equation to find a relation for quality of vapor at any point. The equations are solved using SIMPLE algorithm in the finite volume method and the results compared with real BWR (PB2 BWR/4 NPP) and the boiling data. Comparison shows that the present model is satisfactorily improved in accuracy.

  10. Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  11. Thermodynamic modeling of the processes in a boiling water reactor to buildup the magnetic corrosion product deposits

    International Nuclear Information System (INIS)

    Highlights: ► Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. ► Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe2O4]. ► GEM calculations applied to the boiling zone match with the EPMA and EXAFS findings. ► Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations. - Abstract: The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by using Gibbs Energy Minimization (GEM-Selector code) calculations of thermodynamic equilibrium at in situ temperatures and pressures. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe2O4] spinel solid solutions. GEM calculations applied to the boiling zone match with the electron probe microanalysis (EPMA) and Extended X-ray Absorption Fine Structure (EXAFS) findings, indicating that zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations under Zn water chemistry conditions. GEM results have helped to explain the existence of magnetic product deposits on the surface of the fuel element and the processes that take place in the reactor.

  12. Hybrid simulation of boiling water reactor dynamics using a university research reactor

    International Nuclear Information System (INIS)

    A ''hybrid'' reactor/simulation (HRS) testing arrangement has been developed and experimentally verified using The Pennsylvania State University (Penn State) TRIGA Reactor. The HRS uses actual plant components to supply key parameters to a digital simulation (and vice versa). To implement the HRS on the Penn State TRIGA reactor, an experimental or secondary control rod drive mechanism is used to introduce reactivity feedback effects that are characteristic of a boiling water reactor (BWR). The simulation portion of the HRS provides a means for introducing reactivity feedback caused by voiding via a reduced order thermal-hydraulic model. With the model bifurcation parameter set to the critical value, the nonlinearity caused by the neutronic-simulated thermal/hydraulic coupling of the hybrid system is evident upon attaining a limit cycle, thereby verifying that these effects are indeed present. The shape and frequency of oscillation (∼ 0.4 Hz) of the limit cycles obtained with the HRS are similar to those observed in operating commercial BWRs. A control or diagnostic system specifically designed to accommodate (or detect) this type of anomaly can be experimentally verified using the research reactor based HRS

  13. Fracture toughness of highly irradiated stainless steels in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Demma, A. [Electric Power Research Inst., Palo Alto, California (United States); Carter, R. [Electric Power Research Inst., Charlotte, North Carolina (United States); Jenssen, A. [Studsvik Nuclear (Sweden); Torimaru, T. [Nippon Nuclear Fuel Development Co. Ltd, Oarai-machi, Ibaraki (Japan); Gamble, R. [Sartrex Corp., Rockville, Maryland (United States)

    2007-07-01

    Austenitic stainless steels in boiling water reactor (BWR) core structures can experience significant fracture toughness reductions at elevated fluence levels. One of the gaps identified by EPRI is the lack of data over the full range of radiation exposure anticipated for BWRs. This paper describes an experimental project started in 2005 to generate additional fracture toughness data of highly irradiated stainless steels at appropriate fluences, in support of a methodology for evaluating the serviceability of internal components in BWRs. The irradiated austenitic stainless steels retrieved from disposed BWR internal components and their irradiation and fabrication histories are described as well as an updated evaluation of the relationship between fracture toughness and neutron fluence for BWR internals. The effect of specimen orientation on fracture toughness is also being investigated. Microstructural and microchemical analyses of the various materials tested are also presented to complement the fracture toughness results. The fracture toughness results indicate: (1) there is a distinct orientation effect on the toughness, (2) there is no apparent variation in JIC with respect to fluence within the test range (from 3.3 to 9.1 10{sup 21} n/cm{sup 2}, E > 1MeV); any variation with fluence is embedded within the testing and material scatter, and (3) the four specimens corresponding to a material irradiated at approximately 5.2 and 5.9 10{sup 21} n/cm{sup 2} have distinctly lower toughness compared to the other tests. The reason for the low toughness of this material is discussed. (author)

  14. State-of-the-art and prospects for development of innovative simplified boiling-water reactor VK-300

    International Nuclear Information System (INIS)

    At present RDIPE is close to the completion of the detail design of an innovative simplified boiling-water reactor VK-300. A nuclear power plant equipped with a VK-300 reactor facility is intended for small- and medium-size power systems as well as for electricity and heat cogeneration. The innovative VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype reactor VK-50 in RIAR, Dimitrovgrad, and the experience of designing such reactors as SBWR (GE) and SWR (Siemens). Thus, the reactor pressure vessel, fuel elements and moisture separators developed for the WWER-1000 reactor facility were taken for VK-300. The presentation will be focused on the most important design features of VK-300. More attention will be given to the specific features of the reactor design relating to a) the original and efficient scheme of coolant circulation and separation, b) the top placement of CPS drive mechanisms, and c) a unique system for reactor core emergency cooling. Reactor passive safety features will be given a special emphasis. The prospects for developing in Russia nuclear power units with VK-300 reactor facility will be analyzed. (authors)

  15. Future reactor experiments

    International Nuclear Information System (INIS)

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper

  16. Future reactor experiments

    Science.gov (United States)

    Wen, Liangjian

    2015-07-01

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  17. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    International Nuclear Information System (INIS)

    The objective of this paper is the simulation and analysis of the Boiling Water Reactor (BWR) lower head during a severe accident. The Couple computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation

  18. Application of quenched and tempered SPV 50 steel to primary containment vessel of boiling water reactor

    International Nuclear Information System (INIS)

    The experiments on the welding of the steel plates for pressure vessels, JIS G 3115, SPV 50, were carried out, and the results were evaluated, in order to use quenched and tempered SPV 50 steel for the MARK-2 type primary containment vessels for boiling water reactors, because the former steel SGV 49 with the thickness limit of 38 mm is not usable to the larger MARK-2 type containment vessels. The chemical composition of the experimental specimens with 40 mm and 70 mm thickness is shown. The quenching temperature is 930 deg. C, and the tempering temperature is 660 deg. C and 630 deg. C for the thickness of 40 mm and 70 mm, respectively. The stress relieving was conducted by the conditions 575 deg. C x 10 h and 575 deg. C x 17 h for the thickness of 40 mm and 70 mm, respectively. The welding conditions are shown, such as welding method, the shape of edge preparation, filler material, preheating temperature, voltage and current, and heat input. The experimental results of tensile test, Charpy test and drop weight test are shown for the parent material and welded joints. The tests of brittle fracture behavior and crack propagation characteristics were conducted, and the results were evaluated. The application of the quenched and tempered SPV 50 steel to the containment vessels was studied by these test results, in comparison with the ASME Code, Sec. 3. This steel was decided to be adopted for the MARK-2 type containment vessels from the viewpoint of the licensing and safety. (Nakai, Y.)

  19. Nuclear power plant with pressure vessel boiling water reactor VK-300 for district heating and electricity supply

    International Nuclear Information System (INIS)

    The viability for Russia of the Boiling Water Reactor (BWR) concept has been shown by a number of feasibility studies fulfilled for perspective sites with increased energy demands. Russia has long (31 year) successful experience in operation of NPPs with the vessel-type boiling reactor VK-50 which is located in the city of Dimitrovgrad. Taking into account the large Russian district heating market, it is expedient to apply this concept (BWR) not only for electricity supply, but also for district heating. This is a way to increase of nuclear power plant competitiveness along with good safety performance. The safety and protection of nuclear heat customer is guaranteed by reliable technical means which are well checked at Russian nuclear sites. (author)

  20. A model for stationary and dynamic impression of undercooled boiling in coupled thermohydraulic and neutron physics calculations of nuclear reactors

    Science.gov (United States)

    Mueller, Roland Guenther

    1987-06-01

    In order to account for subcooled boiling in calculations of neutron physics and thermal hydraulics of light water reactors (where vapor bubbles strongly influence the nuclear chain reaction), a dynamic model is derived from the time-dependent conservation equations. It contains methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. It enables the complete two-phase flow region to be treated consistently. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement is reached. The results from the coupling of the new calculation model with a neutron kinetics program proves its suitability for the steady-state and transient calculation of reactor cores.

  1. Response of the Gamma TIP Detectorsin a Nuclear Boiling Water Reactor

    OpenAIRE

    Fridström, Richard

    2010-01-01

    In order to monitor a nuclear boiling water reactor fixed and movable detectors are used, such as the neutron sensitive LPRM (Local Power Range Monitors) detectors and the gamma sensitive TIP (Traversing Incore Probe) detectors. These provide a mean to verify the predictions obtained from core simulators, which are used for planning and following up the reactor operation. The core simulators calculate e.g. the neutron flux and power distribution in the reactor core. The simulators can also si...

  2. SIMULATE-3K: Enhancements and Application to Boiling Water Reactor Transients

    International Nuclear Information System (INIS)

    The SIMULATE-3K (S-3K) reactor analysis code has been applied to a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) transients since 1993. Over the years, many changes have occurred in the S-3K channel hydraulics and ex-core component modeling. This paper summarizes those changes and outlines the status of existing vessel and steam line models. Examples are given for BWR transients that can be analyzed with S-3K

  3. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    An important goal of nuclear reactor instrumentation is the continuous monitoring of the state of the reactor and the detection of deviations from the normal behaviour at an early stage. Early detection of anomalies enables one to make the necessary steps in order to prevent further damage of nuclear fuel. In the present paper, an on-line core monitoring system is described by means of which boiling anomaly in nuclear reactor fuel assemblies can be detected. (author). 9 refs, 7 figs

  4. TARMS, an on-line boiling water reactor operation management system

    International Nuclear Information System (INIS)

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.)

  5. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    International Nuclear Information System (INIS)

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  6. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  7. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shripad T. Revankar; Seungmin Oh

    2003-09-30

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  8. Severe accident sequence assessment for boiling water reactors: program overview

    International Nuclear Information System (INIS)

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

  9. Experimental and numerical investigation of sub-cooled boiling, condensation, and void flashing in nuclear heating reactor test loop

    International Nuclear Information System (INIS)

    This paper describes experimental and numerical investigations of sub-cooled boiling, condensation, and void flashing in the HRTL-5 test loop, which simulates the primary loop of a 5 MW nuclear heating reactor. A drift-flow model of two-phase with four governing equations was used, in which sub-cooled boiling, condensation, and void flashing have been taken into account. Based on the mathematical model, a program has been developed for analyzing the natural circulation system. As parameters, inlet sub-cooling, system pressure, and heat flux are varied. For comparison, some simplified models, which are designed to reveal the importance of sub-cooled boiling, condensation, flashing in the HRTL-5 test loop, are adopted in the program. The results show: (1) subcooled boiling, condensation, and void flashing may have great influence on the distribution of the void fraction and more intense at low system pressure; (2) the calculation of them is correlative and interactive other than independent; (3) for a system with short heated section, long riser, and low pressure, it is possible to reach 'boiling out of the core', where there is almost no void in the heated section, but much in the riser. (orig.)

  10. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼ 21 Btu/Ibm. A sensitivity study with regard to the steam separator pressure-loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty in the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼ 500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power

  11. Boiling Heat Transfer Experiments by using Transparent Heated Microtube

    Science.gov (United States)

    Huang, Shih-Che; Kawanami, Osamu; Kawakami, Kazunari; Honda, Itsuro; Kawashima, Yousuke; Ohta, Haruhiko

    For detailed study of the relationship between boiling bubble behavior and inner wall temperature during flow boiling in microtubes, a transparent heated microtube, whose inner wall was coated with a thin gold film, was employed. Boiling behavior could be observed clearly, and the inner wall temperature of the tube was measured simultaneously with direct heating of the film. Ionized water was used as a test fluid. The experimental conditions were as follows: tube diameter, 1 mm; inlet liquid subcooling, 10 K; mass velocity, 100 kg/m2s and heat flux, up to 469 kW/m2 in the open system. As a result, the frequencies of fluctuation of the inner wall temperature and flow rate were divided into four regions. In addition, the fluctuation range of flow rate increased with increasing heat flux however, this fluctuation decreased drastically for heat flux over 212 kW/m2. The fluctuation of void fraction coincided with that of inner wall temperature.

  12. Intelligent information data base of flow boiling characteristics in once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Valuable experimental knowledge with flow boiling characteristics of the helical-coil type once-through steam generator was converted into an intelligent information data base program. The program was created as a windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis of any helical-coil type once-through steam generator, (2) analysis and comparison with the experimental data, (3) reference and graph display of the steady state experimental data, (4) reference of the flow instability experimental data and display of the instability threshold correlated by each parameter, (5) summary of the experimental apparatus. (6) menu bar such as a help and print. In the steady state analysis, the region lengths of subcooled boiling, saturated boiling, and super-heating, and the temperature and pressure distributions etc. for secondary water calculated. Steady state analysis results agreed well with the experimental data, with the exception of the pressure drop at high mass velocity. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor with helical-coil type steam generator

  13. Piping reliability analysis for recirculation safe ends of a boiling water reactor

    International Nuclear Information System (INIS)

    This paper presents a piping reliability analysis for the eight recirculation inlet-nozzle safe ends of a boiling water reactor (BWR) nuclear power plant. The analysis is based on principles of probabilistic fracture mechanics. On the basis of observed cracks in the pipe safe ends, the crack is modeled with a semi-elliptical shape initiating at the pipe inner wall. Crack samples are generated using a Monte Carlo simulation technique and an importance sampling scheme. Leak probabilities are estimated through the first ten years of plant lifetime. For the estimated plant operating time of 3 1/2 years, a 20% to 30% probability of safe end leaking is predicted. This prediction correlated well with actual findings at the plant in which one safe end out of eight was leaking after 3 1/2 years. (orig.)

  14. Method for increasing the stability of a boiling water cooled reactor with natural coolant circulation and a boiling water cooled reactor with natural coolant circulation (its versions)

    International Nuclear Information System (INIS)

    The invention is aimed at improving the safety of a boiling water reactor with natural coolant circulation and increasing the reactor core power density by increasing the coolant flowrate and neutron flux stability as well as by reducing the medium compressibility effectiveness in pressure compensator in dynamic modes. The reactor vessel includes the core, draught section, heat exchangers and a pressure compensator. A part of the pressure compensator is separated by a barrier with calibrated openings possessing a limited capacity and hydrolocks. The calibrated openings in the barrier are located below the coolant level and a part of space separated by a barrier is filled with gas from external system. The part of the barrier projecting above the coolant level is adjacent to heat exchangers. In transitional regimes with the change of pressure in the circulation circuit a hydrolock facilitates to reactor vessel projection against repressing and keeps the barrier from excessive power load

  15. Determination of local boiling in light water reactors by correlation of the neutron noise

    International Nuclear Information System (INIS)

    The power limit of swimming-pool type reactors depends on the phenomenon of the appearance of burn-out. In order to determine this limit we have attempted to detect the local boiling which usually occurs before the burn out. Local boiling has been simulated by an electrically heated plate placed in the core of the reactor Siloette. The study of local boiling, which is based on the properties of the correlation functions for the neutron noise of detectors placed in the core, shows that a privileged frequency occurs in the power spectrum of the noise. It is intended in the future to determine the influence of various parameters on this characteristic frequency. (author)

  16. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  17. A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

    Science.gov (United States)

    Sorensen, Kirk; Juhasz, Albert

    2007-01-01

    Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

  18. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  19. Conceptual design of a self-sustainable pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Parametric studies have been performed on a seed-blanket Th-U233 fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts required substantial reduction of the core power density in order to operate under nominal PWR system conditions. Boiling flow regime in the seed area allows better heat removal, which in turn, may potentially allow increasing the power density of the core. In addition, the reduced moderation improves the breeding performance. A 2-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to 104 W/cc, created a map of designs with their corresponding fissile inventory ratio (FIR) values. It was found that several options have the potential to achieve the main objective - a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. (author)

  20. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors

    International Nuclear Information System (INIS)

    The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author)

  1. French Fuel Elements for Natural Uranium-Graphite-CO2 Reactors: Experience Gained and Future Prospects

    International Nuclear Information System (INIS)

    The purpose of this paper is to summarize fuel element development for this reactor type in France in the light of the experience gained with these elements. Progress in the laboratory in work on basic materials (uranium alloys and magnesium alloys) and on the design and assembly of the fuel elements is very briefly recapitulated from the first G2 element studies (1954) to the current work on EDF3 and beyond. The following are then considered: (1) Experience gained in fabricating the G2, G3. EDF1 and EDF2 fuel elements, i.e. (a) the technical problems of mass production, (b) the production cost and the effects on it of the ever-increasing complexity of proposed elements and (c) planning problems, transition from prototype to mass production; (2) In-Pile operating experience with elements actually used in the G2 and G3 reactors, and the experimental irradiations carried out on EDF1 and EDF2. Finally, an attempt is made to determine the cost of the fuel cycle for this reactor type and future developments. (author)

  2. CAD analysis for molten pool experiment and assessment of CIF under subcooled boiling condition

    International Nuclear Information System (INIS)

    In-vessel retention (IVR) of core melt is a key severe accident mitigation measure adopted by some advanced light water reactors. Decay heat removal for IVR is through the external reactor vessel cooling (ERVC) by flooding the reactor cavity during severe accident. As long as the heat flux on the vessel outer surface is lower than the critical heat flux (CHF), the reactor vessel can be sufficiently cooled to maintain the RPV integrity. Heat flux on the RPV outer surface is mainly determined by the core melt behavior, whereas CHF depends on the thermal-hydraulic parameters in the cooling channel outside the RPV. In the present studies, computational fluid dynamics (CFD) models for IVR experiments were carried out, to understand various phenomena related to molten pool and cooling channel. The UCLA experiment and SLUTAN experiment were simulated by using CFD method, the heat source driven natural convection heat transfer coefficient in UCLA experiment, and the two phase boiling phenomena in SLUTAN experiment were investigated in this paper. The CFD calculation results were compared with the experiments data

  3. Pulsation characteristics of boiling water cooled reactor two fuel assembly model

    International Nuclear Information System (INIS)

    The results of experimental studies into the pulsation characteristics of the natural circulation circuit model for the boiling water cooled reactor are given. Influence of nonidentity of fuel assembly power on stability of coolant flow rate was investigated. The methods for avoiding the whole circuit and interassembly hydrodynamic instabilities are suggested

  4. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-02-22

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling... October 14, 2009 (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available...

  5. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-03-09

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling... October 14, 2009, (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available...

  6. A computational study on instrumentation guide tube failure during a severe accident in boiling water reactors

    International Nuclear Information System (INIS)

    This paper focuses on the nature and timing of Instrumentation Guide Tube (IGT) failure in case of severe core melt accident in a Nordic type Boiling Water Reactor (BWR). First, a 2D structural analysis of a RPV lower head is performed to determine global vessel deformation, timing and mode of failure. Next, a structural analysis is also performed on a 3D IGT section taking into account the influence of global vessel deformation and thermo-mechanical load from the melt pool. We have found that the IG tube was not clamped in the housing at the time when welding ring of the IGT nozzle has been melted and global failure of the vessel wall has not started yet. This suggests that IGT failure is the dominant failure mode in the considered case of a large (~200 tons) melt pool. (author)

  7. NABUB a non-saturated model of coolant boiling in a fast reactor sub-assembly

    International Nuclear Information System (INIS)

    A theoretical model is described of sodium boiling in a fast reactor sub-assembly in which the usual assumptions of a saturated vapour are not made. Instead, vapour pressure is calculated in a perfect gas basis, which enables some allowance to be made for the possible presence of non-condensables, which may inhibit the condensation f the vapour. Indications are given of the circumstances under which such inhibition might be expected to show the most marked effects, and some sample results ontained by the code are presented. These show that the coolant voiding pattern is most sensitive to restrictions on the condensing flux in the 100 to 200w/cm2 range. If unrestricted condensation is assumed, the results of the code are in excellent agreement with more conventional saturation models. (author)

  8. Plant operation performance improvements of the General Electric (GE) boiling water reactors (BWR'S)

    International Nuclear Information System (INIS)

    This paper summarizes some of the plant operation performance improvement techniques developed by the General Electric Company Nuclear Energy Business Operation for the General Electric Boiling Water Reactors (GE BWR's). Through the use of both thermal and plant hardware operating margins, substantial additional flexibility in plant operation can be achieved resulting in significant improvements in plant capacity and availability factor and potential fuel cycle economics for the currently operating or requisition GE BWR plants. This list of techniques includes expanding the BWR thermal power/moderator flow operating domain to the maximum achievable region, operation with a single recirculation loop out of service and operation at rated thermal power with reduced feedwater temperatures. These plant improvements and operating techniques can potentially increase plant capacity factor by 1% to 2% and provide additional fuel cycle economics savings to the GE BWR's owners

  9. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼21 Btu/lbm. A sensitivity study with regard to the steam separator pressure - loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty In the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power. Implications of the Present Study: With the exception of guidance to avoid SRV cycling, these preliminary results have provided further support for the validity of this response strategy. SABRE calculations have shown that the reactor is probably slightly unstable in natural circulation operation, but relief valve cycling prevents the occurrence of instabilities at or near design pressure. Thus, unstable operation should not be a concern when boron injection and HPCI are available and depressurization is unnecessary. The reactor water level for injection flow corresponding to HPCI operation has been shown to be acceptable and consistent with earlier estimates based on the NSAC results; however, condensation effects have a significant influence on the equilibrium reactor water level. In addition, the SABRE results reinforce PP and L's concerns that

  10. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    OpenAIRE

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use ...

  11. Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

    OpenAIRE

    Hsoung-Wei Chou; Chin-Cheng Huang

    2015-01-01

    The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to ...

  12. Reactor physics calculations on MOX fuel in boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    The loading of MOX (Mixed Oxide) fuel in BWRs (Boiling Water Reactors) is considered in this paper in a ''once-through'' strategy. The fuel assemblies are of the General Electric 8 x 8 type, whereas the reactor is of the General Electric BWR/6 type. Comparisons with traditional UOX (Uranium Oxide) fuel assemblies revealed that the loading of MOX fuel in BWRs is possible, but this type of fuel creates new problems that have to be addressed in further detail. The major ones are the SDM (Shutdown Margin) and the stability of the cores at BOC (beginning of cycle), which were demonstrated to be significantly lowered. The former requires a new design of the control rods, whereas a modification of the Pu isotopic vector allows improving the latter. Another issue with the use of the MOX fuel assemblies in a ''once-through'' strategy is the increased radiotoxicity of the discharged fuel assemblies, which is much higher than of the UOX fuel assemblies. (author)

  13. Effectiveness of a Large Number of Control Rods in the Second Charge of the Halden Boiling Water Reactor

    International Nuclear Information System (INIS)

    The reactivity worth of various control-rod configurations has been measured in the second fuel charge of the Halden Boiling Heavy Water Reactor (HBWR) under low power conditions. The second fuel charge of HBWR consists of 7-rod UO2 cluster elements with 1.5% enrichment. A total of 30 control rods is placed in the open positions of the hexagonal fuel-lattice structure. In older to facilitate theoretical comparisons, measurements have been made on symmetrical control-rod configurations only. The experiment consisted of measuring the critical water level for the clean core and with the different rod configurations inserted to various distances from the bottom of the reactor. The temperature dependence of the reactivity worth was investigated by performing measurements, using a ring of 6 control rods, at the three different temperatures 34°C, 150°C and 220°C. Comparisons of the experimentally-determined critical water levels and the calculated critical water levels are presented. The critical water levels are calculated both by a method in which the control rods are homogenized together with fuel and moderator to form a control-rod zone, and also by a heterogeneous method in which the fuel elements and control rods are regarded as line sinks to thermal neutrons and the fuel elements are regarded as line sources of fast neutrons. (author)

  14. Modeling of the acoustic boiling noise of sodium during an assembly blockage in sodium-cooled reactors

    International Nuclear Information System (INIS)

    In the framework of the fourth generation of nuclear reactors safety requirements, the acoustic boiling detection is studied to detect subassembly blockages. Boiling, that might occur during subassembly blockages and that can lead to clad failure, generates hydrodynamic noise that can be related to the two-phase flow. A bubble dynamics study shows that the sound source during subassembly boiling is condensation. This particular phenomenon generates most noise as a high subcooling is present in the subassembly and because of the high thermal diffusivity of sodium. This result leads to an estimate of the form of the acoustic spectrum that will be filtered and amplified during propagation inside the liquid. And even though it is unlikely that bubbles will be present inside the subassembly, due to the very gradual temperature profile at the wall and due to the geometry that leads to a strong confinement of the vapor, the historical bubble dynamics approach gives some insight in previous measurements. Additionally, some hypotheses can be disproved. These theoretical ideas are validated with a small water experiment, yet it also shows that a simple experience in sodium doesn't lead to a better knowledge of the acoustic source. A theoretical analysis also revealed that a realistic experiment with a simulant fluid, such as water or mercury, isn't representative. A similar conclusion is obtained when studying cavitation as a simulant acoustic source. As such, the acoustic detection of boiling, in comparison with other detection systems, isn't sufficiently developed yet to be applied as a reactor protective system. (author)

  15. Revision of nucleated boiling mechanisms

    International Nuclear Information System (INIS)

    The boiling occurrence plays an important role in the power reactors energy transfer. But still, there is not a final theory on the boiling mechanisms. This paper presents a critical analysis of the most important nucleated boiling models that appear in literature. The conflicting points are identified and experiments are proposed to clear them up. Some of these experiments have been performed at the Thermohydraulics laboratory (Bariloche Atomic Center). (Author)

  16. Diffraction Experiments at the IBR-2 Pulsed Reactor with Methane Cold Neutron Source

    CERN Document Server

    Balagurov, A M; Mironova, G M; Pole, A V; Simkin, V G

    2000-01-01

    A new methane cold neutron source has been tested at the IBR-2 pulsed reactor at the Frank Laboratory of Neutron Physics. In a paper the results of experiments at neutron diffractometers HRFD and DN-2 which are placed at the IBR-2 from the methane moderator side are given. A comparison with the results obtained with the conventional water comb-like moderator is performed. The perspectives of the cold source for various kinds of neutron diffraction experiments, including atomic and magnetic structural analysis and real time experiments are discussed. It is shown, that for a huge number of the experiments which are performing at both HRFD and DN-2 the methane cold neutron source provides the better conditions than water comb-like moderator.

  17. Interpretation of corrosion potential data from boiling-water reactors under hydrogen water chemistry conditions

    International Nuclear Information System (INIS)

    A method was devised to estimate electrochemical conditions at the entrance to the recirculation piping of a boiling water reactor under hydrogen water chemistry (HWC) conditions from electrochemical corrosion potential (ECP) measurements made in remote autoclaves. The technique makes use of the mixed potential model to estimate ECP in the autoclaves and compares estimates to measured values in an optimization on the concentrations of hydrogen peroxide and oxygen in the recirculation system. The algorithm recognizes that H2O2 decomposes in sampling lines and that transit times between the recirculation system and monitoring points depend upon flow rates and sampling line diameters. An analysis was made of ECP data from three monitoring locations in the Barseback BWR in Sweden, as a function of H2 concentration in the feedwater for two flow rates (5,500 kg/s and 6,300 kg/s for the four recirculation loops). HWC did not displace ECP below a critical value of -0.23 VSHE at the lower flow rate until the reactor water [H2] exceeded 0.15 ppm, corresponding to a feedwater H2 level of > 0.93 ppm. At the higher flow rate of 6,300 kg/s (divided equally between four recirculation loops), protection was not predicted until the feedwater [H2] exceeded 1.2 ppm, corresponding to a reactor water [H2] of ∼ 0.195 ppm. The difference was attributed to the greater persistence of H2O2 at high feedwater [H2] at the higher flow rate, possibly because of the lower transit time from the core to the recirculation system

  18. Introduction to the study of boiling in water reactors

    International Nuclear Information System (INIS)

    In order to plot the low and high power transfer function for a reactor using its background, a conventional method is proposed here for estimating the efficiency of a CC5 chamber associated to a direct current detection system. (author)

  19. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    International Nuclear Information System (INIS)

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs

  20. Development and Capabilities of ISS Flow Boiling and Condensation Experiment

    Science.gov (United States)

    Nahra, Henry; Hasan, Mohammad; Balasubramaniam, R.; Patania, Michelle; Hall, Nancy; Wagner, James; Mackey, Jeffrey; Frankenfield, Bruce; Hauser, Daniel; Harpster, George; Nawrocki, David; Clapper, Randy; Kolacz, John; Butcher, Robert; May, Rochelle; Chao, David; Mudawar, Issam; Kharangate, Chirag R.; O'Neill, Lucas E.

    2015-01-01

    An experimental facility to perform flow boiling and condensation experiments in long duration microgravity environment is being designed for operation on the International Space Station (ISS). This work describes the design of the subsystems of the FBCE including the Fluid subsystem modules, data acquisition, controls, and diagnostics. Subsystems and components are designed within the constraints of the ISS Fluid Integrated Rack in terms of power availability, cooling capability, mass and volume, and most importantly the safety requirements. In this work we present the results of ground-based performance testing of the FBCE subsystem modules and test module which consist of the two condensation modules and the flow boiling module. During this testing, we evaluated the pressure drop profile across different components of the fluid subsystem, heater performance, on-orbit degassing subsystem, heat loss from different modules and components, and performance of the test modules. These results will be used in the refinement of the flight system design and build-up of the FBCE which is manifested for flight in late 2017-early 2018.

  1. A potential of boiling water power reactors with a natural circulation of a coolant

    International Nuclear Information System (INIS)

    The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

  2. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    OpenAIRE

    Alejandro Nuñez-Carrera; Raúl Camargo-Camargo; Gilberto Espinosa-Paredes; Adrián López-García

    2012-01-01

    The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the...

  3. Investigation of void effects in boiling water reactor fuels using neutron tomography

    OpenAIRE

    Loberg, John

    2006-01-01

    In a boiling water reactor (BWR), the void is correlated to dry out and the power level of the reactor. However, measuring the void is very difficult so it is therefore calculated with an accuracy that leaves room for improvements. Typically the uncertainty is ± 3% for 40% average void in the reactor. If the void could be determined with improved accuracy, both safety and economical features could be improved. X-ray tomography has previously been done on BWR fuel models in order to determine ...

  4. Reactor operation experience

    International Nuclear Information System (INIS)

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  5. New Reactor Siting, Licensing and Construction Experience. Proceedings of the 2. CNRA International Workshop on 'New Reactor Siting, Licensing and Construction Experience'

    International Nuclear Information System (INIS)

    This report documents the proceedings from the 2. Workshop on New Reactor Siting, Licensing and Construction Experience. A total of 45 specialists from 16 countries and international organisations attended. The meeting was sponsored by the OECD Nuclear Energy Agency Committee on Nuclear Regulatory Activities and hosted by the US Nuclear Regulatory Commission (U.S.NRC). The objectives of the workshop were to provide a forum to exchange information on lessons learned from siting, licensing and constructing new nuclear power plants around the world. Key focus areas included siting practices and regulatory positions that have been enhanced as a result of the Fukushima accident; lessons learned from licensing and design review approaches and challenges, construction experience and recommendations for regulatory oversight; and regulatory cooperation on generic and design specific issues through the MDEP specific working groups. The workshop was structured in 4 technical sessions, each followed by ample time for panel discussions. The first technical session was devoted to regulatory cooperation on generic and design specific issues, MDEP working groups (EPR, AP1000), vendor inspection co-operation, digital I and C, and codes and standards. The second technical session was intended to discuss and share regulatory positions on siting practices and enhancements as a result of lessons learned from Fukushima accident. The third technical session addressed the construction experience and regulatory oversight of new reactor construction activities. And the fourth technical session included presentations on the lessons learned from regulatory licensing reviews of new reactor designs

  6. The control of a boiling water reactor power plant for example Muehleberg

    International Nuclear Information System (INIS)

    Simplified fluid circuit flow diagrams are given for two boiling water reactor types, the first having an outer circuit with the boiling water vessel, turbine, condenser and feed-pump and an inner circuit circulating water within the pressure vessel; the second type has a primary loop for the pressure vessel, a heat exchanger and a secondary loop for the turbine and condenser. The first type has been used at Muehleberg, Leibstadt and Kaiseraugst, and the second at Beznau, and Goesgen. A control circuit illustration is given based on Muehleberg and incorporating a proportional (P) controller in the boiling water side of the outer loop and two PID controllers in the condensate return line. A PI regulator is included in the inner loop. (G.C.)

  7. Neutronic analysis and validation of boiling water reactor core designed by MCNPX code

    International Nuclear Information System (INIS)

    Highlights: • MCNPX code is used to design a model for BWR core. • The fuel enrichment is distributed in such a way to flat the power. • Validation of the BWR core model designed by MCNPX code. • Calculate Pu and its isotopes concentration at different burnup. - Abstract: This paper presents a design of boiling water reactor BWR model using MCNPX to develop benchmarks for checking the fuel management computer code packages. MCNPX code based on Monte Carlo method, is used to design a three dimensional model for BWR fuel assembly in typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. This design is used to study the thermal neutron flux and the pin by pin power distribution through the BWR core assemblies. The fuel used in BWR core is UO2 with three different types of enrichment (0.711%, 1.76% and 2.19%). This enrichment is distributed in such a way as to flatten the power. The effect of different enrichment values on the radial normalized power distribution is analyzed. The spent fuel in the reactor can be recycled, and plutonium and its isotopes can be extracted

  8. Spectral measurements of direct and scattered gamma radiation at a boiling-water reactor site

    International Nuclear Information System (INIS)

    Quantitative surveys of direct and scattered gamma radiation emitted from the steam-power conversion systems of a boiling-water reactor and other on-site radiation sources were made using a directionally shielded HPGe gamma spectrometry system. The purpose of this study was to obtain data on the relative contributions and energy distributions of direct and scattered gamma radiation in the site environs. The principal radionuclide of concern in this study is 16N produced by the 16O(n,p)16N reaction in the reactor coolant. Due to changes in facility operation resulting from the implementation of hydrogen water chemistry (HWC), the amount of 16N transported from the reactor to the main steam system under full power operation is excepted to increase by a factor of 1.2 to 5.0. This increase in the 16N source term in the nuclear steam must be considered in the design of new facilities to be constructed on site as well as the evaluation of existing facilities with respect to ALARA (As Low As Reasonably Achievable) dose limits in unrestricted areas. This study consisted of base-line measurements taken under normal BWR chemistry conditions in October, 1987 and a corresponding set taken under HWC conditions in July, 1988. Ground-level and elevated measurements, corresponding to second-story building height, were obtained. The primary conclusion of this study is that direct radiation from the steam-power conversion system is the predominant source of radiation in the site environs of this reactor and that air scattering (i.e. skyshine) does not appear to be significant. (orig.)

  9. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

  10. Cracking in stabilized austenitic stainless steel piping of German boiling water reactors - characteristic features and root cause

    International Nuclear Information System (INIS)

    Cracks have been found in the welds of piping systems made from stabilized austenitic stainless steels in German boiling water reactors (BWR). In the course of the intensive failure analysis metallographic examinations, microstructural investigations by electron microscopy, corrosion experiments and welding tests have been performed. The results show that cracking under the given medium conditions is due to intergranular stress corrosion cracking (IGSCC) in those parts of the heat affected zone (HAZ) which are overheated during welding and where solution of titanium carbides and subsequent precipitation of chromium carbides and depletion of chromium along the affected grain boundaries could occur. (orig.)

  11. Critical Heat Flux during Flow Boiling Experiment with Surfactant Solutions

    International Nuclear Information System (INIS)

    Some additives enhance heat transfer, although, the magnitude and mechanism of enhancement are not consistent or clearly understood. A low concentration of surfactant can also reduce the solution's surface tension considerably, and its level of reduction depends on the amount and type of surfactant present in solution. The surfactant concentrations are usually low enough that the addition of surfactant to water causes no significant change in saturation temperature and most other physical properties, except viscosity and surface tension. Reduced surface tension influences the activation of nucleation sites, bubble growth and dynamics, affecting the boiling heat transfer coefficient. Surfactants effect on CHF (Critical Heat Flux) was determined during flow boiling at atmospheric pressure in closed loop filled with water solutions of tri-sodium phosphate (TSP, Na3PO4.12H2O). TSP was added to the containment sump water to adjust pH level during accidents in nuclear power plants. CHF was measured for four water surfactant solutions at different mass fluxes (100 - 500 kg/m2sec) and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Wettability was determined by measuring the contact angle at different concentration cases that will substantiate any CHF increase

  12. Corrosion products, activity transport and deposition in boiling water reactor recirculation systems

    International Nuclear Information System (INIS)

    The deposition of activated corrosion products in the recirculation loops of Boiling Water Reactors produces increased radiation levels which lead to a corresponding increase in personnel radiation dose during shut down and maintenance. The major part of this dose rate is due to cobalt-60. The following areas are discussed in detail: - the origins of the corrosion products and of cobalt-59 in the reactor feedwaters, - the consolidation of the cobalt in the fuel pin deposits (activation), - the release and transport of cobalt-60, - the build-up of cobalt-60 in the corrosion products in the recirculation loops. Existing models of the build-up of circuit radioactivity are discussed and the operating experiences from selected reactors are summarised. Corrosion chemistry aspects of the cobalt build-up in the primary circuit have already been studied on a broad basis and are continuing to be researched in a number of centers. The crystal chemistry of chromium-nickel steel corrosion products poses a number of yet unanswered questions. There are major loopholes associated with the understanding of activation processes of cobalt deposited on the fuel pins and in the mass transfer of cobalt-60. For these processes, the most important influence stems from factors associated with colloid chemistry. Accumulation of data from different BWRs contributes little to the understanding of the activity build-up. However, there are examples that the problem of activity build-up can be kept under control. Although many details for a quantitative understanding are still missing, the most important correlations are visible. The activity build-up in the BWR recirculation systems cannot be kept low by a single measure. Rather a whole series of measures is necessary, which influences not only cobalt-60 deposition but also plant and operation costs. (author) 26 figs., 13 tabs., 90 refs

  13. Practical application of neutron noise analysis at boiling water reactors

    International Nuclear Information System (INIS)

    The present status in the development of neutron noise methods for diagnostic purposes at BWRs is assessed with respect to practical applications. Three items of interest are briefly reviewed. They are concerned with local phenomena found in neutron noise signals at the higher frequency ranges (above several Hertz). The detection of vibrating in-core instrument tubes and the impacting of fuel element boxes were a problem in which neutron noise analysis substantially contributed. The possibility of detecting bypass flow boiling from neutron noise signatures is a recently proposed concept. Most of the research efforts have been applied to the experimental determination of local characteristics of the two-phase flow which dominates the noise sources in a BWR. Steam velocity measurements in fuel bundles by neutron noise techniques and the derivation of semi-empirical data, e.g. void fraction, bundle power and inlet flow rate, and possibly flow pattern recognition are features for practical use. But there are still effects which are not yet completely understood and require further experimental and theoretical investigations. (Auth.)

  14. Studies on improvements in the control methods of boiling water reactor plant

    International Nuclear Information System (INIS)

    In order to improve the performance of regulation and load following control of boiling water reactor plant, optimal control theory is applied and new types of control method are developed. Case-α controller is first formulated on the basis of the optimal linear regulator theory applied to the linealized model of the system; it is then modified by adding a integration-type action in a feed back loop and by the use of variable gain and reference for adapting to the power level requested. Case-#betta# controller consists of a hierarchical control scheme which has classical P.I. type sub-loop controllers at the first level and a linear optimal regulator at the second level. The controller is designed on the basis of the optimal regulator theory applied to the multivariate autoregressive system model which is obtained from the identification experiments, where the system model is determined with the conventional sub-loop controllers included. The results of the simulation experiments show these control methods proposed have performed fairly well and will be useful for the improvement of the performance of nuclear power plant control. In addition, it is suggested that these control methods will be also attractive for the control of other production plants because these were developed in the attempt to solve the problems deviated from so called 'The gap between the optimal contro theory and actual systems.' (author)

  15. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  16. A two-step method for developing a control rod program for boiling water reactors

    International Nuclear Information System (INIS)

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift

  17. Investigation of BWR [boiling water reactor] instability phenomena using RETRAN-03

    International Nuclear Information System (INIS)

    In 1988, LaSalle, a boiling water reactor (BWR)/5, experienced severe flux oscillations following a trip of both recirculation pumps. The flux oscillations were terminated by an automatic scram at 118% of rated neutron flux. As a result of this event, the U.S. Nuclear Regulatory Commission has asked the BWR utilities to develop procedural or hardware changes that will assure protection of all safety limits. The rapid growth of the oscillations at LaSalle, and the fact that previous stability analyses had predicted the plant to be very stable, emphasizes that a better understanding of this phenomenon is needed before the success of the long-term fixes can be assured. The intent of the Electric Power Research Institute's work was to use BWR transient methods to model reactor instabilities and investigate the factors that dominate this phenomenon. The one-dimensional transient code RETRAN-03 (Ref. 1) was used. The following conclusions are drawn: (1) RETRAN has demonstrated the ability to model BWR instability (nonlinear oscillations). (2) The general system behavior predicted by RETRAN in BWR stability analyses matches theoretical prediction and plant data. (3) These one-dimensional, time-domain results have increased the understanding of BWR stability phenomena and have helped optimize the long-term solutions being developed by the utilities

  18. Detection of boiling by Piety's on-line PSD-pattern recognition algorithm applied to neutron noise signals in the SAPHIR reactor

    International Nuclear Information System (INIS)

    A partial blockage of the cooling channels of a fuel element in a swimming pool reactor could lead to vapour generation and to burn-out. To detect such anomalies, a pattern recognition algorithm based on power spectra density (PSD) proposed by Piety was further developed and implemented on a PDP 11/23 for on-line applications. This algorithm identifies anomalies by measuring the PSD on the process signal and comparing them with a standard baseline previously formed. Up to 8 decision discriminants help to recognize spectral changes due to anomalies. In our application, to detect boiling as quickly as possible with sufficient sensitivity, Piety's algorithm was modified using overlapped Fast-Fourier-Transform-Processing and the averaging of the PSDs over a large sample of preceding instantaneous PSDs. This processing allows high sensitivity in detecting weak disturbances without reducing response time. The algorithm was tested with simulation-of-boiling experiments where nitrogen in a cooling channel of a mock-up of a fuel element was injected. Void fractions higher than 30 % in the channel can be detected. In the case of boiling, it is believed that this limit is lower because collapsing bubbles could give rise to stronger fluctuations. The algorithm was also tested with a boiling experiment where the reactor coolant flow was actually reduced. The results showed that the discriminant D5 of Piety's algorithm based on neutron noise obtained from the existing neutron chambers of the reactor control system could sensitively recognize boiling. The detection time amounts to 7-30 s depending on the strength of the disturbances. Other events, which arise during a normal reactor run like scrams, removal of isotope elements without scramming or control rod movements and which could lead to false alarms, can be distinguished from boiling. 49 refs., 104 figs., 5 tabs

  19. Proceedings of the International Workshop on Boiling Water Reactor Stability

    International Nuclear Information System (INIS)

    With regard to technical understanding of the phenomena, the participants agreed that the causes of instability appear to be well understood, but there are many variables involved, and their correlation with instability conditions is not always certain. Most codes claimed to be capable of predicting oscillations and unstable conditions, based on post-test analyses of data from actual events, but there do not seem to be any blind predictions available which accurately predict an instability event before the actual test results are released. As a result, reactor owners have decided that the best course is to avoid, with sufficient margin, certain regions in the power-flow map where regions of instability are known to exist, rather than try to predict them very accurately. The meeting concluded that the safety significance of BWR instability is rather limited, and current estimates of plant risk do not show it to be a dominant contributor. This is because the installed plant protection systems will shut a reactor down when the oscillations exceed power limits, and any fuel damage which might occur will be localized and containable. However, it was also agreed that an instability event could increase uncertainties in the human error rate, because operators who have never seen an unstable reactor may take actions which are not necessarily the best for the particular situation. In addition, although an instability event may not cause any harm to the public, it may cause some fuel failures, and these are certainly a concern to a reactor owner, for economic and radiation protection reasons. Finally, it was also agreed that BWR instability is certainly considered to be significant by the public, where acceptance of the technology would erode if a plant is perceived to be in an uncontrolled state, regardless of the actual risk inherent in the situation

  20. Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

    International Nuclear Information System (INIS)

    Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs

  1. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  2. The synergic impact of the boiling and water radiolysis on the pressurized water reactor fuel cladding's chemical environment

    Energy Technology Data Exchange (ETDEWEB)

    Dobrevski, I.; Zaharieva, N. [Bulgarian Academy of Sciences, Inst. for Nuclear Research and Nuclear Energy, Sofia (Bulgaria)

    2010-07-01

    By the presence of local boiling at the cladding surfaces of pressurized water reactors (PWRs), including WWER-1000 Units, the behaviors of gases dissolved in water phase (coolant) is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. On the other hand it is known that the hydrogen is added to primary coolant of PWRs, in order to avoid the production of oxidants as radiolysis of water products. It is clear that if boiling strips out dissolved hydrogen, the creation of local oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O{sub 2}) and hydrogen (H{sub 2}), but also hydrogen peroxide (H{sub 2}O{sub 2}) will be produced. While the resulting by water radiolysis hydrogen and oxygen can be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in the wall water phase and will act as an important factor for creation of oxidizing conditions in fuel cladding environment, together with some water radiolytical radicals: ·OH, HO{sub 2}·/ O{sub 2}{sup -}. Summarizing of the above mentioned allows the conclusion that creation of oxidizing conditions in the nuclear fuel cladding environment is not a direct boiling consequence but, in fact, is a result (consequence) of the synergic impact of the boiling- and water radiolysis- processes on the Pressurized Water Reactor fuel cladding surface areas. The PWRs experiences confirm that the density of SNB (sub-cooled nucleate boiling), resp. steaming rate, control the degree of the above mentioned water radiolysis processes. If it is not possible to moderate the steaming rate of the fuel cladding surfaces in PWRs, the only way to avoid the cladding damages caused by the local oxidizing conditions, is the applying of cladding materials

  3. Conceptual design and thermal-hydraulic characteristics of natural circulation Boiling Water Reactors

    International Nuclear Information System (INIS)

    A natural circulation boiling water reactor (BWR) with a rated capacity of 600 MW (electric) has been conceptually designed for small- and medium-sized light water reactors. The components and systems in the reactor are simplified by eliminating pumped recirculation systems and pumped emergency core cooling systems. Consequently, the volume of the reactor building is -- 50% of that for current BWRs with the same rated capacity; the construction period is also shorter. Its thermal-hydraulic characteristics, critical power ratio (CPR) and flow stability at steady state, decrease in the minimum CPR (ΔMCPR) at transients, and the two-phase mixture level in the reactor pressure vessel (RPV) during accidents are investigated. The two-phase mixture level in the RPV during an accident does not decrease to lower than the top of the core; the core uncovery and heatup of fuel cladding would not occur during any loss-of-coolant accident

  4. A multi-cycle BWR [boiling water reactor] core reload design analysis system (MCAS)

    International Nuclear Information System (INIS)

    This paper describes the design, construction, and application of a software system (MCAS) for performing boiling water reactor reload core design analysis. MCAS provides for the execution of studies which analyze alternative reload strategies over a range of cycles. Studies are performed by preparing and executing sequential SIMULATE-E Haling depletions and storing the results on a data base for subsequent reporting and analysis. Application of MCAS has shown that the ability of efficiently and accurately predict the effects of next cycle design decisions on future cycles is a valuable capability. This capability results in the proper selection of BWR [boiling water reactor] reload fuel bundle enrichment and batch size as necessary for reload fuel supply planning and early identification and resolution of design problems which would prove expensive if discovered at a later time

  5. Simulation and fault-detection of a pressure control servosystem in a Boiling Water Reactor

    International Nuclear Information System (INIS)

    This master thesis describes a Simnon model of a boiling water reactor to be used in simulating faults and disturbances. These faults and disturbanses will be detected by noise analysis. Some methods in identification and noise analysis are also described and are applied on some malfunctions of a servo. A Pascal program for recursive parameter identification was also written and tested. This program is to be used in an expert system for noise analysis on the nuclear power plant Barsebaeck. (author)

  6. Evaluation of pressurized thermal shock in transitional condition for boiling water reactor pressure vessel

    International Nuclear Information System (INIS)

    The structural integrity for Pressurized Thermal Shock (PTS) was evaluated for the RPVs of Japanese Boiling Water Reactors (BWRs). It has been clarified that the BWR RPVs have the sufficient margin of fracture toughness by calculating the stress intensity factor in transitional condition and the acceptance criteria for RPV shell plate which is assumed to be neutron-irradiated in core region for 60 years. (author)

  7. Flow processes during subcooled boiling in fuel rod clusters of water-cooled reactors

    International Nuclear Information System (INIS)

    The theoretical fundamentals for the thermohydraulic calculation of fuel rod clusters in light water-cooled reactors are presented with special regard to boiling on fuel rods in unsaturated water. It is shown which preconditions concerning the structure of the two-phase flow must be met in order to apply the methods of single-phase continuum mechanics to two-phase flows. (orig./TK)

  8. A numerical study of boiling flow instability of a reactor thermosyphon system

    Energy Technology Data Exchange (ETDEWEB)

    Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der [Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew [Shell Research and Technology Centre, Badhuisweg 3, 1031 CM Amsterdam (Netherlands)

    2006-04-01

    A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed. [Author].

  9. Operating and test experience with Experimental Breeder Reactor number 2 (EBR-II), the Integral Fast Reactor (IFR) prototype

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor number 2 (EBR-II) has operated for 30 years, the longest for any liquid metal cooled reactor (LMR) power plant in the world. Given the scope of what has been developed and demonstrated over those years, it is arguably the most successful test reactor operation ever. Tests have been carried out on virtually every fast reactor fuel type. The reactor itself has been extensively studied. The most dramatic safety tests, conducted on 3 April, 1986, showed that an LMR with metallic fuel could safely accommodate loss of flow or loss of heat-sink without scram. EBR-II operated as the Integral Fast Reactor (IFR) prototype, demonstrating important innovations in safety, plant design, fuel design and actinide recycle. The ability to accommodate anticipated transients without scram passively resulted in significant simplification of the reactor plant, primarily through less reliance on emergency power and not having to require the secondary sodium or steam systems to be safety grade. These features have been quantified in a probabilistic risk assessment (PRA) conducted for EBR-II, demonstrating considerable safety advantages over other reactor concepts. Fundamental to the superior safety and operating characteristics of this reactor is the metallic U-Pu-Zr alloy fuel. Performance of the fuel has been fully proven: achieved burnup levels exceed 20 at.% in the lead test assemblies. A complete set of fuel performance and safety limits has been developed and was carried forward in formal safety documents supporting conversion of the core to IFR fuel. The last major demonstration planned was to assess the performance of recycled actinides in the fuel and to confirm that passive safety characteristics are maintained with recycled actinide fuel in the core. (author)

  10. Improvements of fuel failure detection in boiling water reactors using helium measurements

    International Nuclear Information System (INIS)

    To certify a continuous and safe operation of a boiling water reactor, careful surveillance of fuel integrity is of high importance. The detection of fuel failures can be performed by off-line gamma spectroscopy of off-gas samples and/or by on-line nuclide specific monitoring of gamma emitting noble gases. To establish the location of a leaking fuel rod, power suppression testing can be used. The accuracy of power suppression testing is dependent on the information of the delay time and the spreading of the released fission gases through the systems before reaching the sampling point. This paper presents a method to improve the accuracy of power suppression testing by determining the delay time and gas spreading profile. To estimate the delay time and examine the spreading of the gas in case of a fuel failure, helium was injected in the feed water system at Forsmark 3 nuclear power plant. The measurements were performed by using a helium detector system based on a mass spectrometer installed in the off-gas system. The helium detection system and the results of the experiment are presented in this paper. (authors)

  11. Time domain model sensitivity in boiling water reactor stability analysis using TRAC/BF1

    International Nuclear Information System (INIS)

    Boiling water nuclear reactors (BWRs) may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate because of the tight coupling of flow to power, especially under gravity-driven circulation. To predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model is developed for a typical BWR. Using this tool, it is demonstrated that density waves may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases are analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. As predicted by others, the two-phase friction controls the extent of the oscillation. Because of this sensitivity, existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from one case to another. It is found that higher dimensional nuclear feedback models reduce the extent of the oscillation

  12. Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

    Science.gov (United States)

    Yoshida, Hiroyuki; Takase, Kazuyuki

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low.

  13. Severe accident mitigation features of the economic simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper provides an overview of the Economic Simplified Boiling Water Reactor (ESBWR)severe accident mitigation systems. The major severe accident types are described and the systems credited for mitigating the severe accidents are discussed, including the Basemat Internal Melt Arrest Coolability (BiMAC) device, the Passive Containment Cooling System (PCCS), and the advantages of suppression pool water for scrubbing during containment venting. The ruggedness of the containment and reactor building designs for accommodating beyond design accident conditions is also discussed. (author)

  14. The effects of aging on Boiling Water Reactor core isolation cooling system

    International Nuclear Information System (INIS)

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes

  15. Correlation Fourier diffractometry: 20 Years of experience at the IBR-2 reactor

    Science.gov (United States)

    Balagurov, A. M.; Bobrikov, I. A.; Bokuchava, G. D.; Zhuravlev, V. V.; Simkin, V. G.

    2015-05-01

    The high-resolution Fourier diffractometer (HRFD) was commissioned at the IBR-2 pulsed reactor at FLNP JINR in 1994. The specific feature of the HRFD design is the use of fast Fourier chopper for modulating the primary neutron beam intensity and the correlation method of diffraction data acquisition. This allowed to reach with HRFD extremely high resolution (Δ d/ d ≈ 0.001) over a wide range of inter-planar spacings at a relatively short flight path between chopper and sample ( L = 20 m). Over time, a lot of diffraction experiments on crystalline materials, the main goal of which was to study their atomic and magnetic structures, were performed at HRFD. Successful implementation of the Fourier diffractometry technique at the IBR-2 reactor stimulated the construction of yet another Fourier diffractometer intended for internal mechanical stress studies in bulk materials (FSD, Fourier Stress Diffractometer). In this paper the experience of using this technique at the IBR-2, which is a long-pulse neutron source, is considered, the examples of HRFD studies are given, and possible solutions for existing technical problems of using correlation diffractometry and ways of increasing the intensity and resolution of HRFD are discussed.

  16. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  17. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs

  18. Non normal modal analysis of oscillations in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas

    2013-07-01

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

  19. Non normal modal analysis of oscillations in boiling water reactors

    International Nuclear Information System (INIS)

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

  20. Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

    2004-11-01

    One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

  1. Phenomenology and thermo-hydraulic stability of the CAREM-25 reactor: Evaluation of subcooled boiling effect

    International Nuclear Information System (INIS)

    In this article the phenomenology present in self/pressurized, integral, natural circulation, low thermodynamic quality nuclear reactors similar to CAREM-25 is investigated. In particular, analytical relations for the mass flow rate, the core mean enthalpy and the location of the two phase boundary are derived in terms of the so-called natural variables of the system: the nuclear power, the condensation power and the system pressure. In addition, some consequences of the flashing phenomenon in the reactor thermal-hydraulics are discussed emphasizing those affecting the reactor stability. The reactor stability performance was studied by using the HUARPE code which is a low diffusive code. The stability results obtained by neglecting the subcooled effect in the system are presented in the so-called the stability maps in which the results are presented for a wide range of conditions. The stability effect caused by the presence of subcooled boiling in the reactor core was also examined. In order to investigate such a consequence, the code was slightly modified such that the predicted vapor profile in the hot leg is similar to that estimated by RELAP system code at steady state conditions. The simple implemented algorithm allows varying a free parameter with which a broad number of cases can be studied. This is important since the subcooled boiling predictions generally have large uncertainties and therefore to cover a large number of situations is desired to derive confident conclusions. The results show the existence of vapor created by means of subcooled boiling enhances the system stability for a wide range of conditions. For this reason from this preliminary investigation, it is concluded neglecting the subcooled effect in CAREM-25 stability studies is a conservative criterion (author))

  2. The Simplified Boiling Water Reactor (SBWR) turbine generator foundation

    International Nuclear Information System (INIS)

    A spring-mounted turbine generator foundation is well suited to the design objectives of the SBWR. Because of the experience gained from more than 200 nuclear and non-nuclear installations, the structural feasibility of the design concept has been proven. The only exception is that there is no known seismic experience for this design in a major earthquake. Analyses are performed to compare the seismic response of the spring-mounted design to that of a conventional design. These analyses demonstrate that the design is seismically feasible. Because of the many advantages to be realized, the SBWR turbine generator foundation design will continue based on the spring-mounted concept. (author)

  3. A diagnostic expert system for a boiling water reactor using a dynamic model

    International Nuclear Information System (INIS)

    A diagnostic expert system for abnormal disturbances in a BWR (Boiling Water Reactor) plant has been developed. The peculiar feature of this system is a diagnostic method which combines artificial intelligence technique with numerical analysis technique. The system has three diagnostic functions, 1) identification of anomaly position (device or sensor), 2) identification of anomaly mode and 3) identification of anomaly cause. Function 1) is implemented as follows. First, a hypothesis about anomaly propagation paths is built up by qualitative reasoning, using knowledge of causal relations among observed signals. Next, the abnormal device or sensor is found by applying model reference method and fuzzy set theory to test the hypothesis, using knowledge of plant structure and function, heuristic strategy of diagnosis and module type dynamic simulator. This simulator is composed of basic transfer function modules. The simulation model for the testing region is built up automatically, according to the requirement from the diagnostic task. Function 2) means identification of dynamic characteristics for an anomaly. It is realized by tuning model parameters so as to reproduce the abnormal signal behavior using the non-linear programing method. Function 3) derives probable anomaly causes from heuristic rules between anomaly mode and cause. A basic plant dynamic model was built up and adjusted to dynamic characteristics for one BWR plant (1100MWe). In order to verify the diagnostic functions of this system, data for several abnormal events was compiled by modifying this model. The diagnostic functions were proved useful, through the simulated abnormal data

  4. Development of open code system for core design of boiling water reactor

    International Nuclear Information System (INIS)

    A new core design system for the Boiling Water Reactor (BWR), HANCS, has been developed. HANCS consists of HIDEC, ALLIS and NORMA, which are open source codes. HIDEC which consists of MVP2.0 and ORIGEN2.1 performs the assembly calculation. ALLIS generates the nuclear constants library for the core calculation. NORMA is introduce in order to perform the core calculation. HANCS was developed by coupling these codes with some other utility programs. HANCS was verified by comparing the calculation results by CASMO-SIMULATE as the reference code. In the verification, the results of the core calculation, such as k-effective, the relative power, the void fraction and the fuel temperature, were compared for the initial loading core and the equilibrium core. In the initial loading core analysis, the calculation results of HANCS agreed well with those of CASMO-SIMULATE under both the zero power condition and the full power operation. In the equilibrium core analysis, although the difference of the void fractions between HANCS and CASMO-SIMULATE was found, the void fractions finally agreed well with those of CASMO-SIMULATE by changing the thermal-hydraulic options of HANCS. The other results also agreed well. It is concluded by the verification that HANCS is appropriate for the BWR core analysis. (author)

  5. Modeling of fuel-rod behavior during reactor power cycling and ramping experiments with computer code FRAPCON-2

    International Nuclear Information System (INIS)

    Modelling of fuel-rod behavior during reactor power cycling and ramping (including power-cooling mismatch experiments) with the computer code FRAPCON-2 is discussed. FRAPCON-2 computer calculations, using different mechanical models (Rigid Pellet, Deformable Pellet and Finite Element Mechanical Models) are compared with experimental results. The range of conditions over which FRAPCON-2 may be applied for PWR fuel rod behavior modelling during reactor power cycling and ramping are illustrated

  6. Operating experiences of the research reactors

    International Nuclear Information System (INIS)

    Nuclear research reactors are devices of wide importance, being used for different scientific research tasks, for testing and improving reactor systems and components, for the production of radioisotopes, for the purposes of defence, for staff training and for other purposes. There are three research reactors in Yugoslavia: RA, RB and TRIGA. Reactors RA and RB at the 'Boris Kidric' Institute of Nuclear Sciences are of heavy water type power being 6500 and 10 kW, and maximum thermal neutron flux of 1014 and 1011(n/cm2s), respectively. TRIGA reactor at the 'Jozef Stefan' Institute in Ljubljana is of 250 kW power and maximum thermal neutron flux of 1013(n/cm2s). Reactors RA and RB use soviet fuel in the form of uranium dioxide (80% enriched) and metallic uranium (2%). Besides, RB reactor operates with natural uranium too. TRIGA reactor uses american uranium fuel 70% and 20% enriched, uranium being mixed homogeneously with moderator (ZrH). Experiences in handling and controlling the fuel before irradiation in the reactor, in reactor and after it are numerous and valuable, involving either the commercial arrangements with foreign producers, or optimal burn up in reactor or fuel treatment after the reactor irradiation. Twenty years of operating experience of these reactors have great importance especially having in mind the number of trained staff. Maintenance of reactors systems and fluids in continuous operation is valuable experience from the point of view of water reactor utilization. The case of the RA reactor primary cycle cobalt decontamination and other events connected with nuclear and radiation security for all three reactors are also specially emphasized. Owing to our research reactors, numerous theoretical, numerical and experimental methods are developed for nuclear and other analyses and design of research and power reactors,as well as methods for control and protection of radiation. (author)

  7. The Nuclear option for U.S. electrical generating capacity additions utilizing boiling water reactor technology

    International Nuclear Information System (INIS)

    The technology status of the Advanced Boiling Water (ABWR) and Simplified Boiling Water (SBWR) reactors are presented along with an analysis of the economic potential of advanced nuclear power generation systems based on BWR technology to meet the projected domestic electrical generating capacity need through 2005. The forecasted capacity needs are determined for each domestic North American Electric Reliability Council (NERC) region. Extensive data sets detailing each NERC region's specific generation and load characteristics, and capital and fuel cost parameters are utilized in the economic analysis of the optimal generation additions to meet this need by use of an expansion planning model. In addition to a reference case, several sensitivity cases are performed with regard to capital costs and fuel price escalation

  8. Transposition of 2D Molten Corium–Concrete Interactions (MCCI) from experiment to reactor

    International Nuclear Information System (INIS)

    Highlights: • Simple scaling rules and approaches for MCCI are outlined and discussed. • A detailed MCCI code is compared with simplified approaches. • The assumption of a quasi-steady-state energy balance is an adequate simplification. • Simplified procedures are a useful code-independent scaling-up method for MCCI. - Abstract: In the course of a severe accident in a light water reactor, the interactions of corium with the concrete structures of the reactor cavity (Molten Corium–Concrete Interactions or MCCI) may have a significant impact on the long-term integrity of the containment. The 2D behaviour of the melt pool contained in the reactor cavity under dry or top flooding conditions is considered as one of the key phenomena. The “scaling” issue is usually resolved by – in a first step – identifying the impact of physical mechanisms on the process and – in a second step – evaluating these mechanisms at scaled conditions regarding time and length. The conditions for the MCCI change with time due to the evolution of the melt’s state defined by e.g., its composition, temperature and solid fraction, and due to the change of cavity contour and the decreasing decay heat. Here, simplified models are investigated with the objective to infer from laboratory-scale experiments how basic and important parameters like the temperature of the melt and the erosion depth evolve with time if transposed to reactor scale. Due to the simplifications in the models under consideration, the MCCI is analysed assuming “ideal” boundary conditions as e.g., an evolution of a cavity contour with time while retaining its geometrical shape (sphere, cylinder, etc.). Based on these idealised assumptions, generic trends for physical parameters like melt temperature, heat flux at the pool boundary surface, concrete fraction in the melt, viscosity, etc. can be deduced. Simple scaling methods are introduced and checked for consistency by comparison calculations with the

  9. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, M.P. [Instituto de Estudos Avancados - CTA, Sao Paolo (Brazil); Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.

  10. Corrosion product deposition on fuel element surfaces of a boiling water reactor

    International Nuclear Information System (INIS)

    Over the last decade the problem of corrosion products deposition on light water reactor fuel elements has been extensively investigated in relation to the possibility of failures caused by them. The goal of the present study is to understand in a quantitative way the formation of such kind of deposits and to analytically understand the mechanism of formation and deposition with help of the quasi-steady state concentrations of a number of 3d metals in reactor water. Recent investigations on the complex corrosion product deposits on a Boiling Water Reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The buildup of magnetic corrosion product deposits (crud) on the fuel cladding of the BWR, Kernkraftwerk Leibstadt (KKL) Switzerland has hampered the Eddy-current based measurements of ZrO2 layer thickness. The magnetic behavior of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main components identified by 2D μXRD mapping inside the layer were hematite and spinel phases with the common formula (MxFey)[M(1-x)Fe(2-y)]O4, where M = Zn, Ni, Mn. With μXRD it was clearly shown that the cell parameter of analyzed spinel is different from the one of the pure endmembers (ZnFe2O4, NiFe2O4 and MnFe2O4) proving the existence of solid solutions. These

  11. Fast reactor operating experience

    International Nuclear Information System (INIS)

    At the beginning of electricity generation from nuclear power there was the breeder, which fulfilled its duty in a number of smaller test and experimental reactors within national programs. Over the years, some of those reactors have attained impressive availabilities, while others have helped to improve our knowledge by the negative results they contributed. Worldwide a decisive step was taken by the mid- to late sixties in the planning and construction of medium sized demonstration fast breeder power plants (250 to 350 MW). In the Federal Republik of Germany, this step is taken belatedly in building the SNR-300. BN-350 in the USSR, Phenix in France, and PFR in the United Kingdom have now been in operation for some ten years. Over that period, valuable experience has been accumulated in sodium technology. The operating behavior of all components and systems working in sodium is called excellent; the hazards associated with sodium, the fire hazard in particular, thus often seem to be greatly overrated. Leakages have been brought under control. It has always been possible so far to trace them back to systemic faults produced in the welding process. The ability of fast sodium cooled reactors to produce more nuclear fuel than they consume has been demonstrated in Phenix, whose breeding ration has been measured to be 1.16. The first true large breeder, Super Phenix in France, is to be commissioned already in 1985. In building another three breeder power plants the European partners in an association hope to achieve the commercial breakthrough of the breeder line. (orig.)

  12. Two-phase flow in the upper plenum of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    The end part of the Emergency Core Spray System (ECSS) of the Boiling Water Reactors (BWRs) at Forsmark Nuclear Power Plant (NPP) is situated in the Upper Plenum. It consists of a pipe network equipped with water injection nozzles. In case of Lost-of-Coolant Accidents (LOCAs), the ECSS should maintain the core covered by water and, at the same time, rapidly cool and decompress the reactor by means of cold water injection. In similar reactors, some welds belonging to the ECSS support have, after a period of time, shown crack indications. Inspection, repair or replacement of these welds is time consuming and expensive. For this reason, it has now been decided to permanently remove the end part of the ECSS and to replace it by water injection in the Downcomer. However, this removal should not be accompanied by undesirable effects like an increase in the moisture of the steam used for operating the turbines. To investigate the effect of this removal on the steam moisture, a CFD analysis of the two-phase flow in the Upper Plenum of Unit 3, with and without ECSS, has been carried out by means of a two-phase Euler model in FLUENT 6.0. The inlet conditions are given by an analysis of the core kinetics and thermal hydraulics by mean of the POLCA-code. The outlet conditions, i. e. the steam separator pressure drops, are given by empirical correlations from the experiments carried out at the SNORRE facility. The predicted the mass flow-rates to each separator, together with empirical correlations for the moisture content of the steam leaving the separators and the steam dryer, indicate a slight decrease in the steam moisture when the ECSS is removed. Also, a minor decrease in pressure losses over the Upper Plenum is achieved with this removal. On the other hand, rounding the sharp edges of the inlet openings to the steam separators at the shroud cover may give a large reduction in pressure losses

  13. Investigation on a corrosion product deposit layer on a boiling water reactor fuel cladding

    International Nuclear Information System (INIS)

    Recent investigations on the complex corrosion product deposits on a boiling water reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The magnetic behaviour of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements' oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main phases identified by 2D μXRD mapping inside the layer are hematite and spinel phases with the common formula MxFey(M(1-x)Fe(2-y))O4, where M = Zn, Ni, Mn. It has been shown that the solid solutions of these phases were obtained with rather large differences between the parameter cell of the known spinels (ZnFe2O4, NiFe2O4 and MnFe2O4) and the investigated material. The comparison of EPMA with μXRD analysis shows that the ratio of Fe2O3/MFe2O4 (M = Zn, Ni, Mn) phases in the lower sample equals ∼1/2 and in the higher one ∼1/1 within the analyzed volume of the samples. It has been shown that this ratio, together with the thickness of the corrosion product deposit layer, effect its magnetic properties.

  14. Boiling Water Reactor Fuel Assembly Axial Design Optimization Using Tabu Search

    International Nuclear Information System (INIS)

    In this paper the implementation of the tabu search (TS) optimization method to a boiling water reactor's (BWR's) fuel assembly (FA) axial design is described. The objective of this implementation is to test the TS method for the search of optimal FA axial designs. This implementation has been linked to the reactor core simulator CM-PRESTO in order to evaluate each design proposed in a reactor cycle operation. The evaluation of the proposed fuel designs takes into account the most important safety limits included in a BWR in-core analysis based on the Haling principle. Results obtained show that TS is a promising method for solving the axial design problem. However, it merits further study in order to find better adaptation of the TS method for the specific problem

  15. The D and D of the Experimental Boiling Water Reactor (EBWR)

    International Nuclear Information System (INIS)

    Argonne National Laboratory has completed the D ampersand D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D ampersand D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort

  16. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik [Nuclear Physics and Biophysics Research Division Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

  17. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    International Nuclear Information System (INIS)

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m

  18. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    Science.gov (United States)

    Trianti, Nuri; Nurjanah, Su'ud, Zaki; Arif, Idam; Permana, Sidik

    2015-09-01

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid's temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

  19. Boiling in variable gravity under the action of an electric field: results of parabolic flight experiments

    International Nuclear Information System (INIS)

    Results from the variable gravity pool boiling experiments performed during the 52nd ESA parabolic flight campaign are reported in this paper. During a typical parabola, the gravity acceleration changes from 1.8gE (high gravity) to ∼0gE (low gravity) and finally back to 1.8gE. The two high gravity periods and the microgravity period are each roughly maintained for 20 seconds while the transition from high gravity to low gravity and vice versa occurs over a period of 3–5 seconds. Use of the high feedback frequency microheater array allowed quasi-steady boiling data over the continuous range of gravity levels (0gE-1.8gE). The experimental apparatus consisted of a boiling chamber with a 7×7 mm2 microheater array in a 10×10 configuration. Each heater in the array was individually controlled to maintain a constant temperature. The array could be operated in a full configuration or a selectively powered reduced set of 3×3 heaters. Experiments were performed with FC-72 as the test fluid, the pressure was maintained at a constant value between 1 and 1.13 atm and the subcooling ranged from 27 to 11 K. An external electric field was imposed over the boiling surface by means of a grid consisting of 4 rods, laid parallel to the surface; voltages up to 10 kV were applied. The electric field was effective in reducing the size of the detaching bubbles, and increasing the heat transfer compared to the values in low-g, although its effectiveness decayed as the heat flux/superheat increased. The current results compared well with previous results obtained in the ARIEL apparatus that was operated in orbital flight.

  20. Boiling in variable gravity under the action of an electric field: results of parabolic flight experiments

    Science.gov (United States)

    Di Marco, P.; Raj, R.; Kim, J.

    2011-12-01

    Results from the variable gravity pool boiling experiments performed during the 52nd ESA parabolic flight campaign are reported in this paper. During a typical parabola, the gravity acceleration changes from 1.8gE (high gravity) to ~0gE (low gravity) and finally back to 1.8gE. The two high gravity periods and the microgravity period are each roughly maintained for 20 seconds while the transition from high gravity to low gravity and vice versa occurs over a period of 3-5 seconds. Use of the high feedback frequency microheater array allowed quasi-steady boiling data over the continuous range of gravity levels (0gE-1.8gE). The experimental apparatus consisted of a boiling chamber with a 7×7 mm2 microheater array in a 10×10 configuration. Each heater in the array was individually controlled to maintain a constant temperature. The array could be operated in a full configuration or a selectively powered reduced set of 3×3 heaters. Experiments were performed with FC-72 as the test fluid, the pressure was maintained at a constant value between 1 and 1.13 atm and the subcooling ranged from 27 to 11 K. An external electric field was imposed over the boiling surface by means of a grid consisting of 4 rods, laid parallel to the surface; voltages up to 10 kV were applied. The electric field was effective in reducing the size of the detaching bubbles, and increasing the heat transfer compared to the values in low-g, although its effectiveness decayed as the heat flux/superheat increased. The current results compared well with previous results obtained in the ARIEL apparatus that was operated in orbital flight.

  1. Cycle studies: material balance estimation in the domain of pressurized water and boiling water reactors. Experimental qualification

    International Nuclear Information System (INIS)

    This study is concerned with the physics of the fuel cycle the aim being to develop and make recommendations concerning schemes for calculating the neutronics of light water reactor fuel cycles. A preliminary study carried out using the old fuel cycle calculation scheme APOLLO1- KAFKA and the library SERMA79 has shown that for the compositions of totally dissolved assemblies from Pressurized Water Reactors (type 17*17) and also for the first time, for Boiling Water Reactor assemblies (type 8*8), the differences between calculation and measurement are large and must be reduced. The integration of the APOLLO2 neutronics code into the fuel cycle calculation scheme improves the results because it can model the situation more precisely. A comparison between APOLLO1 and APOLLO2 using the same options, demonstrated the consistency of the two methods for PWR and BWR geometries. Following this comparison, we developed an optimised scheme for PWR applications using the library CEA86 and the code APOLLO2. Depending on whether the information required is the detailed distribution of the composition of the irradiated fuel or the average composition (estimation of the total material balance of the fuel assembly), the physics options recommended are different. We show that the use of APOLLO2 and the library CEA86 improves the results and especially the estimation of the Pu239 content. Concerning the Boiling Water Reactor, we have highlighted the need to treat several axial sections of the fuel assembly (variation of the void-fraction, heterogeneity of composition). A scheme using Sn transport theory, permits one to obtain a better coherence between the consumption of U235, the production of plutonium and burnup, and a better estimation of the material balance. (author)

  2. Risk based ISI application to a boiling water reactor

    International Nuclear Information System (INIS)

    The ASME Section XI Working Group on Implementation of Risk-Based Examination produced a code case to define risk-based selection rules that could be used for In-Service Inspection (ISI) of Class 1, 2, and 3 piping. To provide guidelines for practical implementation of the code case, EPRI sponsored work to develop evaluation procedures and criteria. As part of an EPRI sponsored pilot study, these procedures have been applied to a BWR plant. Piping within the scope of the existing Section XI program has been analyzed. The results of this effort indicate that implementation of RBISI programs can significantly reduce the cost and radiation exposure associated with in-service inspections. The revised program was compared to the previous program and a net gain in safety benefit was demonstrated

  3. Ion chromatography campaigns at several boiling water reactors

    International Nuclear Information System (INIS)

    Water chemistry characterization campaigns using on-line ion chromatography were conducted at several BWRs during 1989. Some of the highlights from these campaigns, along with equipment and IC methods are presented in this paper. Monitoring copper ion in final feedwater and filter demineralizer effluents at levels <0.2 ppb enabled optimum operation of the condensate demineralizer system. The need for new precoats was based on ion trends, not pressure drop or volume limitations. Conductivity transients, due to power reductions, were characterized. Magnesium, calcium, sulfate and chromate were the major contributors. Start up, water chemistry transients, were monitored, revealing organics from sealing compounds used in the condenser and turbine. On-line IC is an effective tool for understanding power plant chemistry and improving operations of condensate cleanup systems

  4. Parametrisation for optimisation of reload patterns for boiling water reactors

    International Nuclear Information System (INIS)

    Parametrisation of reload patterns for BWRs is attempted for the purpose of optimising them. This is done in two stages. The first stage involves an algorithm for constitution of a reload pattern out of a given set of fuel bundles. It is designed to constitute patterns with features of Low Leakage Loading (LLL) and checker board arrangement of fresh and exposed fuel bundles in the central region. The characteristics of the Reload Patterns can be manipulated by means of only two input parameters. The idea of the dependence the Haling power peaking and the cycle energy of the ''Biparametric Reload Patterns (BRP)'' on the two input parameters is presented through case studies. The idea of the ranges of these characteristic quantities and their mutual relation is given by the case studies both for general as well as the optimum BRPs. A preliminary optimization procedure for biparametric reload patterns is presented. This method is augmented by a multi-parametric algorithm to reshuffling the radial exposure distribution interactively so as to exhaust any possibility of improvement in a given reload pattern. When tested against this procedure, the optimum biparametric reload patterns are seen to provide only a limited scope for improvement showing that they are very close to the optimum reload patterns. The extent of possible improvement is illustrated. The entire procedure is incorporated in a 2-dimensional code CORECOOK. The computer time needed for this task of optimisation of reload pattern is comparable to that required for one routine 3-D core follow-up calculation. This algorithm has been used for obtaining the reload patterns for all the five cores loaded in BWRs at Tarapur (India) since 1988. (author). 25 refs, 5 figs, 2 tabs

  5. Parametrization for optimization of reload patterns for boiling water reactors

    International Nuclear Information System (INIS)

    Parametrization of the reload patterns for BWRs is attempted with the aim of optimizing them. This is done in two stages. The first involves an algorithm for the construction of a reload pattern out of a given set of fuel bundles. It is designed to construct patterns which feature low leakage loading (LLL) and a chequerboard arrangement of fresh and exposed fuel bundles in the central region. The characteristics of the reload patterns can be manipulated by means of only two input parameters. The dependence of the Haling power peaking and the cycle energy of the ''biparametric reload patterns (BRP)'' on the two input parameters is shown through case studies. The ranges of these characteristic quantities and their mutual relation are given by case studies for both the general as well as the optimum BRPs. A preliminary optimization procedure for BRPs is presented. This method is augmented by a multi-parametric algorithm to reshuffle the radial exposure distribution interactively so as to exhaust any possibility of improvement in a given reload pattern. When tested against this procedure, the optimum BRPs are seen to have only limited scope for improvement, showing that they are very close to the optimum reload pattern. The extent of possible improvement is illustrated. The entire procedure is incorporated in a 2-D code CORECOOK. The computer time needed for optimization of a reload pattern is comparable to that required for one routine 3-D core followup calculation. This algorithm has been used for obtaining the reload patterns for all five cores in the BWRs at Tarapur (India) since 1988. (Author)

  6. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  7. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    International Nuclear Information System (INIS)

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  8. Power distribution control within the scope of the advanced nuclear predictor for boiling water reactors

    International Nuclear Information System (INIS)

    In boiling water reactors the Advanced Nuclear Predictor (FNR) has proved to be a valuable tool in improving plant operating efficiency. The system is described in its main features and capabilities. As a logical extension, a power distribution control system has been developed, based on a reduced but accurate core model, which in itself can be used for fast prediction of core states. The system provides prediction of optimal operating strategies as well as on-line control, observing all constraints imposed on the permissible operating region. (orig.)

  9. Present status of maintenance technologies for boiling-water-reactor power plants

    International Nuclear Information System (INIS)

    Toshiba places the highest priority on maintenance technologies for boiling-water-reactor (BWR) power plants. These activities are based on our motto, 'Ensuring stable operation of BWRs throughout the plant life cycle'. A quarter of a century has passed since the construction of the first such plant in which Toshiba was involved, and preventive maintenance is therefore a matter of great importance for BWRs. This paper presents an overview of plant monitoring and diagnosis, preventive maintenance of equipment, and ensuring the high quality of plant improvement or annual inspection work. (author)

  10. Nuclear power plant with boiling water reactor VK-300 for district heating and electricity supply

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, Y.N.; Lisitza, F.D.; Romenkov, A.A.; Tokarev, Y.I. [RDIPE, Moscow (Russian Federation)

    1998-07-01

    The paper considers specific design features of a pressure vessel boiling water reactor with coolant natural circulation and three-step in-vessel steam separation (at draught tube outlet of the upcomer, within zone of overflow from the upcomer to downcomer and in cyclon-type separators). Design description and analytical study results are presented for the passive core cooling system in the case of loss of preferred power and rupture in primary circuit pipeline. Specific features of a primary containment (safeguard vessel) are given for an underground NPP sited in a rock ground. (author)

  11. Calculations of severe accident progression in the General Electric Simplified Boiling Water Reactor

    International Nuclear Information System (INIS)

    General Electric is designing a new nuclear power plant: the Simplified Boiling Water Reactor (SBWR). The SBWR is a passive plant in which the core cooling and decay heat removal safety systems are driven by gravity. To model the plant response to severe accidents, MAAP-SBWR, an advanced version of the Modular Accident Analysis Program (MAAP), has been developed. The main feature of the new code is a flexible containment model. The challenges in modeling the SBWR, the code structure and models, and a sample application to the SBWR are discussed

  12. Uncertainty calculation of emergency core cooling system for boiling water reactor (BWR-5)

    International Nuclear Information System (INIS)

    A brief description about uncertainty calculation of emergency core cooling system for boiling water reactor (BWR-5) is presented in this paper. Based on methodology of PSA level 1 and draft description of ECCS's document supplied by TOSHIBA (Code PSO-00-00097, July 2000) the event tree is built. The fault trees of three of subsystems HPCSS, LPCSS, LPCIS can be developed due to the simplified P and ID of ECCS and the reliability data accompanied. The computer code used to develop fault tree is KIRAP-tree and one used to find cut set and calculated uncertainty is KCUT. (author)

  13. A parametric analysis of decay ratio calculations in a boiling water reactor model

    International Nuclear Information System (INIS)

    The results of an investigation of the effects of several parameters on the reactivity instability of a Boiling Water Reactor (BWR) calculational model are summarized. Calculations were performed for a typical BWR operated at low flow conditions, where reactivity instabilities are more likely to occur. The parameters investigated include the axial power shape (characterized by two separate parameters), the core pressure, and operating flow. All calculations were performed using the LAPUR code which was developed at the Oak Ridge National Laboratory for the dynamic modeling of large BWR's. 4 refs., 8 figs

  14. Results of Reactor Materials Experiments Investigating 2-D Core-Concrete Interaction and Debris Coolability

    International Nuclear Information System (INIS)

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor materials experiments and associated analysis to achieve the following objectives: 1) resolution of the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs of future plants. With respect to the second objective, there remain uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a first step in bridging this gap, a large scale Core-Concrete Interaction experiment (CCI-1) has been conducted as part of the MCCI program. This test investigated the interaction of a 400 kg core-oxide melt with a crucible made of siliceous concrete along two walls and the base. The two remaining walls were made of non-ablative magnesium oxide. The initial phase of the test was conducted under dry conditions. After a predefined ablation depth was achieved, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the test facility and an overview of results from this test. (authors)

  15. Application of reliability techniques to prioritize BWR [boiling water reactor] recirculation loop welds for in-service inspection

    International Nuclear Information System (INIS)

    In January 1988 the US Nuclear Regulatory Commission issued Generic Letter 88-01 together with NUREG-0313, Revision 2, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' to implement NRC long-range plans for addressing the problem of stress corrosion cracking in boiling water reactor piping. NUREG-0313 presents guidelines for categorizing BWR pipe welds according to their SCC condition (e.g., presence of known cracks, implementation of measures for mitigating SCC) as well as recommended inspection schedules (e.g., percentage of welds inspected, inspection frequency) for each weld category. NUREG-0313 does not, however, specify individual welds to be inspected. To address this issue, the Lawrence Livermore National Laboratory developed two recommended inspection samples for welds in a typical BWR recirculation loop. Using a probabilistic fracture mechanics model, LLNL prioritized loop welds on the basis of estimated leak probabilities. The results of this evaluation indicate that riser welds and bypass welds should be given priority attention over other welds. Larger-diameter welds as a group can be considered of secondary importance compared to riser and bypass welds. A ''blind'' comparison between the probability-based inspection samples and data from actual field inspections indicated that the probabilistic analysis generally captured the welds which the field inspections identified as warranting repair or replacement. Discrepancies between the field data and the analytic results can likely be attributed to simplifying assumptions made in the analysis. The overall agreement between analysis and field experience suggests that reliability techniques -- when combined with historical experience -- represent a sound technical basis on which to define meaningful weld inspection programs. 13 refs., 8 figs., 5 tabs

  16. Safety System Design Concept and Performance Evaluation for a Long Operating Cycle Simplified Boiling Water Reactor

    International Nuclear Information System (INIS)

    The long operating cycle simplified boiling water reactor is a reactor concept that pursues both safety and the economy by employing a natural circulation reactor core without a refueling, a passive decay heat removal, and an integrated building for the reactor and turbine. Throughout the entire spectrum of the design basis accident, the reactor core is kept covered by the passive emergency core cooling system. The decay heat is removed by the conventional active low-pressure residual heat removal system. As for a postulated severe accident, the suppression pool water floods the lower part of the reactor pressure vessel (RPV) in the case when core damage occurs, and the in-vessel retention that keeps the melt inside the RPV is achieved by supplying the coolant. The containment adopts a parallel-double-steel-plate structure similar to a hull structure, which contains coolant between the inner and outer walls to absorb the heat transferred from the inside of the containment. Consequently, the containment structure functions as a passive containment cooling system (PCCS) to remove the decay heat in case of an accident. This paper describes the PCCS performance evaluation by using TRAC code to show one of the characteristic plant features. The core damage frequency for internal events was also evaluated to examine the safety level of the plant and to show the adequacy of the safety system design

  17. Subcooled film boiling heat transfer on a high temperature sphere in very dilute Al{sub 2}O{sub 3} nano-fluids

    Energy Technology Data Exchange (ETDEWEB)

    Hyun Sun Park; Dereje Shiferaw; Bal Raj Sehgal [Royal Institute of Technology, Department of Energy Technology, Division of Nuclear Power Safety, Drotning Kristinas Vag 33A, SE-10044, Stockholm (Sweden)

    2005-07-01

    Full text of publication follows: nano-fluids, or conventional liquids, e.g., water, with small concentration of nano-particles uniformly suspended, have attracted attention as a new heat transport medium with enhanced thermo-physical properties. Up to the present, only exploratory experiments on nano-fluids have been reported. Das et al (Int. J. Heat Mass Transfer 43, pp 3701-3707, 2003) conducted boiling experiments with water containing 38 nm Al{sub 2}O{sub 3} nano-particles. They observed deterioration in the nucleate boiling heat transfer due to the deposition of nano-particles. Boiling experiments conducted by Vassallo et al (Int. J. Heat Mass Transfer 47, pp 407-411, 2004) using silica nano-fluid using 0.4 mm diameter NiCr wire showed three times higher critical heat flux (CHF) and the wire traversed the film boiling region before it failed. Another independent experiment performed on 1 cm{sup 2} square plate with a very low concentration of nano-particles ranging from 0.01 to 0.05 g/liter and at under pressure (2.89 psia), nano-fluids resulted in drastic 2{approx}3 times enhancement of the CHF (You and Kim, Appl. Phys. Lett. 83. No 16, 2003). However in all the aforementioned studies no appropriate explanation of the CHF enhancement has been advanced. The measured 2-3 times higher critical heat flux for very dilute nano-fluids may have high significance if such nano-fluids could be employed in heat transport systems. Recently, we investigated the effect of nano-particles on film boiling, which governs heat transfer during accident conditions in a reactor plant, e.g., in coolability of a degraded core, or a particulate debris bed or a core melt, and in steam explosions. Our previous experiments performed on film boiling in nano-fluids having larger concentrations of 5, 10, and 20 g/liter than those in You's experiments showed that the nano-fluids lower the film boiling temperature, decrease the film boiling heat transfer and provide a much thicker and

  18. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    International Nuclear Information System (INIS)

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

  19. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1

    International Nuclear Information System (INIS)

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17th, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  20. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O2; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  1. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  2. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  3. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  4. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  5. IBR-2 reactor power start-up and first physical experiments on its beams

    International Nuclear Information System (INIS)

    The design is described of the IBR-2 pulse fast reactor designated for investigations into nuclear physics, condensed medium physics, molecular biology, elementary particle physics (fundamental properties of the neutron) as well as for solving various applied problems. The IBR-2 core is loaded with plutonium dioxide fuel of about 90 kg total mass. The fuel elements are sodium, cooled with a 300 deg C inlet temperature. The cooling system is two-circuit, two-loop with the 100 m3/h sodium flow rate. The reactor is surrounded by water neutron moderators which are ''scanned'' by 14 horizontal channels. The power pulse is formed by the reactivity modulator in the form of two coaxially positioned mobile neutron reflectors - the basic one (BR) and additional one (AR). The BR rotation frequency equals 1500 min-1. The pulse frequency is varied discretely by the AR, being at rest or rotating at a lower rate. In 1982 the IBR-2 attained the average power 2 MW at a BR rotation frequency equal to 1500 mm-1 and a pulse frequency equal to 25 Hz, which corresponds to pulse reactor power of 270 MW. The reactor main frequency mode 5 Hz at 2 MW power was realized in 1984, the reactor peak power achieved 1350 MW. Results of first physical studies conducted during the reactor power start-up are presented

  6. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    International Nuclear Information System (INIS)

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  7. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Manglik, R M; Athavale, A; Kalaikadal, D S; Deodhar, A; Verma, U

    2011-09-02

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  8. Final environmental statement for La Crosse Boiling Water Reactor: (Docket No. 50-409)

    International Nuclear Information System (INIS)

    A Final Environmental Statement for the Dairyland Power Cooperative for the conversion from a provisional to a full-term operating license for the La Crosse Boiling Water Reactor, located in Vernon County, Wisconsin, has been prepared by the Office of Nuclear Reactor Regulation. This statement provides a summary of environmental impacts and adverse effects of operation of the facility, and a consideration of principal alternatives (including removal of LACBWR from service, alternative cooling methodology, and alternative waste treatment systems). Also included are the comments of federal, state, and local governmental agencies and certain non-governmental organizations on the La Crosse Draft Environmental Statement and staff responses to these comments. After weighing environmental, economic, and technical benefits and liabilities, the staff recommends conversion from a provisional operating license to a full-term operating license, subject to specific environmental protection limitations. An operational monitoring program shall be established as part of the Environmental Technical Specifications. 64 refs., 20 figs., 48 tabs

  9. Analysis of boiling water reactors capacities for the 100% MOX fuel recycling

    International Nuclear Information System (INIS)

    The electro-nuclear park exploitation leads to plutonium production. The plutonium recycling in boiling water reactors performs a use possibility. The difference between the neutronic characteristics of the uranium and the plutonium need to evaluate the substitution impact of UOX fuel by MOX fuel on the reactor operating and safety. The analysis of the main points reached to the following conclusions: the reactivity coefficients are negative, during a cooling accident the re-divergence depends on the isotopic vector of the used plutonium, the efficiency lost of control cross resulting from the plutonium utilization can be compensate by the increase of the B 4C enrichment by 10B and the change of the steel structure by an hafnium structure, the reactivity control in evolution can be obtained by the fuel poisoning (gadolinium, erbium) and the power map control by the plutonium content monitoring. (A.L.B.)

  10. Optimal axial enrichment distribution of the boiling water reactor fuel under the Haling strategy

    International Nuclear Information System (INIS)

    The axial enrichment distribution of boiling water reactor fuel is optimized to improve uranium utilization subject to constraints on thermal margins. It is assumed that the reactor is operated under the Haling strategy, so that determination of the enrichment distribution can be decoupled from the poison management. This nonlinear optimization problem is solved using a method of approximation programming, where each iteration step is formulated in terms of linear goal programming to handle infeasible problems. The core is represented by an axial one-dimensional model. The average enrichment of a two-region fuel can be slightly reduced by increasing the enrichment of the lower half rather than the upper half. The optimal solutions for a 24-region fuel, in which the enrichments of indivdual nodes can differ from one another, display double-humped enrichment distributions. The natural uranium blanket design is also investigated, and it is concluded that blanketed fuel is practically optimal using the Haling strategy

  11. TRAC-BD1: transient reactor analysis code for boiling-water systems

    Energy Technology Data Exchange (ETDEWEB)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented.

  12. Reducing radiation levels at boiling water reactors of a commercial nuclear power plant fleet

    International Nuclear Information System (INIS)

    Boiling Water Reactors (BWRs) have suffered from high radiation fields in the primary loop, typically measured by the 'BRAC' (BWR Radiation Level Assessment and Control) reactor recirculation system (RRS) dose rates. Reactor water chemistry and activated corrosion product measurements are important in understanding changes in radiation fields in components and systems of a BWR. Several studies have been conducted at Exelon Nuclear's 14 BWRs in order to understand more fully the cause and effect relationships between reactor water radioactive species and radiation levels. Various radiation control strategies are utilized to control and reduce radiation levels. The proper measurement of radioactive soluble and insoluble species is a critical component in understanding radiation fields. Other factors that impact radiation fields include: noble metal applications; hydrogen injection; zinc addition; chemistry results; cobalt source term; fuel design and operation. Chemistry and radiation field trending and projections are important tools that assist in assessing the potential for increased radiation fields and aiding outage planning efforts, including techniques to minimize outage dose. This paper will present the findings from various studies and predictor tools as well as provide recommendations for continued research efforts in this field. Current plant data will be shared on reactor water radioactive species, plant radiation levels, zinc addition amounts and other chemistry controls. (author)

  13. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE-ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

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

    International Nuclear Information System (INIS)

    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)

  15. Numerical simulation and artificial neural network modeling of natural circulation boiling water reactor

    International Nuclear Information System (INIS)

    Numerical simulation of natural circulation boiling water reactor is important in order to study its performance for different designs and under various off-design conditions. Numerical simulations can be performed by using thermal-hydraulic codes. Very fast numerical simulations, useful for extensive parametric studies and for solving design optimization problems, can be achieved by using an artificial neural network (ANN) model of the system. In the present work, numerical simulations of natural circulation boiling water reactor have been performed with RELAP5 code for different values of design parameters and operational conditions. Parametric trends observed have been discussed. The data obtained from these simulations have been used to train artificial neural networks, which in turn have been used for further parametric studies and design optimization. The ANN models showed error within ±5% for all the simulated data. Two most popular methods, multilayer perceptron (MLP) and radial basis function (RBF) networks, have been used for the training of ANN model. Sequential quadratic programming (SQP) has been used for optimization

  16. Reactor antineutrino experiments

    CERN Document Server

    Lu, Haoqi

    2014-01-01

    Neutrinos are elementary particles in the standard model of particle physics. There are 3 flavors of neutrinos that oscillate among themselves. Their oscillation can be described by a 3$\\times$3 unitary matrix, containing three mixing angles $\\theta_{12}$, $\\theta_{23}$, $\\theta_{13}$, and one CP phase. Both $\\theta_{12}$ and $\\theta_{23}$ are known from previous experiments. $\\theta_{13}$ was unknown just two years ago. The Daya Bay experiment gave the first definitive non-zero value in 2012. An improved measurement of the oscillation amplitude $\\sin^{2}2(\\theta_{13})$ = $0.090^{+0.008}_{-0.009}$ and the first direct measurement of the $\\bar\

  17. Prediction of void fraction in subcooled flow boiling

    International Nuclear Information System (INIS)

    The information on heat transfer and especially on the void fraction in the reactor core under subcooled conditions is very important for the water-cooled nuclear reactors, because of its influence upon the reactivity of the systems. This paper gives a short overview of subcooled boiling phenomenon and indicates the simplifications made by the RELAP5 model of subcooled boiling. RELAP5/MOD3.2 calculations were compared with simple one-dimensional models and with high-pressure Bartolomey experiments.(author)

  18. Los Alamos experiments and their impacts on fast reactor safety

    International Nuclear Information System (INIS)

    Results of two sets of recent Los Alamos transition-phase experiments are reported herein. The two sets of experiments addressed two different behaviors of boiling pools of molten fuel, molten steel and steel vapor, in the transition phase of a core-disruptive accident (CDA) in a liquid-metal fast breeder reactor (LMFBR). The transient boilup experiments simulated the recriticality-induced motions of a boiling pool within a single subassembly during the subassembly-pool subphase of the transition phase. The melting wall experiments simulated the melting and entrainment of subassembly duct wall steel into a boiling pool during the same subphase. From the results of the transient boilup experiment we identified behaviors and phenomena that argue against an energetic disassembly from the subassembly-pool subphase. From the melting wall experiments we determined that a stable boiling pool is unlikely by showing that significant amounts of wall steel would likely be rapidly entrained and lead to pool collapse. 8 refs., 3 figs

  19. Corrosion product deposits on boiling-water reactor cladding: Experimental and theoretical investigation of magnetic properties

    Science.gov (United States)

    Orlov, A.; Degueldre, C.; Wiese, H.; Ledergerber, G.; Valizadeh, S.

    2011-09-01

    Recent Eddy current investigations on the cladding of nuclear fuel pins have shown that the apparent oxide layers are falsified due to unexpected magnetic properties of corrosion product deposits. Analyses by Scanning Electron Microscopy (SEM) or Electron Probe Micro Analysis (EPMA) demonstrated that the deposit layer consists of complex 3-d element oxides (Ni, Mn, Fe) along with Zn, since the reactor operates with a Zn addition procedure to reduce buildup of radiation fields on the recirculation system surfaces. The oxides crystallise in ferritic spinel structures. These spinels are well-known for their magnetic behaviour. Since non-magnetic zinc ferrite (ZnFe 2O 4) may become magnetic when doped with even small amounts of Ni and/or Mn, their occurrence in the deposit layer has been analyzed. The magnetic permeability of zinc ferrite, trevorite and jacobsite and their solid solutions are estimated by magnetic moment additivity. From the void history examination, the low elevation sample (810 mm) did not face significant boiling during the irradiation cycles suggesting growth of (Mn0.092+Zn0.752+Fe0.293+)[(Fe1.713+Mn0.032+Ni0.132+)O] crystals with theoretical value of the magnetic permeability for the averaged heterogeneous CRUD layer of 9.5 ± 3. Meanwhile, (Mn0.162+Zn0.552+Fe0.293+)[(Fe1.713+Mn0.042+Ni0.252+)O] crystallizes at the mid elevation (1810 mm) with theoretical magnetic permeability for the CRUD layer of 4.2 ± 1.5 at the investigated azimuthal location. These theoretical data are compared with the magnetic permeability of the corrosion product deposited layers gained from reactor pool side Eddy current (EC) analyses (9.0 ± 1.0 for low and 3.5 ± 1.0 for high elevation). The calculated thicknesses and magnetic permeability values of the deposition layers (estimated by MAGNACROX multifrequency EC method) match together with these estimated using an "ion magnetic moment additivity" model.

  20. Corrosion product deposits on boiling-water reactor cladding: Experimental and theoretical investigation of magnetic properties

    International Nuclear Information System (INIS)

    Recent Eddy current investigations on the cladding of nuclear fuel pins have shown that the apparent oxide layers are falsified due to unexpected magnetic properties of corrosion product deposits. Analyses by Scanning Electron Microscopy (SEM) or Electron Probe Micro Analysis (EPMA) demonstrated that the deposit layer consists of complex 3-d element oxides (Ni, Mn, Fe) along with Zn, since the reactor operates with a Zn addition procedure to reduce buildup of radiation fields on the recirculation system surfaces. The oxides crystallise in ferritic spinel structures. These spinels are well-known for their magnetic behaviour. Since non-magnetic zinc ferrite (ZnFe2O4) may become magnetic when doped with even small amounts of Ni and/or Mn, their occurrence in the deposit layer has been analyzed. The magnetic permeability of zinc ferrite, trevorite and jacobsite and their solid solutions are estimated by magnetic moment additivity. From the void history examination, the low elevation sample (810 mm) did not face significant boiling during the irradiation cycles suggesting growth of (Mn0.092+Zn0.752+Fe0.293+)[(Fe1.713+Mn0.032+Ni0.132+)O4] crystals with theoretical value of the magnetic permeability for the averaged heterogeneous CRUD layer of 9.5 ± 3. Meanwhile, (Mn0.162+Zn0.552+Fe0.293+)[(Fe1.713+Mn0.042+Ni0.252+)O4] crystallizes at the mid elevation (1810 mm) with theoretical magnetic permeability for the CRUD layer of 4.2 ± 1.5 at the investigated azimuthal location. These theoretical data are compared with the magnetic permeability of the corrosion product deposited layers gained from reactor pool side Eddy current (EC) analyses (9.0 ± 1.0 for low and 3.5 ± 1.0 for high elevation). The calculated thicknesses and magnetic permeability values of the deposition layers (estimated by MAGNACROX multifrequency EC method) match together with these estimated using an 'ion magnetic moment additivity' model.

  1. The effectiveness of early hydrogen water chemistry on corrosion mitigation for boiling water reactors

    International Nuclear Information System (INIS)

    For mitigating intergranular stress corrosion cracking (IGSCC) in an operating boiling water reactor (BWR), the technology of hydrogen water chemistry (HWC) aiming at coolant chemistry improvement has been adopted worldwide. However, the hydrogen injection system is usually in an idle and standby mode during a startup operation. The coolant in a BWR during a cold shutdown normally contains a relatively high level of dissolved oxygen from intrusion of atmospheric air. Accordingly, the structural materials in the primary coolant circuit (PCC) of a BWR could be exposed to a strongly oxidizing environment for a short period of time during a subsequent startup operation. At some plants, the feasibility of hydrogen water chemistry during startup operations has been studied. It is technically difficult to directly procure water chemistry data at various locations of an operating reactor. Accordingly, the impact of startup operation on water chemistry in the PCC of a BWR operating under HWC can only be theoretically evaluated through computer modelling. In this study, a well-developed computer code DEMACE was used to investigate the variations in redox species concentration and in electrochemical corrosion potential (ECP) of components in the PCC of a domestic BWR during startup operations in the presence of HWC. Simulations were carried out for [H2]FWs ranging from 0.0 to 2.0 parts per million (ppm) and for power levels ranging from 3.8% to 11.3% during startup operations. Our analyses indicated that for power levels with steam generation in the core, a higher power level would tend to promote a more oxidizing coolant environment for the structural components and therefore lead to less HWC effectiveness on ECP reduction and corrosion mitigation. At comparatively lower power levels in the absence of steam, the effectiveness of HWC on ECP reduction was much better. The effectiveness of HWC in the PCC of a BWR during startup operations is expected to vary from location to

  2. Extracted neutron beams experimental facilities and program of the first experiments at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Structural specific features of the IBR-2 pulse research biological hidered reactor. The characteristics of spectrometer for investigating the small angle neutron scattering and the CORA facility intended for investigating the structure and dynamics of condensed media by means of the therrol neutron scattering as well as the DN-2 diffractometer for investigating the atomic structure and crystallographic characteristics of monocrystals, having large (>10 A) elementary cell size and the equipment of the ultracold neutron channel are given. Biological shields of the reactor and experimental facilities are assembled of concrete blocks and standard building constructions and attains in the most dangerous regions 1 m. The shield ensures for the personnel a safety level of ionizing radiations and effectively shields the facilities from mutual effects caused by scattered radiation. The program of physical investigations planned at the IBR-2 continues the investigations started at the IBR-30 reactor

  3. N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

    Czech Academy of Sciences Publication Activity Database

    Obalová, L.; Jirátová, Květa; Karásková, K.; Chromčáková, Ž.

    2012-01-01

    Roč. 191, č. 1 (2012), s. 116-120. ISSN 0920-5861 R&D Projects: GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : N2O * catalytic decomposition * fixed bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.980, year: 2012

  4. Implementation of a source term control program in a mature boiling water reactor

    International Nuclear Information System (INIS)

    The implementation and results of a source term control program at the James A. FitzPatrick Nuclear Power Plant (JAF), a mature boiling water reactor (BWR) facility that has been in commercial operation since 1975, are discussed. Following a chemical decontamination of the reactor water recirculation piping in the Reload 8/Cycle 9 refueling outage in 1988, hydrogen water chemistry (HWC) and feedwater Zn addition were implemented. This is the first application of both HWC and feedwater Zn addition in a BWR facility. The radiological benefits and impacts of combined operation of HWC and feedwater Zn addition at JAF during Cycle 9 are detailed and summarized. The implementation of hydrogen water chemistry resulted in a significant transport of corrosion products within the reactor coolant system that was greater than anticipated. Feedwater Zn addition appears to be effective in controlling buildup of other activated corrosion products such as 60Co on reactor water recirculation piping; however, adverse impacts were encountered. The major adverse impact of feedwater Zn addition is the production of 65Zn that is released during plant outages and operational transients

  5. Implementation of a source term control program in a mature boiling water reactor.

    Science.gov (United States)

    Vargo, G J; Jarvis, A J; Remark, J F

    1991-06-01

    The implementation and results of a source term control program implemented at the James A. FitzPatrick Nuclear Power Plant (JAF), a mature boiling water reactor (BWR) facility that has been in commercial operation since 1975, are discussed. Following a chemical decontamination of the reactor water recirculation piping in the Reload 8/Cycle 9 refueling outage in 1988, hydrogen water chemistry (HWC) and feedwater Zn addition were implemented. This is the first application of both HWC and feedwater Zn addition in a BWR facility. The radiological benefits and impacts of combined operation of HWC and feedwater Zn addition at JAF during Cycle 9 are detailed and summarized. The implementation of hydrogen water chemistry resulted in a significant transport of corrosion products within the reactor coolant system that was greater than anticipated. Feedwater Zn addition appears to be effective in controlling buildup of other activated corrosion products such as 60Co on reactor water recirculation piping; however, adverse impacts were encountered. The major adverse impact of feedwater Zn addition is the production of 65Zn that is released during plant outages and operational transients. PMID:2032839

  6. Stability tests in the Grand Gulf unit 1 boiling water reactor

    International Nuclear Information System (INIS)

    This paper summarizes the results of a series of tests performed on January 31, 1987, to determine the stability of the second reload core in the Grand Gulf Unit 1 boiling water reactor (BWR). The subject of BWR stability is relevant for commercial BWR operation. Utilities are required to evaluate reactor stability for every reload core unless plant technical specifications provide for monitoring of neutron flux oscillations in the so-called limit-cycle detect and suppress region at low flows. The parameter of merit for stability calculations or measurements is the asymptotic decay ratio (DR). The definition of asymptotic DR guarantees that as long as its value is < 1.0, the reactor is stable. The DR also yields a quantitative measure of relative stability: DRs below 0.5 are considered very stable. A noise analysis technique was implemented in a portable computer system, which uses standard commercially available hardware, and was used to perform stability measurements on line. This technique has proven to be fairly accurate for high DRs, when the reactor is close to the stability threshold. For low DR conditions, however, the technique yields only reasonable accuracy. An attempt to quantify this accuracy has been made, and the resulting error bands are presented

  7. Predicted impact of power coastdown operations on the water chemistry for two domestic boiling water reactors

    International Nuclear Information System (INIS)

    A theoretical model was adapted to evaluate the impact of power coastdown on the water chemistry of two commercial boiling water reactors (BWRs) in this work. In principle, the power density of a nuclear reactor upon a power level decrease would immediately be lowered, followed by water chemistry variations due to reduced radiolysis of water and extended coolant residence times in the core and near-core regions. It is currently a common practice for commercial BWRs to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal feedwater hydrogen concentration may be different after a power coastdown is implemented in a BWR. A computer code DEMACE was used in the current study to investigate the impact of various power coastdown levels on major radiolytic species concentrations and electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of two domestic reactors operating under either normal water chemistry or HWC. Our analyses indicated that under a rated core flow rate the oxidizing species concentrations and the ECP did not vary monotonously with decreases in reactor power level at a fixed feedwater hydrogen concentration. In particular, ECP variations basically followed the patterns of hydrogen peroxide in the select regions and exhibited high values at power levels of 95% and 90% for Chinshan-1 and Kuosheng-1, respectively. (author)

  8. Phased array UT application for boiling water reactor vessel bottom head

    International Nuclear Information System (INIS)

    Stress Corrosion Cracking (SCC) on welds of reactor internals is one of the most important issues in nuclear plants since 1990's. Demands to inspect the reactor internals are increasing. This paper focuses on the development and the application of the phased array ultrasonic testing (PAUT) for the reactor internals located in Boiling Water Reactor (BWR) vessel bottom head (e.g., shroud support). The Toshiba PAUT technologies and technique has been developed and applied to in-Vessel inspections (IVIs) as our universal nondestructive testing (NDT) technologies. Though it was difficult to detect and size cracks in Alloy 182 welds (i.e. weld metal of the shroud support and a CRD stub tube), the efficiency of the PAUT techniques is shown in recent IVI activities. For example the PAUT techniques are applied to crack depth sizing in the weld between the CRD stub tube and RPV bottom build-up in recent years. An immersion technique by the PAUT enables to perform the UT examination on a complex geometric surface to be inspected. The PAUT techniques are developed to detect and size flaws on the shroud support Alloy 182 welds. The techniques include detection from the outside and the inside of RPV. These techniques are applied to the simulated shroud support mockups with SCC-simulated flaws. The examination result is proven to have a good agreement with their actual. As a result, the efficiency of the PAUT techniques is confirmed. (author)

  9. Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

    International Nuclear Information System (INIS)

    Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.

  10. A nuclear desalination complex with a VK-300 boiling type reactor facility

    International Nuclear Information System (INIS)

    RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m3/day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

  11. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  12. Dose Rates Near Water Moderator of the IBR-2 Reactor Experiment and Analysis

    CERN Document Server

    Golikov, V V; Shabalin, E P

    2002-01-01

    Adsorbed dose rates in metals and in hydrogenous materials (polyethylene and water) have been measured at the neutron beam channel No. 3 of the IBR-2 reactor just behind the light water moderator [1]. Three methods have been applied; all of them gave the comparable results, if accounting for some corrections due to nonuniformity of the irradiation field. In metals (copper) it appeared to be 0.013 W/g/MW of the reactor power with an accuracy to {\\pm}3%; in polyethylene and water - (0.090\\pm 0.009) and (0.053\\pm 0.003) W/g/MW, respectively.

  13. Dose Rates Near Water Moderator of the IBR-2 Reactor: Experiment and Analysis

    CERN Document Server

    Golikov, V V; Shabalin, E P

    2002-01-01

    Adsorbed dose rates in metals and in hydrogenous materials (polyethylene and water) have been measured at the neutron beam channel No. 3 of the IBR-2 reactor just behind the light water moderator [1]. Three methods have been applied; all of them gave the comparable results, if accounting for some corrections due to nonuniformity of the irradiation field. In metals (copper) it appeared to be 0.013 W/g/MW of the reactor power with an accuracy to {\\pm}3%; in polyethylene and water - (0.090\\pm 0.009) and (0.053\\pm 0.003) W/g/MW, respectively.

  14. Non Invasive Water Level Monitoring on Boiling Water Reactors Using Internal Gamma Radiation: Application of Soft Computing Methods

    International Nuclear Information System (INIS)

    To provide best knowledge about safety-related water level values in boiling water reactors (BWR) is essentially for operational regime. For the water level determination hydrostatic level measurement systems are almost exclusively applied, because they stand the test over many decades in conventional and nuclear power plants (NPP). Due to the steam generation especially in BWR a specific phenomenon occurs which leads to a water-steam mixture level in the reactor annular space and reactor plenum. The mixture level is a high transient non-measurable value concerning the hydrostatic water level measuring system and it significantly differs from the measured collapsed water level. In particular, during operational and accidental transient processes like fast negative pressure transients, the monitoring of these water levels is very important. In addition to the hydrostatic water level measurement system a diverse water level measurement system for BWR should be used. A real physical diversity is given by gamma radiation distribution inside and outside the reactor pressure vessel correlating with the water level. The vertical gamma radiation distribution depends on the water level, but it is also a function of the neutron flux and the coolant recirculation pump speed. For the water level monitoring, special algorithms are required. An analytical determination of the gamma radiation distribution outside the reactor pressure vessel is impossible due to the multitude of radiation of physical processes, complicated non-stationary radiation source distribution and complex geometry of fixtures. For creating suited algorithms Soft Computing methods (Fuzzy Sets Theory, Artificial Neural Networks, etc.) will be used. Therefore, a database containing input values (gamma radiation distribution) and output values (water levels) had to be built. Here, the database was established by experiments (data from BWR and from a test setup) and simulation with the authorised thermo

  15. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  16. Ultrasonic flaw detection and sizing methods for cracks in the nozzle corner area at boiling water reactor vessels

    International Nuclear Information System (INIS)

    The demonstration of inservice inspection methods with ultrasound for the nozzel inner corner at boiling water reactor pressure vessels has shown that a detectability of cracks with a depth in the range of 5-10 mm is possible if optimal inspection parameters are chosen. The investigations concerning the choice of the optimal parameters is presented

  17. Ultrasonic flaw detection and sizing methods for cracks in the nozzle corner area at boiling water reactor vessels

    International Nuclear Information System (INIS)

    The demonstration of inservice inspection methods with ultrasound for the nozzle inner corner at boiling water reactor pressure vessels has shown that a detectability of cracks with a depth in the range of 5 - 10 mm is possible if optimal inspection parameters are chosen. The investigations concerning the choice of the optimal parameters is presented. 4 refs

  18. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Science.gov (United States)

    2010-01-01

    ...) of 10 CFR 50.34—Post-Accident Sampling for Boron, Chloride, and Dissolved Gases; and 3. Paragraph (f... design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Generic changes that... design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10...

  19. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to: (1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, (2) assess the RELAP5 and TRACE computer code against the experimental data, and (3) develop mathematical model and heat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal-hydraulic codes assessment

  20. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  1. A method of simulating voids in experimental studies of boiling water reactors

    International Nuclear Information System (INIS)

    The coolant density in boiling water reactors may vary from 3 at pressures up to 1000 p.s.i. In order to study the effect of reduced water density on reactivity in unpressurized experimental systems, the effective water density is reduced by packing small beads of highly expanded polystyrene into the fuel clusters and flooding the interstices with water. Coolant densities of from 0.4 to 0.6 gm/cm3 may be produced with the introduction of only about 0.4 gm/cm3 of non-hydrogeneous material. This memorandum describes the production, properties and handling of polystyrene beads and the tests carried out to establish the validity of the technique. (author)

  2. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  3. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    International Nuclear Information System (INIS)

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it's Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components

  4. A semiempirical prediction of the decay ratio for the boiling water reactors start-up process

    International Nuclear Information System (INIS)

    During the start-up of a commercial boiling water reactor (BWR), the power and the coolant flow are continuously monitored. In order to prevent power instability events, the decay ratio (DR) could also be monitored. The process can be made safer if the operator could anticipate the DR too. DR depends on the power, the flow and many other quantities such as axial and radial neutron flux distribution, feed water temperature, void fraction, etc. A simple relationship for DR is derived. Three independent variables seem to be enough: the power, the flow and a single parameter standing for all other quantities which affect the DR. The relationship is validated with data from commercial BWR start-ups. A practical procedure for the start-up of a BWR is designed; it could help preventing instability events

  5. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

  6. Evaluation of instrumentation for detection of inadequate core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    This report is a review of the Approach to Inadequate Core Cooling issue in Boiling Water Reactors (BWR). The report consists of seven sections. The principal conclusion is that the condition of the reference leg, and operator awareness of that condition are of primary importance in level indication reliability for safety. An indication of reference leg level and temperature displayed to the operators would be a useful enhancement of reliability and a guide to further operator action in all circumstances. We conclude that the BWR practice of multiple, redundant coolant level measurements, with overlapping ranges, can be a reliable basis for indication of approach to an ICC condition, and, in correlation with the other control and safety systems of modern BWRs, will prevent unsafe conditions

  7. A model for fuel rod and tie rod elongations in boiling water reactor fuel bundles

    International Nuclear Information System (INIS)

    A structural model is developed to determine the relative axial displacements of the spring held fuel rods to the tie rods in Boiling Water Reactor fuel bundles. An irradiation dependent relaxation model, which considers a two stage relaxation process dependent upon the fast fluence is used for the compression springs. The changes in spring compression resulting from the change in the length of the zircaloy fuel cladding due to irradiation enhanced anisotropic creep and growth is also considered in determining the time dependent variation of the spring forces. The time dependence of the average linear heat generation rates and their axial distributions is taken into account in determining the fuel cladding temperatures and fast fluxes for the various fuel rod locations within each of the BWR fuel bundles whose relative displacements were measured and used in this verification study. (orig.)

  8. Simulator evaluation of the Boiling Water Reactor Owners' Group (BWROG) graphics display system (GDS)

    International Nuclear Information System (INIS)

    This report describes the evaluation of a Graphic Display System (GDS). The GDS was developed by the Boiling Water Reactor Owners' Group (BWROG) to aid control room operators in detecting abnormal operating conditions, assessing the safety status of the plant, executing corrective action and monitoring plant response. The objective of the evaluation was to obtain recommendations for improving the usefulness of the GDS and to assess its usefulness under simulated accident operating conditions. The GDS presented 19 operator selectable displays on a high resolution color CRT monitor. The displays included safety function status, key parameters in bar and trend formats, and two-dimensional limits plots associated with the execution of symptom-based emergency procedures. Almost all of the operators, 94%, considered the GDS to be a useful device. The GDS was considered to be more useful for complex transients than for more straightforward events or routine operation

  9. Evaluation of a passive containment cooling system for a simplified BWR [boiling water reactor

    International Nuclear Information System (INIS)

    Simplified boiling water reactors (BWRs) are characterized for the adoption of a passive containment cooling system (PCCS) and a passive emergency core cooling system (ECCS). TOSPAC, which had been developed as the preliminary design code for several PCCS concepts, was compared with TRAC for verification. TOSPAC analyses were also performed to show the effectiveness of the isolation condenser (IC) as a PCCS over a wide range of break spectra. The selected reference plant for the analysis is a natural circulation BWR plant with 1,800-MW(thermal) power. The ECCS consists of a gravity-driven cooling system (GDCS) and depressurization valves. The IC and drywell cooler are considered for the PCCS. The IC units and drywell coolers are placed in the IC pool and GDCS pool, respectively

  10. Crack growth of intergranular stress corrosion cracks in austenitic stainless steel pipes of boiling water reactors

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) piping is considered from the crack growth rate point of view. Crack growth rate of sensitized austenitic stainless steel welds is dependent on the degree of sensitization of the material and the severity of the environment as well as the stress state. In evaluation of actual crack growth rate there are three major sources of uncertainty: knowledge of actual crack size and shape, actual stress distribution in he area of the crack and the degree of sensitization. In the report the crack growth calculations used in the USA and in Sweden are presented. Finally, the crack growth rate predictions based on mechanistic modelling of IGSCC and some needs of further research in Finland are considered

  11. New model of cobalt activity accumulation on stainless steel piping surfaces under boiling water reactor conditions

    International Nuclear Information System (INIS)

    A new technique for on-line measurement of corrosion amount and activity accumulation was developed. Cobalt activity accumulation tests were conducted under the normal water chemistry (NWC) condition (electrochemical corrosion potential (ECP): +0.15 V vs. SHE) and the hydrogen water chemistry (HWC) condition (ECP -0.30 V vs. SHE, -0.42 V vs. SHE) to evaluate cobalt activity accumulation under HWC conditions in boiling water reactors (BWRs). Total corrosion decreased and cobalt activity accumulation increased as ECP decreased. Experimental data were reproduced by a new model, in which cobalt activity deposits on oxide particle surfaces by absorption or replacement. This model estimated the cobalt activity accumulation under HWC conditions (ECP <-0.42 V vs. SHE) after 10000 h to be 12 times as large as that under NWC conditions (ECP +0.15 V vs. SHE). (author)

  12. Experimental investigations on load reduction in the pressure suppression system of boiling water reactors

    International Nuclear Information System (INIS)

    For the load specification of pressure suppression systems in boiling water reactors the periodic pressure pulses from a condensation phenomenon, called chugging, are of great importance. The research indicates, that the chugging mechanism is mainly induced by the BORDA-effect at the sharp edge of the vent pipe outlet. Based on these insights, simple vent pipe outlet mitigators are developed and tested, which effect in a passive mode a significant reduction of the dynamic pressure pulses from this condensation phenomenon. The results also yield the proof of multivent effect, of time window for single chugging event occurrence at a multivent configuration and the assurance of the reproducibility of this dynamic condensation phase. (orig.)

  13. Study of a Heavy-Water Reactor with Boiling Heavy-Water Coolant

    International Nuclear Information System (INIS)

    Among the possible types of heavy-water reactor, those cooled by heavy water would appear to combine the advantages of excellent neutron economy and a well-tried cladding material; this allows optimum utilization of uranium under the present conditions of technology. Placing the reactor, the handling equipment, and the heat exchangers together in a prestressed concrete vessel appreciably simplifies operating problems by reducing the number of hermetic seals in contact with the pressurized heavy water. This arrangement is only effective if a large proportion of the heat transfer is by phase change, so as to keep the amount of coolant to a minimum. The Commissariat à Energie Atomique has made a study of a boiling heavy-water reactor under a co-operation agreement with the Siemens and Sulzer Companies and with the participation of the Socia Company. The paper describes the main features of these projects as well as the main technological problems raised by this design which relate to the thermal insulation of the concrete vessel in the presence of a two-phase fluid; the handling equipment which must function in steam at 300°C; and the accessibility of the exchangers. (author)

  14. On-line test of power distribution prediction system for boiling water reactors

    International Nuclear Information System (INIS)

    A power distribution prediction system for boiling water reactors has been developed and its on-line performance test has proceeded at an operating commercial reactor. This system predicts the power distribution or thermal margin in advance of control rod operations and core flow rate change. This system consists of an on-line computer system, an operator's console with a color cathode-ray tube, and plant data input devices. The main functions of this system are present power distribution monitoring, power distribution prediction, and power-up trajectory prediction. The calculation method is based on a simplified nuclear thermal-hydraulic calculation, which is combined with a method of model identification to the actual reactor core state. It has been ascertained by the on-line test that the predicted power distribution (readings of traversing in-core probe) agrees with the measured data within 6% root-mean-square. The computing time required for one prediction calculation step is less than or equal to 1.5 min by an HIDIC-80 on-line computer

  15. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    International Nuclear Information System (INIS)

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner's Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section

  16. Eulerian two-phase computational fluid dynamics for boiling water reactor core analysis

    International Nuclear Information System (INIS)

    Traditionally, the analysis of two-phase boiling flows has relied on experimentally-derived correlations. This approach provides accurate predictions of channel-averaged temperatures and void fractions and even peak assembly temperatures within an assembly. However, it lacks the resolution needed to predict the detailed intra-channel distributions of temperature, void fraction and steaming rates that are needed to address the fuel reliability concerns which result from longer refueling cycles and higher burnup fuels, particularly for the prediction of potential fuel pin cladding failures resulting from growth of tenacious crud. As part of the ongoing effort to develop a high-fidelity, full-core, pin-by-pin, fully-coupled neutronic and thermal hydraulic simulation package for reactor core analysis], capabilities for Eulerian-Eulerian two-phase simulation within the commercial Computational Fluid Dynamics code Star-CD are being extended and validated for application to Boiling Water Reactor (BWR) cores. The extension of the existing capability includes the development of wall heat partitioning and bubble growth models, implementation of a topology map based approach that provides the necessary capability to switch between the liquid and vapor as the continuous phase on a cell-by-cell basis and the development of appropriate models for the inter-phase forces that influence the movement of bubbles and droplets. Two applications have been identified as an initial demonstration and validation of the implemented methodology. First, the model is being applied to an Atrium-10 fuel assembly from Cycle 11 of the River Bend Nuclear Power Plant. Second, the model is being applied to an international benchmark problem for validation of BWR assembly analysis methods. (authors)

  17. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  18. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    International Nuclear Information System (INIS)

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days

  19. Performance studies of a new core cooling monitor in a boiling water reactor

    International Nuclear Information System (INIS)

    Performance studies of a new type of core cooling monitors have been carried out in the Barsebaeck Nuclear Power Station during the operation periods 1988-10-04 to 1989-07-05, 1989-08-03 to 1990-09-05 and 1990-09-28 to 1991-07-04. The results showed that the monitors, which were placed inside the reactor core, are very sensitive to variations of the reactor operating conditions, and that 34 months of irradiation did not influence the signals from the monitors. Experiments were also carried out in a 160 bar loop, where sudden uncovers of the monitors were achieved by decreasing the liquid level of the coolant surrounding the monitors. The experiments included the pressures of 5, 20, 50, 70 and 155 bar, and the responses to uncover were in the ranges between 11 and 82 mV/sec or a total step change of 2 V at typical BWR conditions. This is of the order of two decades higher than the responses from monitors based on thermocouple readings. The monitors can be operated in two modes, the core cooling mode and the temperature mode. In the former mode the electrical current is 3-4 A, and in the latter mode, where the monitor actually serves as a thermometer, the current is in the order of 50-100 mA. In the laboratory the monitors have been studied for temperatures up to 1265 deg. C, which is very useful in case of a severe reactor accident. Thus, during such events the temperatures in the reactor core could be followed up to this level and the monitors could also be used to activate certain safety equipment. The function as well as the design of the instrument is verified in laboratory experiments, computer calculations and reactor tests and is now ready for implementation in the BWR instrumentation. In summary: 1. The proposed monitor can operate in two modes; the core cooling mode and the temperature mode. 2. Laboratory studies have shown that the responses to uncover are two decades higher than signals from monitors based on thermocouple readings. 3. No effects of

  20. Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

    International Nuclear Information System (INIS)

    With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

  1. Bubble spreading during the boiling crisis: modelling and experimenting in microgravity

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Y; Lecoutre, C; Chatain, D

    2016-01-01

    Boiling is a very efficient way to transfer heat from a heater to the liquid carrier. We discuss the boiling crisis, a transition between two regimes of boiling: nucleate and film boiling. The boiling crisis results in a sharp decrease in the heat transfer rate, which can cause a major accident in industrial heat exchangers. In this communication, we present a physical model of the boiling crisis based on the vapor recoil effect. Under the action of the vapor recoil the gas bubbles begin to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes its spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. In the experiments both in the Mir spa...

  2. Verification of computer code for calculation of coolant radiolysis in the VVER reactor core with regard for boiling in its upper part

    International Nuclear Information System (INIS)

    Code Bora for WWER coolant radiolysis calculation considering single jets boiling in the reactor core top part is developed on the basis of computer codes MOPABA-H2 (radiolysis of aqueous solutions) and SteamRad (radiolysis of vapor). Physico-chemical processes taking place in boiling core coolant are complex and diversified. Still, for the solution of certain problems their simulation can be simplified. The approach of reasonable simplification was used for development of code Bora: mathematical model assumed is purposed for simulation of phenomena only in the area of interest; the number of simulated chemical reactions and particles shall be reasonably minimum; complexity of interphase mass transfer calculation procedure shall be adequate to actually available accuracy of modeling. The analysis of new experimental initial yields of water radiolysis products data and kinetic parameters of elementary chemical reactions with their participation has been carried out. Some changes have been introduced in the mechanism of liquid water and aqueous solutions of ammonia radiolysis have been significantly revised on the basis of this analysis. Examples of the calculations provided for code Bora verification are presented. Despite of very simple simulation of interphase mass transfer, Bora allows to obtain average chemical composition of two-phase coolant at BWR core outlet with the accuracy sufficient for engineering calculations. The report also presents the results of two-phase coolant chemical composition test calculation for reactor core top part coolant boiling in pressurized water reactor. (author)

  3. Fuel lattice design in a boiling water reactor using an ant-colony-based system

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis, E-mail: joseluis.montes@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Facultad de Ciencias, Universidad Autonoma del Estado de Mexico (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Ortiz, Juan Jose, E-mail: juanjose.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Perusquia, Raul, E-mail: raul.perusquia@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico)

    2011-06-15

    Research highlights: > We present an ant-colony-based system for BWR fuel lattice design and optimization. > Assessment of candidate solutions at 0.0 MWd/kg {sup 235}U seems to have a limited scope. > Suitable heuristic rules enable more realistic fuel lattice designs. > The election of the objective has a large impact in CPU time. > ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U{sup 235} enrichment and Gd{sub 2}O{sub 3} concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U{sup 235} enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U{sup 235} enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U{sup 235} enrichment; whereas, the k-infinity was inside the {+-}100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a

  4. Fuel lattice design in a boiling water reactor using an ant-colony-based system

    International Nuclear Information System (INIS)

    Research highlights: → We present an ant-colony-based system for BWR fuel lattice design and optimization. → Assessment of candidate solutions at 0.0 MWd/kg 235U seems to have a limited scope. → Suitable heuristic rules enable more realistic fuel lattice designs. → The election of the objective has a large impact in CPU time. → ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U235 enrichment and Gd2O3 concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U235 enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U235 enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U235 enrichment; whereas, the k-infinity was inside the ±100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a powerful tool to tackle this step of

  5. TRIGA reactor operating experience

    International Nuclear Information System (INIS)

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  6. Reactor Neutrino Experiments: $\\theta_{13}$ and Beyond

    CERN Document Server

    Qian, X

    2014-01-01

    We review the current-generation short-baseline reactor neutrino experiments that have firmly established the third neutrino mixing angle $\\theta_{13}$ to be non-zero. The relative large value of $\\theta_{13}$ (around 9$^\\circ$) has opened many new and exciting opportunities for future neutrino experiments. Daya Bay experiment with the first measurement of $\\Delta m^2_{ee}$ is aiming for a precision measurement of this atmospheric mass-squared splitting with a comparable precision as $\\Delta m^2_{\\mu\\mu}$ from accelerator muon neutrino experiments. JUNO, a next-generation reactor neutrino experiment, is targeting to determine the neutrino mass hierarchy with medium baselines ($\\sim$50 km). Beside these {\\color{black} opportunities enabled by the large $\\theta_{13}$}, the current-generation (Daya Bay, Double Chooz, and RENO) and the next-generation (JUNO, RENO-50, and PROSPECT) reactor experiments, with their unprecedented statistics, are also leading the precision era of the 3-flavor neutrino oscillation phys...

  7. Decontamination and decommissioning of the Experimental Boiling Water Reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The Experimental Boiling Water Reactor (EBWR), located on the Argonne National Laboratory-East (ANL-E) site, started operations in 1957. The initial rating was 20 MW(t). The rating was eventually increased to 70 MW(t) in 1959 and 100 MW(t) in 1962. The reactor was shut down in 1967 and all of the fuel was removed from the facility. The facility was placed in dry lay-up until 1986. ANL-E personnel started the decontamination and decommissioning (D ampersand D) effort in 1986. Supporting equipment such as the external steam system and some of the upper reactor components, the core riser and the top fuel shroud, were removed at that time. Characterization of the facility was also undertaken. The contract to complete the EBWR D ampersand D Project was issued in December 1993. The initial schedule called for the final effort to be divided into five phases that were to be completed over a four year period. However, this schedule was subsequently consolidated, at the request of ANL-E, to a thirteen month period, with the on-site work to be completed by the end of 1994. The EBWR D ampersand D Project is approximately 88% complete. A small quantity of reactor internals remains to be volume reduced along with the removal of the SFSP water treatment system. Upon completion of this work the facility will be decontaminated and a final survey completed. The planned completion of on-site work is scheduled for July 1995

  8. Reactor stability estimation in a boiling water reactor using a multivariate autoregressive model

    International Nuclear Information System (INIS)

    As to the stability phenomenon in the core of a BWR, as test methods have become sophisticated to yield high quality data, the information obtained from actual test data has given a better insight into the way of coupling neutron kinetics and thermal hydroulic phenomena. In the dynamic characteristics of a BWR, the reactor core and pressure regulator characteristics are dominant, and both are strongly coupled. The reactor core stability is defined by the invessel reactor dynamics without the effect of plant controllers, therefore, it is necessary to decouple the reactor core dynamics from the pressure regulator in order to estimate accurately the stability performance. In this study, autoregressive technique was applied to both artificial disturbance data, that is, pressure perturbation data and steady state noise data, and it was demonstrated that this model figging yielded more realistic stability performance index than ordinary correlation method, and that the stability index was able to be identified from noise data. The auto regressive fitting of small pressure perturbation test data and the evaluation by transient model and by noise analysis are reported. (Kako, I.)

  9. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    Energy Technology Data Exchange (ETDEWEB)

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

  10. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    International Nuclear Information System (INIS)

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path

  11. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    Energy Technology Data Exchange (ETDEWEB)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment.

  12. Potential uses of high gradient magnetic filtration for high-temperature water purification in boiling water reactors

    International Nuclear Information System (INIS)

    Studies of various high-temperature filter devices indicate a potentially positive impact for high gradient magnetic filtration on boiling water reactor radiation level reduction. Test results on in-plant water composition and impurity crystallography are presented for several typical boiling water reactors (BWRs) on plant streams where high-temperature filtration may be particularly beneficial. An experimental model on the removal of red iron oxide (hematite) from simulated reactor water with a high gradient magnetic filter is presented, as well as the scale-up parameters used to predict the filtration efficiency on various high temperature, in-plant streams. Numerical examples are given to illustrate the crud removal potential of high gradient magnetic filters installed at alternative stream locations under typical, steady-state, plant operating conditions

  13. The stability analysis using two fluids (SAT) code for boiling flow systems: Volume 4, Experiments and model validation

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.P.; Dykhuizen, R.C.; Su, M.G.; Jain, P.

    1988-12-01

    This report presents analyses of dynamic instability and frequency response characteristics of boiling flow systems based on an unequal velocity, unequal temperature two-fluid model of such flow. The dynamic instability analyses in the time domain are incorporated into three options of a computer code SAT (steady state, or equilibrium point analyses; linear stability analysis; and nonlinear analysis). The frequency response analysis is incorporated into a fourth option FREQ. Results from dynamic instability experiments carried out in a Refrigerant-113 boiling flow rig are also reported as are comparison of these with linear stability analysis predictions. Descriptions of the model, the computational techniques, the computer codes, the experiments and model validation are divided into the following volumes: Volume 1, theoretical model and computational formulation; Volume 2, coding description; Volume 3, user's manual; and Volume 4, experiments and model validation. Instability experiments run in our Refrigerant-113 boiling flow facility are described in this document. Results from these experiments are compared with predictions of the theoretical model. Instability experiment data from two other facilities and frequency response results from one are compared with theoretical model predictions also. 19 refs., 41 figs.

  14. Recriticality in a BWR [boiling water reactor] following a core damage event

    International Nuclear Information System (INIS)

    This report describes the results of a study conducted by Pacific Northwest Laboratory to assist the US Nuclear Regulatory Commission in evaluating the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Based on a conservative bounding analysis, this report concludes that there is a potential for recriticality in BWRs if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event will most likely not generate a pressure pulse significant enough to fail the vessel. Instead, a quasi-steady power level would result and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event. Two strategies are identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached. The first strategy involves initiating boration injection at or before the time of core reflood if the potential for control blade melting exists. The second strategy involves initiating residual heat removal suppression pool cooling to remove the heat load generated by the recriticality event and thus extend the time available for boration. 31 figs., 17 tabs

  15. Non linear analysis of out of phase oscillations in boiling water reactors

    International Nuclear Information System (INIS)

    Out of phase oscillations have been observed recently in many boiling water reactors during stability tests and also in start-up conditions. Many authors have attempted to explain these regional oscillations, but the explanations given are not complete. In this paper, we develop a non-linear phenomenological model that can explain, both in phase and out of phase oscillations. The neutronic loop has been described on the basis of an expansion in terms of λ-modes. Furthermore, for a semiquantitative representation of the dynamics, reduced order model have been obtained reducing the number of regions, modes and energy groups considered in the problem. In this line, we propose a model that qualitatively explains the dynamic behavior of these oscillations verifying that in phase oscillations only appear when the azimuthal model has not enough thermal-hydraulic feedback to overcome the eigenvalue separation and also, that it is possible that self-sustained out of phase oscillations arise due to the different thermal-hydraulic properties of the two reactor core lobes, if the modal reactivities have appropriate feedback gains. (author)

  16. Feasibility of core management system by data communication for boiling water reactors

    International Nuclear Information System (INIS)

    A core management system by data communication has been designed and proposed for more efficient operation of boiling water reactor (BWR) plants by faster transmission and centralized management of information. The system comprises three kinds f computers: a process computer for monitoring purposes at the reactor site, a center computer for administration purposes at the head office, and a large scientific computer for planning and evaluation purposes. The process and the large computers are connected to the center computer by a data transmission line. To demonstrate the feasibility of such a system, the operating history evaluation system, which is one of the subsystems of the core management system, has been developed along the above concept. Application to the evaluation of the operating history of a commercial BWR shows a great deal of merit. Quick response and a significant manpower reduction can be expected by data communication and minimized intervention of human labor. Visual display is also found to be very useful in understanding the core characteristics

  17. New strategies of reloads design and models of control bars in boiling water reactors

    International Nuclear Information System (INIS)

    In this work the results obtained when analyzing new strategies in the reload designs of nuclear fuel and models of control bars, for boiling water reactors are presented. The idea is to analyze the behaviour of the reactor during an operation cycle, when the heuristic rules are not used (commonly used by expert engineers in both designs). Specifically was analyzed the rule of low leak and the load strategy Control Cell Core for the design of a fuel reload. In a same way was analyzed the rule of prohibiting the use of the intermediate positions in the control bars, as well as the construction of bar models based on load strategies type Control Cell Core. In the first analysis a balance and transition cycle were used. For the second analysis only a transition cycle was used, firstly with the reloads designed in the first analysis and later on with reloads built by other methods. For the simulation of the different configurations proposed in both cases, was used the code Simulate-3. To obtain the designs in both studies, the heuristic techniques or neural networks and taboo search were used. The obtained results show that it can be omitted of some rules used in the ambit for the mentioned designs and even so to obtain good results. To carry out this investigation was used Dell work station under Li nux platform. (Author)

  18. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  19. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    International Nuclear Information System (INIS)

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ''Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs

  20. Dilute chemical decontamination process for pressurized and boiling water reactor applications

    International Nuclear Information System (INIS)

    Westinghouse Electric Corporation (WEC) has developed five chemical processes for nuclear decontamination, based on extensive experimental testing using radioactive pressurized water reactor (PWR) and boiling water reactor (BWR) samples. The dilute chemical decontamination process offers the best combination of effectiveness, low corrosion, low waste volume, and fast field implementation time. This is an alternating multistep process. For PWRs, an oxidation treatment is necessary. Projected contact decontamination factors (DFs) are about 50 on plant Inconel surfaces, with comparable results on stainless steel. Actual test DFs have exceeded 500 in the process test loop. For BWRs, an oxidation step is unnecessary, but very beneficial. DFs of 10 to 20 are achieved without an oxidation treatment. Full process DFs exceed 500 when the oxidation treatment is included. Low corrosion rates are observed, without any adverse effects. Only solid waste is produced by the process. WEC has fabricated a trailer-mounted application system for this process, and is offering it as a decontamination service to commercial customers

  1. Measurement and analysis of structural activation in a boiling water reactor

    International Nuclear Information System (INIS)

    Induced radioactivity of structural materials of a nuclear power plant introduces the possibility of exposure of workers. In order to assess evaluation accuracy of the induced radioactivity, measurements and calculations were performed for gamma-ray dose inside an irradiated reactor pressure vessel of a boiling water reactor. Neutron flux was calculated with two-dimensional Sn transport code DOT3.5 with RZ and RΘ models. Induced radioactivity was calculated with the ORIGEN-79 code, in which three-group activation cross section was produced considering neutron spectrum instead of the original ORIGEN-79 three-group constants. Calculated dose rate by DOT3.5 agreed well with the measured value, and calculational accuracy was improved by taking account of Θ dependence of neutron flux distribution and precise neutron spectrum in activation calculation compared to the calculation with a simplified method such as a single RZ model calculation of neutron flux and activity calculation with the three-group constants built-in the ORIGEN-79 code. (author)

  2. Verification of advanced methods in TARMS boiling water reactor core management system

    International Nuclear Information System (INIS)

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site Boiling Water Reactor (BWR) core operation management system. It covers almost all the functional requirements to the current process computer to increase on-site core management capability, capacity factors, thermal margins, fuel reliability, and so on, by supporting application functions for monitoring the present core power distribution, and for aiding site engineers in making the core operation plans, by predicting future core performance. It is based on a three dimensional, 1.5 energy group, coarse mesh nodal diffusion theory code ''LOGOS02'', and includes advanced methods to increase the accuracy of core power distribution calculations as well as a local peaking factor calculation method by which the effect of neighboring nodes on intra-nodal power distribution can be considered. TARMS has been installed in eight BWR plants and was verified to be an effective BWR core operation management tool. This paper describes its advanced methods and the results of verifications with actual plant data. (author). 3 refs, 6 figs

  3. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  4. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    International Nuclear Information System (INIS)

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval

  5. Anticipated transient without scram analysis of the simplified boiling water reactor following main steam isolation valve closure with boron injection

    International Nuclear Information System (INIS)

    The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system

  6. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    The two-phase flow instability is of interest for the design and operation of many industrial systems such as boiling water reactors (BWRs), chemical reactors, and steam generators. In case of BWRs, the flow instabilities are coupled to the power instabilities via neutronic-thermal hydraulic feedbacks. Since these instabilities produce also local pressure oscillations, the coolant flashing plays a very important role at low pressure. Many frequency-domain codes have been used for two-phase flow stability analysis of thermal hydraulic industrial systems with particular emphasis to BWRs. Some were ignoring the effect of the local pressure, or the effect of 3D power oscillations, and many were not able to deal with the neutronics-thermal hydraulics problems considering the entire core and all its fuel assemblies. The new frequency domain tool uses the best available nuclear, thermal hydraulic, algebraic and control theory methods for simulating BWRs and analyzing their stability in either off-line or on-line fashion. The novel code takes all necessary information from plant files via an interface, solves and integrates, for all reactor fuel assemblies divided into a number of segments, the thermal-hydraulic non-homogenous non-equilibrium coupled linear differential equations, and solves the 3D, two-energy-group diffusion equations for the entire core (with spatial expansion of the neutron fluxes in Legendre polynomials).It is important to note that the neutronics equations written in terms of flux harmonics for a discretized system (nodal-modal equations) generate a set of large sparse matrices. The eigenvalue problem associated to the discretized core statics equations is solved by the implementation of the implicit restarted Arnoldi method (IRAM) with implicit shifted QR mechanism. The results of the steady state are then used for the calculation of the local transfer functions and system transfer matrices. The later are large-dense and complex matrices, (their size

  7. Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

    International Nuclear Information System (INIS)

    Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs

  8. 78 FR 58575 - Review of Experiments for Research Reactors

    Science.gov (United States)

    2013-09-24

    ... COMMISSION Review of Experiments for Research Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Guide (RG) 2.4, ``Review of Experiments for Research Reactors.'' The guide is being withdrawn because... Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because its guidance no longer...

  9. Overview on stability of natural-circulation-cooled boiling water reactors during start-up. An experimental and modeling analysis

    International Nuclear Information System (INIS)

    This paper provides an overview on numerical and experimental work focused on flashing-induced instabilities. These instabilities may occur in natural circulation two-phase systems when operated at low pressure and low power. Therefore they are of special interest for the start-up phase of natural circulation Boiling Water Reactors. The work presented in this paper has been performed within the framework of the NACUSP project (European-Union Fifth Framework Program). Experiments were carried out on a steam/water natural circulation loop (CIRCUS), built at the Delft University of Technology. Information was gained on the characteristics of the flow oscillations and on the void fraction production during flashing in stationary and transient conditions. A 3-D flow-pattern visualization was achieved by means of advanced instrumentation, namely wire-mesh sensors. On the basis of the experimental results, an assessment of existing drift-flux models was performed for flashing flow. The most suitable drift-flux model was implemented in the 4-equations two-phase model FLOCAL, developed at the Forschungszentrum Rossendorf (FZR, Germany). The model allows for the liquid and steam to be in thermal non-equilibrium and, via drift-flux models, to have different velocities. A detail comparison between simulations and experiments is reported. (author)

  10. Design and Experimental Study for Development of Pb-Bi Cooled Direct Contact Boiling Water Small Fast Reactor (PBWFR)

    International Nuclear Information System (INIS)

    A design concept of Pb-Bi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. In the PBWFR, water is injected into hot Pb-Bi above the core, and direct contact boiling takes place in the chimney. The boiling two-phase flow in the chimney serves as a steam lift pump and a steam generator. A two-region core is designed. A decrease in reactivity was estimated to be 1.5 % dk/kk' for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The chimney, cyclone separators and chevron dryers, direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed. For the technical development of the PBWFR, experimental and analytical studies are performed for Pb-Bi direct contact boiling two-phase flow, steel corrosion in Pb-Bi flow, oxygen control and oxygen sensor, and removal of polonium contamination. (authors)

  11. Conceptual design and safety characteristics of the natural circulation boiling water reactor HSBWR-600

    International Nuclear Information System (INIS)

    The HSBWR (Hitachi Small BWR) with a rated capacity of 600 MW electricity has been conceptually designed. The components and systems are simplified by adopting natural circulation and the passive ECCS, and eliminating steam separators. The volume of the reactor building is about 50% of that for current BWRs with the same rated capacity, and the construction period is 32-36 months until commercial operation. The major safety systems are: (1) an accumulated water injection system as an ECCS; (2) an outer pool, which stands outside of the steel primary containment vessel, as a long term cooling system after LOCAs; and (3) a steam driven reactor core isolation cooling system for high pressure water injection. The grace period is one day for core cooling and 3 days for the containment vessel heat removal. The infinite grace period for core cooling is also available as an option. LOCA analysis showed that the core will always be covered by a two-phase mixture, resulting in no core heat-up. The fundamental experiments and analyses showed sufficient capability of the outer pool for long term heat removal. (author). 12 refs, 17 figs, 3 tabs

  12. Operational experience of the Marcoule reactors

    International Nuclear Information System (INIS)

    The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author)

  13. Experiments of Pool Boiling Performance (Boiling Heat Transfer and Critical Heat Flux) on Designed Micro-Structures

    International Nuclear Information System (INIS)

    In general, the evaluation of the boiling performance mainly focuses on two physical parameters: boiling heat transfer (BHT) and critical heat flux (CHF). In the nuclear power plants, both BHT and CHF contribute the nuclear system efficiency and safety, respectively. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on Pin-fin effect analysis. In terms of CHF, critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on the roughness ratio. The extended heat transfer area contributes the boiling heat transfer increase on the structured surface, and its quantitative analysis has been performed. In terms of CHF, the critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. We suggested a capillary limit to CHF delay for modeling capillary induced liquid inflow through microstructured surfaces. The critical size of the capillary limit on the prepared structured surface, determined by a model, could be reasonable explanation points for the experimental results (optimal size for CHF delay). The present experimental results also showed clearly the critical size (10 - 20 μm) for CHF delay, predicted by capillary limit analysis. This study provides fundamental insight into BHT and CHF enhancement of structured surfaces, and an optimal design guide for the required CHF and boiling heat-transfer performance. Finally, this study can contribute the basic understanding of the boiling on designed microstructure surface, and it also suggest the optimal micro scaled structured surface of boiling

  14. Surface activity and radiation field measurements of the TMI-2 reactor building gross decontamination experiment

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C V

    1983-10-01

    Surface samples were collected from concrete and metal surfaces within the Three Mile Island Unit 2 Reactor Building on December 15 and 17, 1981 and again on March 25 and 26, 1982. The Reactor Building was decontaminated by hydrolasing during the period between these dates. The collected samples were analyzed for radionuclide concentration at the Idaho National Engineering Laboratory. The sampling equipment and procedures, and the analysis methods and results are discussed. The measured mean surface concentrations of /sup 137/Cs and /sup 90/Sr on the 305-ft elevation floor before decontamination were, respectively, 3.6 +- 0.9 and 0.17 +- 0.04 ..mu..Ci/cm/sup 2/. Their mean concentrations on the 347-ft elevation floor were about the same. On both elevations, walls were found to be considerably less contaminated than floors. The fractions of the core inventories of /sup 137/Cs, /sup 90/Sr, and /sup 129/I deposited on Reactor Building surfaces prior to decontamination were calculated using their mean concentrations on various types of surfaces. The calculated values for these three nuclides are 3.5 +- 0.4 E-4, 2.4 +- 0.8 E-5, and 5.7 +- 0.5 E-4, respectively. The decontamination operations reduced the /sup 137/Cs surface activity on the 305- and 347-ft elevations by factors of 20 and 13, respectively. The /sup 90/Sr surface activity reduction was the same for both floors, that being a factor of 30. On the whole, decontamination of vertical surfaces was not achieved. Beta and gamma exposure rates that were measured during surface sampling were examined to determine the degree to which they correlated with measured surface activities. The data were fit with power functions of the form y = ax/sup b/. As might be expected, the beta exposure rates showed the best correlation. Of the data sets fit with the power function, the set of December 1981 beta exposure exhibited the least scatter. The coefficient of determination for this set was calculated to be 0.915.

  15. Behavior of bubble in subcooled boiling with forced convection, 2

    International Nuclear Information System (INIS)

    The objective of this research is to observe the bubble behavior in subcooled boiling with forced convection and to investigate the conditions of heaving test with a heater which initiates isolated bubbles and with high speed motion photography. The observation was made at three inlet subcooling of 15, 30 and 45 K with pressure of 0.3 MPa, mass flux of 1000 kg/m2·s, heat flux of 35 kW/m2. At inlet subcooling of 15 K, bubble velocity is nearly constant in the field of visions for upstream and downstream area. But at the higher subcooling, bubble velocity varies greatly and bubble collapse in the camera field. (author)

  16. Dynamic behavior of large oxide-fueled fast reactors during over-power transients due to boiling of sodium

    International Nuclear Information System (INIS)

    The dynamic behavior of large oxide-fueled fast reactors during over-power transients or under-flow situations which result in boiling of sodium used as coolant. The fuel heat transfer was analysed to determine the fuel temperature profile and the Doppler feedback reactivity. The sodium pressure, temperature, mass flow rate and sodium voiding reactivity were obtained by solving the basic coolant hydrodynamics equations. (Author)

  17. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  18. Measurement on the effect of sound wave in upper plenum of boiling water reactor

    International Nuclear Information System (INIS)

    In recent years, the power uprate of Boiling Water Reactors have been conducted at several existing power plants as a way to improve plant economy. In one of the power uprated plants (117.8% uprates) in the United States, the steam dryer breakages due to fatigue fracture occurred. It is conceivable that the increased steam flow passing through the branches caused a self-induced vibration with the propagation of sound wave into the steam-dome. The resonance among the structure, flow and the pressure fluctuation resulted in the breakages. To understand the basic mechanism of the resonance, previous researches were done by a point measurement of the pressure and by a phase averaged measurement of the flow, while it was difficult to detect the interaction among them by the conventional method. In this study, Dynamic Particle Image Velocimetry (PIV) System was applied to investigate the effect of sound on natural convection and forced convection. Especially, when the phases of acoustic sources were different, various acoustic wave effects were checked. (author)

  19. Remarks on boiling water reactor stability analysis. Pt. 1. Theory and application of bifurcation analysis

    International Nuclear Information System (INIS)

    Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants. (orig.)

  20. Stability monitor for Boiling Water Reactors based on the Multivariate Empirical Mode Decomposition

    International Nuclear Information System (INIS)

    Highlights: • Stability monitor based on a novel technique was developed. • Multivariate Empirical Mode Decomposition (MEMD) to BWR data was applied. • Decision rules permitting to raise an instability alarm based on MEMD. • MEMD can estimate the phase of the density wave from measurements of two LPRMs. • This method detects BWR instabilities (DR increase and out-of phase oscillations). - Abstract: In this work a stability monitor based on a novel technique is presented. This monitor permits to launch general alarms indicating incipient high decay ratios (DR) and out-of-phase oscillations, in a simultaneous way time along. The implemented methodology to determine the estimations of DR and out-of-phase oscillations is based on the Multivariate Empirical Mode Decomposition (MEMD) processing the information obtained from all LPRMs located across the core of Boiling Water Reactor (BWR). The extracted modes with the MEMD, called the Intrinsic Mode Functions (IMFs), permit to tracking the oscillation associated to the density wave. The Case 9 (presenting high DRs and apparently out-of-phase oscillations simultaneously) from the Ringhals stability benchmark was used to show the effectiveness of the proposed methodology

  1. Current status of steam dryer performance under power uprate in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Highlights: • Steam dryer performance after extended power uprate is considered. • Effects of Acoustic Side Branches (ASB) on steam dryers is analyzed. • The ABS represents a reduction in the acoustic loads to the steam dryer. • Spectrograms of signals were obtained for frequency analysis. - Abstract: This work is a compilation of the current status of the steam dryer performance after the implementation of power uprates in Boiling Water Reactors (BWR). Some Nuclear Power Plants (NPPs) have reported failures and cracking in the steam dryer by acoustic resonances that cause excessive vibration due to the increase of steam flow. The replacement of the steam dryer, structural reinforcement and the connection of Acoustic Side Branches (ASB) are the main solutions adopted in order to avoid mechanical failures. The signal analysis of the vibration of the main steam lines in a typical BWR5, was performed using the Short-Time Fourier Transform (STFT). Signals were collected by the strain gauges located around the main steam lines (MSL). Two scenarios are analyzed; the first, is the signal obtained before the installation of the ASB, and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer

  2. Remote mechanized equipment for the repair and replacement of boiling water reactor recirculation loop piping

    International Nuclear Information System (INIS)

    Equipment has been assembled for the remote repair or replacement of boiling water reactor nuclear plant piping in the diameter range of 4 to 28 inches (10-71 cm). The objectives of this program were to produce high-quality pipe welds, reduce plant downtime, and reduce man-rem exposure. The repair strategy was to permit repair personnel to install and check out the repair subsystems and then leave the radiation zone allowing the operations to be conducted at a distance of up to 300 feet (91 m) from the operator. The complete repair system comprises subsystems for pipe severing, dimensional gaging, joint preparation, counterboring, welding, postweld nondestructive inspection (conceptual design), and audio, electronic, and visual monitoring of all operations. Components for all subsystems, excluding those for postweld nondestructive inspection, were purchased and modified as needed for integration into the repair system. Subsystems were designed for two sizes of Type 304 stainless steelpipe. For smaller, 12-inch-diameter (30.5 cm) pipe, severing is accomplished by a power hack saw and joint preparation and counterboring by an internally mounted lathe. The 22-inch-diameter (56 cm) pipe is severed, prepared, and counterbored using an externally mounted, single-point machining device. Dimensional gaging is performed to characterize the pipe geometry relative to a fixed external reference surface, allowing the placement of the joint preparation and the counterbore to be optimized. For both pipe sizes, a track-mounted gas tungsten-arc welding head with filler wire feed is used

  3. Aging analysis on ethylene propylene rubber insulations sampled from boiling water reactor containments

    International Nuclear Information System (INIS)

    Frame-retardant ethylene propylene rubber insulations that were service-used for 16 years in boiling-water reactor containments were subjected to material analyses. Value of elongation at break is approximately 270%, indicating that some change in the functional properties has occurred in their normal operation history. Additional artificial aging based on wear-out approach methodology was then conducted in a 110degC thermostatic oven. Seventy days were needed to reduce the mechanical property till it cannot secure the durability against the loss-of-coolant accident simulation test, which is almost three times greater than the prediction based on the acceleration aging test against non-aged samples. Infrared spectroscopy shows that the formation of carbonyl products is also suppressed in service-used cables. Moreover, gel fraction analysis shows that cross-linking is prominent and is enhanced during the additional wear-out aging in service-used cables. The relation between elongation at break value and tensile strength shows similar indication, and demonstrates that chain scission due to oxidation is elevated in the case of acceleration aging. Cross-linking phenomenon in used cables is considered to prevent chemical degradation reaction, by preventing the oxygen permeation into the bulk. This intrinsic feature of service-used cable insulation has a possibility of contributing to the slow aging under the in-service atmosphere, which was already found by the statistical analysis of the aging state data obtained for the corresponding type of insulations. (author)

  4. Multivent effects in a large scale boiling water reactor pressure suppression system

    International Nuclear Information System (INIS)

    The steam-driven GKSS pressure suppression test facility, which contains 3 full scale vent pipes, has been used for 5 years to investigate the postulated loss-of-coolant accident in a Mark II and Type 69 boiling water reactor. Using the results from several of these tests, wetwell boundary load data (peak pressures and spectral power) during the chugging stage, have been evaluated for sparse pool response (one and two vents in the three vent pool) and for full pool response (one, two, or three vent operation in pools of constant wetwell pool area per vent). The sparse pool results indicate the pool-system, chug event boundary loads are strongly dependent on wetwell pool area per vent, with the load increasing with decreasing area. The full pool results show a substantial increase in the pool-system, chug event boundary loads upon a change from single cell to double cell operation; only minor change occurs in going from double to triple cell operation

  5. Fuzzy logic control of water level in advanced boiling water reactor

    International Nuclear Information System (INIS)

    The feedwater control system in the Advanced Boiling Water Reactor (ABWR) is more challenging to design compared to other control systems in the plant, due to the possible change in level from void collapses and swells during transient events. A basic fuzzy logic controller is developed using a simplified ABWR mathematical model to demonstrate and compare the performance of this controller with a simplified conventional controller. To reduce the design effort, methods are developed to automatically tune the scaling factors and control rules. As a first step in developing the fuzzy controller, a fuzzy controller with a limited number of rules is developed to respond to normal plant transients such as setpoint changes of plant parameters and load demand changes. Various simulations for setpoint and load demand changes of plant performances were conducted to evaluate the modeled fuzzy logic design against the simplified ABWR model control system. The simulation results show that the performance of the fuzzy logic controller is comparable to that of the Proportional-Integral (PI) controller, However, the fuzzy logic controller produced shorter settling time for step setpoint changes compared to the simplified conventional controller

  6. Boiling Water Reactor Loading Pattern Optimization Using Simple Linear Perturbation and Modified Tabu Search Methods

    International Nuclear Information System (INIS)

    An automated system for designing a loading pattern (LP) for boiling water reactors (BWRs) given a reference LP and control rod (CR) sequence has been developed. This system employs the advanced nodal code SIMULATE-3 and a BWR LP optimization code FINELOAD-3, which uses a simple linear perturbation method and a modified Tabu search method to select potential optimized LP candidates. Both of these unique methods of FINELOAD-3 were developed to achieve an effective BWR LP optimization strategy and to have high computational efficiency. FINELOAD-3 also adjusts deep CR positions to compensate for the core reactivity deviation caused by fuel shuffling. The objective function is to maximize the end-of-cycle core reactivity while satisfying the specified thermal margins and cold shutdown margin constraints. This optimization system realized the practical application for real BWR LP design. Computer time needed to obtain an optimized LP for a typical BWR/5 octant core with 15 depletion steps is ∼4 h using an engineering workstation. This system was extensively tested for real BWR reload core designs and showed that the developed LPs using this system are equivalent or better than the manually optimized LPs

  7. AXIAL: a system for boiling water reactor fuel assembly axial optimization using genetic algorithms

    International Nuclear Information System (INIS)

    A system named AXIAL is developed based on the genetic algorithms (GA) optimization method, using the 3D steady state simulator code Core-Master-PRESTO (CM-PRESTO) to evaluate the objective function. The feasibility of this methodology is investigated for a typical boiling water reactor (BWR) fuel assembly (FA). The axial location of different fuel compositions is found in order to minimize the FA mean enrichment needed to obtain the cycle length under the safety constraints. Thermal limits are evaluated at the end of cycle using the Haling calculation; the hot excess reactivity and the shutdown margin at the beginning of cycle are also evaluated. The implemented objective function is very flexible and complete, incorporating all the thermal and reactivity limits imposed during fuel design analysis; furthermore, additional constraints can be easily introduced in order to obtain an improved solution. The results show a small improvement in the FA average enrichment obtained with the system related to the reference case that has been studied. The results show that the system converge to an optimal solution, it is observed that the mean fuel enrichment decreases while all the constraints are satisfied. A comparison was also performed using one-point and two-points crossover operator and the results of a sensitivity study for different mutation percentage are also showed

  8. Automatic boiling water reactor loading pattern design using ant colony optimization algorithm

    International Nuclear Information System (INIS)

    An automatic boiling water reactor (BWR) loading pattern (LP) design methodology was developed using the rank-based ant system (RAS), which is a variant of the ant colony optimization (ACO) algorithm. To reduce design complexity, only the fuel assemblies (FAs) of one eight-core positions were determined using the RAS algorithm, and then the corresponding FAs were loaded into the other parts of the core. Heuristic information was adopted to exclude the selection of the inappropriate FAs which will reduce search space, and thus, the computation time. When the LP was determined, Haling cycle length, beginning of cycle (BOC) shutdown margin (SDM), and Haling end of cycle (EOC) maximum fraction of limit for critical power ratio (MFLCPR) were calculated using SIMULATE-3 code, which were used to evaluate the LP for updating pheromone of RAS. The developed design methodology was demonstrated using FAs of a reference cycle of the BWR6 nuclear power plant. The results show that, the designed LP can be obtained within reasonable computation time, and has a longer cycle length than that of the original design.

  9. Automatic boiling water reactor loading pattern design using ant colony optimization algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C.-D. [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China); Nuclear Engineering Division, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)], E-mail: jdwang@iner.gov.tw; Lin Chaung [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China)

    2009-08-15

    An automatic boiling water reactor (BWR) loading pattern (LP) design methodology was developed using the rank-based ant system (RAS), which is a variant of the ant colony optimization (ACO) algorithm. To reduce design complexity, only the fuel assemblies (FAs) of one eight-core positions were determined using the RAS algorithm, and then the corresponding FAs were loaded into the other parts of the core. Heuristic information was adopted to exclude the selection of the inappropriate FAs which will reduce search space, and thus, the computation time. When the LP was determined, Haling cycle length, beginning of cycle (BOC) shutdown margin (SDM), and Haling end of cycle (EOC) maximum fraction of limit for critical power ratio (MFLCPR) were calculated using SIMULATE-3 code, which were used to evaluate the LP for updating pheromone of RAS. The developed design methodology was demonstrated using FAs of a reference cycle of the BWR6 nuclear power plant. The results show that, the designed LP can be obtained within reasonable computation time, and has a longer cycle length than that of the original design.

  10. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    International Nuclear Information System (INIS)

    organizations from 7 countries, SMART, integrated reactor, developed by Korea Atomic Energy Research Institute, Republic of Korea; CAREM, Argentina integrated reactor; MRX, integrated reactor, developed by Japan Atomic Energy Research Institute; UNITERM, NPP with integrated reactor, development by Research and development institute of power engineering (NIKIET), Russian Federation; AHEC-80, Russian NPP with integrated reactor, developed by OKB Mechanical Engineering (OKBM), Nizhny Novgorod, Russia. Moreover, following Boiling Water Reactor (BWR) projects have been subjected to the system comparative analysis. 1) Large Sized Reactors: ABWR, developed by Hitachi, Ltd, Japan, Toshiba Corporation, Japan ? G.E. USA; BWR-90, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; BWR-90+, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; SWR 1000, developed by Siemens Corporation, Germany; ESBWR, developed by General Electric Company, USA. 2) Medium Sized Reactors: SBWR, developed by General Electric Company, USA; HSBWR, developed by Hitachi Company, Ltd. 3) Small Sized Reactors: SSBWR, developed by Hitachi, Ltd, Japan; VK-300, BWR reactor, developed by Research and Development Institute of Power Engineering (NIKIET), Russia. Some data on the analysis of the condition and prospects of energy production and energy consumption, stations and networks in Kazakhstan are given. According to this analysis of nuclear power plants of average and low power are considered to be the most appropriate to construction in Kazakhstan. Recommendations on a choice of the most safe, reliable and economically competitive reactors have been made among the above-mentioned ones PWR, WWER and BWR for construction in Kazakhstan

  11. Local heat transfer from the corium melt pool to the boiling water reactor pressure vessel wall

    International Nuclear Information System (INIS)

    The present study considers in-vessel accident progression after core melt relocation to the lower head of a Boiling Water Reactor (BWR) and formation of a melt pool containing a forest of Control Rod Guide Tubes (CRGTs) cooled by purging flows. Descending streams of melt that flow along cooled surfaces of CRGT, and impinge on the bottom surface of the vessel wall can significantly increase local heat transfer. The area of enhanced heat transfer enlarges with decreasing of the melt Prandtl (Pr) number, while the peaking value of the heat transfer coefficient is a non-monotone function of Pr number. The melt Pr number depends on the melt composition (fractions of metallic and oxidic melt components) and thus is inherently uncertain parameter of the core melting and relocation scenarios. The effect of Pr number in the range of 1.02 - 0.03 on the local and integral thermal loads on the vessel wall is examined using Computational Fluid Dynamics (CFD). Heat transfer models obtained on the base of CFD simulations are implemented in the Phase-change Effective Convectivity Model (PECM) for simulation of reactor-scale accident progression heat transfer in real 3D geometry of the BWR lower plenum. We found that the influence of the low Pr number on the thermal loads in a big melt pool becomes more significant at later time, than rapid acceleration of the creep in the vessel wall. This result suggests that global vessel failure is insensitive to the melt composition in the considered 0.7 m deep melt pool configuration. However, it is not clear yet if the low Pr number effect has an influence on vessel failure mode in the other possible melt pool configurations. (author)

  12. Implementation of automated, on-line fatigue monitoring in a boiling water reactor

    International Nuclear Information System (INIS)

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to a Japanese operating boiling water reactor (BWR), Tsuruga Unit 1, is described. The system uses the influence function approach and rainflow cycle counting methodology, operates on a workstation computer, and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant-unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computes the fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification number-sign 501. Fatigue values are saved automatically on files at times defined by the user for use at a later time. Of particular note, this paper describes some of the details involved with implementing such a system from the utility perspective. Utility installation details, as well as why such a system was chosen for implementation are presented. Fatigue results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. Although the system is specifically set up to address fatigue duty for the feedwater nozzle location, a generic shell structure was implemented so that any other components could be added at a future time without software modifications. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension

  13. A nondiffusive solution method for RETRAN-03 boiling water reactor stability analysis

    International Nuclear Information System (INIS)

    This paper reports that boiling water reactors (BWRs) are susceptible to thermal-hydraulic instabilities that must be considered in BWR design and operation. Early BWRs were designed to be very stable while operating under natural-circulation conditions. As reactor designs have been modified, stability margins have been reduced, and the potential for stability events, such as occurred at the La Salle and Vermont Yankee plants, has increased. These events and other considerations point to the need for a reliable analysis tool for predicting the dynamic behavior of these events. Transient thermal-hydraulic systems analysis codes have been used to analyze hydrodynamic instabilities, and although the results are often reasonable and exhibit the expected behavior, they are sensitive to changes in node and time-step size and a converged solution cannot be demonstrated by reducing the node and time-step sizes. This sensitivity is due to numerical-diffusion that limits the use of most time domain system analysis codes for BWR stability analyses since it directly affects the decay (or growth) ratio compared for stability events. A conservation equation transport model using the method of characteristics has been developed for use with the RETRAN-03 mixture energy and vapor continuity equations. The model eliminates numerical diffusion in the RETRAN solution. The development and validation of a conservation equation transport model for the RETRAN-03 time domain thermal-hydraulic analysis code that extends the range of application to simulating the dynamic behavior of stability events are presented. RETRAN-03 analyses are presented that compare simulations of hydrodynamic instability events with data

  14. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author)

  15. Experience with General Electric's control rods for boiling water reactors

    International Nuclear Information System (INIS)

    Boron carbide undergoes significant volume increase with irradiation. The results are high stresses in the absorber containment. General Electric's Duralife control rods have used high purity stainless steel since 1983 to mitigate stress corrosion failure resulting from these high stresses. GE's Marathon control rod uses both stress corrosion resistant material and an absorber containment system which reduces the local stress resulting from boron carbide swelling. Hot cell examinations have demonstrated the stress corrosion resistance of high purity stainless steel and the importance of eliminating crevices from high neutron fluence and high stress regions. Material selection was also based on reducing control rod end of life activity. (orig.)

  16. Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

    International Nuclear Information System (INIS)

    Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

  17. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station

    International Nuclear Information System (INIS)

    This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs

  18. Reactor BR2

    International Nuclear Information System (INIS)

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  19. The effective convectivity model for simulation of molten metal layer heat transfer in a boiling water reactor lower head

    International Nuclear Information System (INIS)

    The paper is concerned with development of models for assessment of Control Rod Guide Tube (CRGT) cooling efficiency in Severe Accident Management (SAM) for a Boiling Water Reactor (BWR). In case of core melt relocation under a certain accident condition, there is a potential of stratified (with a metal layer atop) melt pool formation in the lower plenum. For simulations of molten metal layer heat transfer we are developing the Effective Convectivity Model (ECM) and Phase-change ECM (PECM). The models are based on the concept of effective convectivity previously developed for simulations of decay-heated melt pool heat transfer. The PECM platform takes into account mushy zone convection heat transfer and compositional convection that enables simulations of non-eutectic binary mixture solidification and melting. The ECM and PECM are validated against various heat transfer experiments for both eutectic and non-eutectic mixtures, and benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is applied to heat transfer simulation of a stratified heterogeneous debris pool in the presence of CRGT cooling. The PECM simulation results show no focusing effect in the metal layer on top of a debris pool formed in the BWR lower plenum and apparent efficacy of the CRGT cooling which can be served as an effective SAM measure to protect the vessel wall from thermal attacks and mitigate the consequences of a severe accident. (author)

  20. The calculation of boiling phenomena in fuel bundles in reactor safety studies using the subchannel thermohydraulics code SABRE

    International Nuclear Information System (INIS)

    This paper briefly describes the subchannel thermo-hydraulics code SABRE, with particular emphasis on the choice of two-phase modelling. The code has been extensively applied to the analysis of fault and severe accident situations in fast reactor cores, and in order to validate the code a range of accident simulation experiments has been analysed. The relative merits of slip models of boiling and full two-fluid representation are discussed, and results are presented comparing the two approaches. It is shown that in many situations the simpler ('3-equation') slip representation is adequate, but that there exist situations where the two-fluid ('6-equation') model is essential to even represent the physical phenomena. An example of such a situation is in wake regions, where vapour and liquid flows may be very different; in the paper this is discussed with particular reference to the flow past a blockage. In this example vapour accumulation may occur behind the blockage, and SABRE calculations of this situation are presented. (orig.)

  1. Effect of moderator density distribution of annular flow on fuel assembly neutronic characteristics in boiling water reactor cores

    International Nuclear Information System (INIS)

    The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. (author)

  2. Effect of gadolinium nitrate concentration on the corrosion compatibility of structural materials in a proposed Indian tube type boiling reactor

    International Nuclear Information System (INIS)

    Gadolinium (Gd3+) is added with nitric acid to moderator heavy water as a neutron poison in nuclear reactors to control reactivity and pH is maintained in the range of 5 to 5.5 to prevent Gd3+ precipitation. Usually ∼15 ppm of Gd3+ is used during actuation of secondary shutdown system and is subsequently removed on ion exchange up to a residual ∼2 ppm before start-up. In the moderator system of a proposed tube type boiling water reactor of Indian origin, a higher concentration (20-400ppm) of Gd(NO3)3 was proposed to be used in the emergency safety shutdown system. With higher concentration of Gd3+, the pH can go down and affect the radiolytic yields and thus affecting the integrity of the structural materials. Considering a long life of 100 years of operation for the proposed reactor, the concentration dependence of Gd3+ on the yields of molecular products like H2 and H2O2 during radiolysis and corrosion compatibility of structural materials like (1) SS 304 LN (proposed structural material for this reactor) and (2) SS 410 (proposed to be used in the valves of the moderator system as an alternative to hard facing alloy, colmonoy) is of interest. The pH and conductivity of the system were observed to be in the range of 5.33-3.76 and 50-870 μS/cm for 20-400 ppm of Gd3+. From the electrochemical studies it was observed that the electrochemical potential increased to more positive potential with increase in Gd3+ concentrations. The yield of H2 and H2O2 was also found to increase with increase in [Gd3+] concentrations. A detailed study on corrosion of the above said alloys at varying [Gd3+] concentrations, temperature, pH and simulated irradiation conditions and its effect on microstructure will be described in the paper. (author)

  3. Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

    International Nuclear Information System (INIS)

    Operating experience has proved so far that BWR control rods cannot be used for the total reactor life time as originally presumed, but instead has to be considered as a consumable article. After only few operating cycles, the mechanism of absorber failure has been shown to be neutron induced boron carbide swelling and stress cracking of the absorber tubes, followed by erosion of the absorber material. In the case that operation of such a control rod is continued in control cells, this can lead to an increase of the local power density distribution in the core and, under certain conditions, can even cause fuel rod damage. A non destructive testing method has been developed called 'UNDERWATER NEUTRON RADIOGRAPHY' applicable for any BWR control rod. 'Lead-control rods' being radiographed are used to evaluate their actual nuclear worth by the help of a special analytical procedure developed and verified by the author. Nuclear worth data plotted against bum up history data will allow to create an 'EMPIRIC MODEL'. This model includes the basic idea of operating control rods of a certain design first in a control position up to a target fluence limited to an amount just below the appearance of control rod washout. Afterwards they have to be moved in a shut down position to work therefor the total remaining holding period. The initial model is applicable to any CR-design as long as sufficient measuring-data and thus data about the nuclear worth are available. The results of these experiences are extrapolated to the whole reactor holding period. After modelling no further measurements of this particular control rod type are necessary in any reactor. The second focal point is to provide an APPROXIMATION EQUATION. By knowing the absorber radius, B4C density and absorber enclosure data an engineer will calculate reliably the working life of any control rod design on control position. indicated as maximum allowable neutron fluence margin until absorber wash-out starts. This

  4. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    OpenAIRE

    Venker, Jeanne

    2015-01-01

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO2) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO2 Brayton cycles, can be assessed as a retrofit measu...

  5. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  6. Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

    International Nuclear Information System (INIS)

    Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

  7. The IBR-2 test reactor

    International Nuclear Information System (INIS)

    Major design criteria, specifications and potential fields of application of the IBR-2 pulsed test reactor (now under construction in Dubna, USSR) are described. The pulsed power bursts will be due to fast periodic reactivity changes by a rotating reflector. The frequency of approximately 100 μs pulsed may be 5, 12.5 or 50 Hz. The IBR-2 reactor will be mostly profitable for slow neutron experiments when investigating solids, nuclei or neutrons themselves using spectroscopic methods. Due to the high peak flux of thermal neutrons (1016-1017 n/cm2xs) the reactor will be superior (for the sort of experiments) to the currently operating SM-2 and HFR high flux steady-state test reactors for many times

  8. Steady-state pool boiling heat transfer on nicr wire surface submerged in Al{sub 2}O{sub 3} nano-fluids

    Energy Technology Data Exchange (ETDEWEB)

    Dereje Shiferaw; Hyun Sun Park; Bal Raj Sehgal [Royal Institute of Technology, Department of Energy Technology, Division of Nuclear Power Safety, Drotning Kristinas Vag 33A, SE-10044, Stockholm (Sweden)

    2005-07-01

    Full text of publication follows: nano-fluids, or conventional liquids, e.g., water, with small concentration of nano-particles uniformly suspended, have attracted attention as a new heat transport medium with enhanced thermo-physical properties. Up to the present, only exploratory experiments on nano-fluids have been reported. Das et al (Int. J. Heat Mass Transfer 43, pp 3701-3707, 2003) conducted boiling experiments with water containing 38 nm Al{sub 2}O{sub 3} nano-particles. They observed deterioration in the nucleate boiling heat transfer due to the deposition of nano-particles. Boiling experiments conducted by Vassallo et al (Int. J. Heat Mass Transfer 47, pp 407-411, 2004) using silica nano-fluid using 0.4 mm diameter NiCr wire showed three times higher critical heat flux (CHF) and the wire traversed the film boiling region before it failed. Another independent experiment performed on 1 cm{sup 2} square plate with a very low concentration of nano-particles ranging from 0.01 to 0.05 g/liter and at under pressure (2.89 psia), nano-fluids resulted in drastic 2{approx}3 times enhancement of the CHF (You and Kim, Appl. Phys. Lett. 83. No 16, 2003). However in all the aforementioned studies no appropriate explanation of the CHF enhancement has been advanced. The measured 2-3 times higher critical heat flux for very dilute nano-fluids may have high significance if such nano-fluids could be employed in heat transport systems. Recently, we investigated the effect of nano-particles on film boiling, which governs heat transfer during accident conditions in a reactor plant, e.g., in coolability of a degraded core, or a particulate debris bed or a core melt, and in steam explosions. Our previous experiments performed on film boiling in nano-fluids having larger concentrations of 5, 10, and 20 g/liter than those in You's experiments showed that the nano-fluids lower the film boiling temperature, decrease the film boiling heat transfer and provide a much thicker and

  9. First experience with the new solid methane moderator at the IBR-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Beliakov, A.A.; Shabalin, E.P. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Tretyakov, I.T. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2001-03-01

    In the 1999 Fall the solid methane moderator (CM) has been installed and tested at full power at the IBR-2 pulsed reactor. Its main features are a beryllium reflector and a light water premoderator. Radiation load on the methane was three times as much as that of IPNS facility, namely, 0.1 W/g. Effects of temperature, operation time, concentration of a hydrogen scavenger, and annealing procedure on both neutron and service performances were studied. Maximum operation time of a newly loaded portion of methane was 4 days. In this time around 30% of methane is transformed into hydrogen, ethane, and high molecular hydrocarbons, and yet no deterioration in cold neutron intensity was detected. Among new knowledges, the most important are two facts observed: two-fold decrease in hydrogen formation rate when methane is poisoned with 2.5% to 5% of ethylene, and low formation rate of solid, inremovable products of radiolysis - (1.5/3)10{sup -7} g/J, which means that after 10 years of operation the methane chamber will be filled with only 100 g of residue. Gain of factor 20 in cold neutron flux was obtained as compared to the routine grooved light water moderator. Presently, it is the highest among the intense pulsed neutron sources. (author)

  10. Boiling Water Reactor Turbine Trip (TT) Benchmark - Vol. IV Volume IV: Summary Results of Exercise 3

    International Nuclear Information System (INIS)

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling of the core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for that purpose. The present volume is the last in a series of four and summarises the results of the third benchmark exercise, which analyses a turbine trip (TT) in a BWR in its entirety, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. Exercise 3 also analyses four extreme scenarios which allowed participants to test the capabilities of their code(s) in terms of coupling and feedback modelling. The data made available from experiments carried out at the plant make the present benchmark particularly valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

  11. Investigation of space- and time-dependent processes in the core of a boiling water reactor with a modular computer model

    International Nuclear Information System (INIS)

    A modular, one-dimensional space-dependent model for determination of the transient behavior in the core of a boiling water reactor was developed for use with an available program system for reactor calculations. It consists of the neutron-kinetic module DYNMOD, the thermohydraulic modules DYNSID 1-3 and the module STATEMP for investigation of the stationary initial values. The coupling between the thermohydraulic and the neutron-kinetic processes is ensured by direct manipulation of the effective cross-sections. Two experiments obtained during the testing of the nuclear plant in Wuergassen were computed with this model to gain information on its capability. Satisfactory results were obtained and the influence of space dependence could be studied using the model. (orig.)

  12. Nuclear-coupled thermal-hydraulic stability analysis of boiling water reactors

    Science.gov (United States)

    Karve, Atul A.

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model we developed from: the space-time modal neutron kinetics equations based on spatial omega-modes, the equations for two-phase flow in parallel boiling channels, the fuel rod heat conduction equations, and a simple model for the recirculation loop. The model is represented as a dynamical system comprised of time-dependent nonlinear ordinary differential equations, and it is studied using stability analysis, modern bifurcation theory, and numerical simulations. We first determine the stability boundary (SB) in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value and then transform the SB to the practical power-flow map. Using this SB, we show that the normal operating point at 100% power is very stable, stability of points on the 100% rod line decreases as the flow rate is reduced, and that points are least stable in the low-flow/high-power region. We also determine the SB when the modal kinetics is replaced by simple point reactor kinetics and show that the first harmonic mode has no significant effect on the SB. Later we carry out the relevant numerical simulations where we first show that the Hopf bifurcation, that occurs as a parameter is varied across the SB is subcritical, and that, in the important low-flow/high-power region, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line. Hence, a point on the 100% rod line in the low-flow/high-power region, although stable, may nevertheless be a point at which a BWR should not be operated. Numerical simulations are then done to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is determined that the NRC requirement of DR flow/high-power region and hence these points

  13. CHF Enhancement of SiC-water nanofluids in Pool Boiling Experiment

    International Nuclear Information System (INIS)

    SiC nanofluids were used for Critical heat flux(CHF) enhancement in the case of water pool boiling. Many kinds of nanofluids have been highlighted as a simple way to gain high thermal performance of fluids. Also, one of the ceramic particle, SiC is received attention these days as a promising material because of its relatively high thermal properties. In this study, SiC nanofluids were investigated to measure its thermal performance in water pool boiling experiment especially for CHF. The volume concentration of SiC nanofluids were 0.0001%, 0.001%, 0.01%. Several characteristic of SiC nanofluids, such as Zeta potential, and contact angle which could be affect on thermal performance of the fluids had been measured. The experiments were conducted under atmospheric pressure. The CHF has been enhanced upto 53.1% at volume concentration 0.01% SiC nanofluids

  14. Cryogenic system for collecting noble gases from boiling water reactor off-gas

    International Nuclear Information System (INIS)

    In boiling water reactors, noncondensible gases are expelled from the main condenser. This off-gas stream is composed largely of radiolytic hydrogen and oxygen, air in-leakage, and traces of fission product krypton and xenon. In the Air Products' treatment system, the stoichiometric hydrogen and oxygen are reacted to form water in a catalytic recombiner. The design of the catalytic recombiner is an extension of industrial gas technology developed for purification of argon and helium. The off-gas after the recombiner is processed by cryogenic air-separation technology. The gas is compressed, passed into a reversing heat exchanger where water vapor and carbon dioxide are frozen out, further cooled, and expanded into a distillation column where refrigeration is provided by addition of liquid nitrogen. More than 99.99 percent of the krypton and essentially 100 percent of the xenon entering the column are accumulated in the column bottoms. Every three to six months, the noble-gas concentrate accumulated in the column bottom is removed as liquid, vaporized, diluted with steam, mixed with hydrogen in slight excess of oxygen content, and fed to a small recombiner where all the oxygen reacts to form water. The resulting gas stream, containing from 20 to 40 percent noble gases, is compressed into small storage cylinders for indefinite retention or for decay of all fission gases except krypton-85, followed by subsequent release under controlled conditions and favorable meteorology. This treatment system is based on proven technology that is practiced throughout the industrial gas industry. Only the presence of radioactive materials in the process stream and the application in a nuclear power plant environment are new. Adaptations to meet these new conditions can be made without sacrificing performance, reliability, or safety

  15. A pilot study for errors of commission for a boiling water reactor using the CESA method

    International Nuclear Information System (INIS)

    Probabilistic Safety Assessment (PSA) typically focuses on the errors leading to the non-performance of required actions (Errors of Omission, EOOs). On the other hand, Errors Of Commission (EOCs) refer to inappropriate, undesired actions that aggravate an accident scenario. The challenges to their treatment in PSA relate to both their identification (which error events should be included in the PSA) and to the quantification of their probabilities. This paper presents the results from a plant-specific study to identify potential EOC vulnerabilities and quantify their risk significance. The study addresses a Boiling Water Reactor (BWR) in Switzerland. It is one of the first EOC analyses ever made for BWRs. The Commission Error Search and Assessment (CESA) method was used to identify EOC scenarios. The EOC probabilities were estimated using the elicitation approach developed as part of the ATHEANA method (A Technique for Human Event Analysis), with input from interviews with plant personnel (with oral as well as written questions). The basis for the quantification was a qualitative analysis of the scenario, the operator response and its procedural basis, and of the opportunities for the EOC and its recovery. The results suggest that the contribution to risk of the most important EOCs is comparable to that of the most important errors of omission, i.e. the required actions typically treated in a PSA; thus, they highlight the significance of EOCs in the overall risk profile of the plant. This study demonstrates the feasibility of a systematic treatment of EOCs for large-scale applications and contributes to understanding the importance of EOCs in the plant risk profile.

  16. Corrosion Products Identification at Normal Water and Hydrogen Water Chemistry in Boiling Water Reactors

    International Nuclear Information System (INIS)

    The corrosion products sampled from condensate and feedwater systems of boiling water reactors (BWRs) at normal water chemistry (NWC) and hydrogen water chemistry (HWC) operating condition were analyzed with dissolution and instrumental simulation methods. The crystallite and amorphous of iron oxides were separated by means of dissolving method with appropriate chemical solution. The iron oxide composition and content were analyzed by X-ray diffraction (XRD) and inductively coupled plasma atomic emission spectrometer (ICP-AES) in this study. The insoluble iron oxides were obtained in influent and effluent of condensate demineralizer comprised mostly crystalline structure of hematite, magnetite and non-crystallite form of amorphous at NWC and HWC environments. Both goethite and lepidocrocite compositions are of minor importance in feed water system. Crystallite and amorphous compositions in the samples will be calculated from the new developing dissolution method. The crystalline phase of corrosion products are varied with water chemistry conditions in BWRs. The oxide characterization of system corrosion products includes compositions, morphology and particle size can effectively provide the ways of solving crud removal problem in different condition for the performance of condensate demineralizer. The feasibility of identifying other iron oxides and hydroxides in corrosion products is briefly discussed and the mechanisms of iron oxide formation formed around BWR piping will also be shown in detail in this report. Moreover, it will be figured out the properties of radioactive corrosion products growing in different operation periods. The results can also assist in plant units to improve the crud reduction countermeasures and to optimize the system water chemistry. (authors)

  17. The installation of PEANO at the Halden Boiling Water Reactor: first test and results

    International Nuclear Information System (INIS)

    After extensive testing of PEANO with data from process simulators, the next step was to set-up an installation in a real process, where the signal validation is performed online. For this purpose the Halden Boiling Water Reactor was used. One implication is that recorded process data from past operation would be used for the training of the system. This type of data is corrupted with errors, faults, process noise or even previous sensor problems, which need to be removed before the data can be used. The pre-processing was therefore a very import step during this installation. A 15 minute average of 29 process signals, spread out over the primary, secondary and tertiary loops, was used. At the end of the design process a fuzzy-neural network resulted containing 5 clusters, that has been trained with over 20.000 patterns. To establish the TCP/IP connection to the process computer and receiving the process data in real-time, some extra software was developed. With this installation it has been shown that it is possible to have the PEANO Server and PEANO Client (monitoring unit) running on one machine (e.g. in the control room), while additional monitoring units are connected from a remote location (e.g. main office building). The first results show that the installation of PEANO is capable of performing its validation task properly, even during transients. Both start-up and shutdown situations can be handled without any problems. In situations where incoming patterns represent unknown process situations that have not been encountered during the training, the 'I don't know' answer was given. To test the ability to detect a sensor failure off-line tests have been run, where sensor faults and drifts were added (author) (ml)

  18. Factors influencing the precoat filtration of boiling water reactor water streams

    International Nuclear Information System (INIS)

    A series of studies on precoat filtration were carried out on condensate and preheater drains in the Swedish and Finnish boiling water reactors (BWRs). The goal was to increase knowledge about the precoat filtration process and to find physical and chemical means to improve the performance of the precoat filters in the condensate polishing plants. To achieve this goal a number of parameters, such as type of resin, bed depth, pH, oxygen and organic contaminant concentrations (measured total organic carbon), and corrosion product particle characteristics, were selected for the study. The work was mainly carried out in the power plants using an experimental facility fed with on-line sampled condensates and drains taken from the plant sampling lines. The main results are that there is a varying influence on precoat filtration from all the aforementioned parameters. The oxygen concentration, the concentration of organic contaminants, and the type of corrosion products are, however, the factors that have the strongest influence within the parameter ranges that are representative for BWR operation. The results are rather similar when the different units are compared. There are, however, some differences that could be mainly attributed to deviations in operation parameters and the subsequent differences in the corrosion product spectra. The mechanism for precoat filtration of corrosion products in BWR condensate is complex. The filtration behavior is to a large extent governed by competition between depth filtration and electrostatic interactions. During the early stages of the filtration cycle, electrostatic interaction is of great importance, whereas depth filtration becomes more important with increasing operating time. Rapid pressure drop buildup rates have been demonstrated to be caused by the presence of amorphous corrosion products. An effect from the presence of organic contaminants has been found, although this should be of little significance

  19. Experimental investigation on critical heat flux and transition boiling of water flow under increased pressure

    International Nuclear Information System (INIS)

    In connection with reactor safety problems (LOCA) a measuring technique has been developed which enables, within the parameter range of medium pressure (0.11 MPa - 1.20 MPa) and low mass flow densities (10 kg/m2s - 500 kg/m2s), exact experimental investigations of critical heat flux and transition boiling of water under quasi-stationary conditions. The system consists of a vertical, temperature-controlled short test section with water flowing upwards inside; an experimental loop controllable to a large extent; a quick automatic data acquisition, and numeric evaluation procedures. Quasi-stationary measured boiling curves, from nucleate boiling to film boiling (circa 450deg C), demonstrate the importance of pressure, mass flow density, and inlet subcooling, the boiling pressure being the most important parameter. The linear course of the boiling curves during transition boiling is remarkable. A frequently suspected hysteresis of the boiling curve could not be detected. The influence of surface effects (contact angle) clearly decreases with increasing pressure. For the empirical correlation of the measured data by means of indices, a statement was chosen that normalizes the heat flux density of transition boiling to the maximum heat flux density at the beginning of the post-CHF range. As a result, the experimental data obtained, and the correlation developed from them, show a better heat transfer in transition boiling than conservatively assumed in general in literature. The temperature-controlled measurements of complete boiling curves supply data for critical heat flow density and the corresponding wall overheating. A comparison with the uncontrolled operation of the test section shows differences of 5-6% only within the range of measurement accuracy of such experiments. (orig.) With 36 figs., 5 tabs

  20. Bubble spreading during the boiling crisis: modelling and experimenting in microgravity

    OpenAIRE

    Nikolayev, Vadim; Beysens, D.; Garrabos, Yves; Lecoutre, Carole; Chatain, D.

    2006-01-01

    International audience Boiling is a very efficient way to transfer heat from a heater to the liquid carrier. We discuss the boiling crisis, a transition between two regimes of boiling: nucleate and film boiling. The boiling crisis results in a sharp decrease in the heat transfer rate, which can cause a major accident in industrial heat exchangers. In this communication, we present a physical model of the boiling crisis based on the vapor recoil effect. Under the action of the vapor recoil ...

  1. Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

    International Nuclear Information System (INIS)

    An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have a strong influence on the Type I and Type II instabilities observed at low and high channel powers, respectively. Also, it was found that the coupled multipoint kinetics model and the modal point kinetics model predict the same threshold power for out-of-phase instability if the coupling coefficient in the former model is half the eigen value separation between the fundamental and the first harmonic mode in the latter model. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (orig.)

  2. Nuclear reactor noise investigations on boiling effects in a simulated MTR-type fuel assembly

    International Nuclear Information System (INIS)

    The work includes validation/testing of existing neutron noise methods under well-controlled circumstances, investigation of boiling phenomena in narrow channels, and development of a novel boiling monitoring method. The work has been performed in the NIOBE facility at the HDR. Noise signals of thermocouples in the channel wall are used for velocity profile monitoring. Flow patterns in the boiling coolant are identified by means of analysis of probaof probability density functions and neutron noise spectra. Local noise effects are studied. (DG)

  3. An in-line diffuse reflection spectroscopy study of the oxidation of stainless steel under boiling water reactor conditions

    International Nuclear Information System (INIS)

    A novel cell unit was constructed to measure in-line the oxide layer build-up on a stainless steel sample by Diffuse Reflection Spectroscopy (DRS; ultraviolet, visible, near infrared) under boiling water reactor (BWR) conditions. The stainless steel samples, observed in the cell through a sapphire window, are contacted with oxidising hot water (300oC, 9.0 MPa). Using a cold finger with the optical fibre probe, the spectroscopic investigations (200-1000 nm wavelength) were performed at a fixed position from the sapphire window. The DRS spectra are the result of the coupling of both absorption (chemical) and interferometric (physical) processes. Analysis of these spectral components allows the independent determination of the oxide layer thickness. The build-up of the oxide layer may be directly observed and quantified, nanometre by nanometre, from 2 to 200 nm. This powerful technique may be used to study the early corrosion rates of stainless steel under BWR conditions and should allow the development of a strategy to reduce corrosion. (Author)

  4. A novel approach for noble metal deposition on surfaces for IGSCC mitigation of boiling water reactor internals

    International Nuclear Information System (INIS)

    A novel in-situ approach has been developed to deposit noble metals on surfaces of materials commonly used in the nuclear power generating industry. The method involves the injection of a noble metal chemical solution directly into the high temperature water that is in contact with a metal surface to be coated with the noble metal. An effective noble metal coating on a surface can be achieved by maintaining the noble metal concentration at a level of 10 to 100 ppb over a period of 48 hours during the injection process. The surface concentration of the noble metal after the treatment was 2 to 3 atomic %, and the noble metal was present to a depth of 200 to 500 A. The concept of noble metal chemical addition (NMCA) technology was successfully used to create a ''noble metal like'' surface on three of the major nuclear materials, 304 SS, Alloy 600 and Alloy 182. The success of this technology was demonstrated by using constant extension rate tensile (CERT) tests, crack growth rate (CGR) tests and electrochemical corrosion potential (ECP) response tests. The NMCA technology in combination with hydrogen has successfully decreased the ECP of surfaces below the critical cracking potential of -0.230 V(SHE), and prevented both crack initiation and crack propagation in simulated boiling water reactor (BWR) environments

  5. Study of the oxide layer formed on stainless steel exposed to boiling water reactor conditions by ion beam techniques

    Science.gov (United States)

    Degueldre, C.; Buckley, D.; Dran, J. C.; Schenker, E.

    1998-01-01

    The build-up of the oxide layer on austenitic steel under boiling water reactor (BWR) conditions was studied by macro- and micro-Rutherford backscattering spectrometry (RBS) and sputtered neutral mass spectroscopy (SNMS). RBS is applicable when the oxide thickness is larger than 20 nm and yields both the layer thickness and its stoichiometry. SNMS provides elemental depth profiles and the oxide thickness when combined with profilometry. Stainless steel strip samples pre-treated (electro- or mechanically polished) or not, exposed in a loop simulating the BWR-conditions for periods ranging from 31 to 291 days and with a low water flow velocity show oxide layers with a thickness of about 300 to 600 nm. There is no significant increase of the oxide layer thickness after 31 days of exposure. The paper confirms the presence of inner and outer oxide layers and also confirms the stoichiometry M 2O 3 in the external part in contact with the oxygenated water. The oxide layer consists not only of an outer layer and an inner layer but also of a deep apparent oxide/metal interface that is attributed to oxide formation through the steel grain boundaries.

  6. Scaling laws and design aspects of a natural-circulation-cooled simulated boiling water reactor fuel assembly

    International Nuclear Information System (INIS)

    In order to study the thermohydraulic behavior of a natural-circulation-cooled boiling water reactor (BWR) fuel assembly, such as void drift, flow pattern distribution, and stability, a scaled loop geometry is designed. For modeling the steam/water flow in a BWR fuel assembly, scaling criteria are derived using the one-dimensional drift-flux model. Thermal equilibrium and subcooled boiling conditions are treated separately, resulting in one overall set of criteria. Scaling on all flow regimes that can be present in a normal fuel assembly leads to fixing both the assembly mass flux and the geometric dimensions. When Freon-12 is used as a modeling fluid, model assembly dimensions must be 0.46 of the prototype. Total power consumption must be reduced by a factor 50. To sustain cooling by natural circulation, a modeled chimney and downcomer are included

  7. Pool boiling heat transfer characteristics of vertical cylinder quenched by SiO{sub 2}-water nano-fluids

    Energy Technology Data Exchange (ETDEWEB)

    Abdurrahim, Bolukbasi; Dogan, Ciloglu [Department of Mechanical Engineering, Ataturk University, Erzurum (Turkey)

    2011-06-15

    This study includes an experimental investigation of the pool boiling heat transfer characteristics of a vertical cylinder quenched by SiO{sub 2}-water nano-fluids. The experiments are performed through a cylindrical rod, at saturated temperature and atmospheric pressure. As the coolant, pure water and SiO{sub 2}-water nano-fluid suspensions at four different concentrations (0.001, 0.01, 0.05 and 0.1 vol.%) are selected. The test specimen heated at high temperatures is plunged into cooling fluids at saturated conditions in a pool. The cooling curves are obtained via taking the temperature-time data of the specimen into account. The experimental results indicate that the pool film boiling heat transfer in nano-fluids is identical to that in pure water. However, during the repetition tests in nano-fluids with high concentrations, the film boiling region disappears, and the critical heat flux increases. In addition, the nucleate pool boiling heat transfer coefficient decreases compared with that of pure water, but a considerable decrease in nucleate pool boiling heat transfer is not observed with the repetition tests. A change in surface characteristics due to the deposition of nano-particles on the surface has a major effect on the quenching process. (authors)

  8. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    International Nuclear Information System (INIS)

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates

  9. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water...

  10. Implementation of the HELIOS code for the Halden boiling water reactor (HBWR)

    International Nuclear Information System (INIS)

    Since the beginning of 1998 the 2-D transport code HELIOS has been used for neutronic physics calculations at the Institutt for energiteknikk. This note describes the system which has been built up to enable HELIOS to quickly and accurately perform the calculations required and is a concise review of the HELIOS system at the Halden Reactor Project. The Halden Reactor is a small but diverse test reactor. A brief outline of the core and its contents is given to give an indication of the complexity of some of the test rigs to be found in this reactor. HELIOS is a neutron and gamma transport code. A brief outline of the code itself is given as well as a description of its implementation at Halden. HELIOS is a very versatile code resulting in many applications. The majority of its uses so far are described in this note. The most common application is the calculation of fuel depletion functions which are used for power determination at Halden. One of the problems of accurately modelling the HBWR is what fraction of the reactor needs to be modelled. An example of a fuel depletion calculation is given to demonstrate the effects of varying the size of model used. This then leads to the minimum size required to accurately model the surroundings of a fuel assembly

  11. Measurement of Effective $\\Delta m_{31}^2$ using Baseline Differences of Daya Bay, RENO and Double Chooz Reactor Neutrino Experiments

    CERN Document Server

    Bezerra, T J C; Suekane, F

    2012-01-01

    In 2011 and 2012, three reactor neutrino experiments, Double Chooz, Daya Bay and RENO showed positive signals of reactor neutrino disappearance and measured a mixing parameter sin^2(2th13) at average baselines 1.05, 1.65 and 1.44km, respectively. It is possible to measure effective Dm31^2 from distortion of neutrino energy spectrum (E dependence of the oscillation) in those experiments. However, since it requires a precise energy calibration, such measurements have not been reported yet. Dm31^2 can also be measured from baseline (L) dependence of the neutrino oscillation. In this paper, Dm31^2 is measured from disappearance probabilities of the three reactor experiments which have different baselines, to be 2.99^{+1.13}_{-1.58}(^{+0.86}_{-0.88}) X 10^{-3}eV^2, where the errors are two (one) dimensional uncertainties. This is consistent with Dm32^2 measured by \

  12. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

    2015-05-15

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

  13. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    International Nuclear Information System (INIS)

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

  14. Acoustics of the Sodium Loop Safety Facility P2 experiment

    International Nuclear Information System (INIS)

    In the highly instrumented P2 experiment to simulate an unprotected loss of flow in an LMBFR core, two high-temperature sodium-immersed microphones were used to obtain the acoustic scenario of the 9-second flow coastdown, boiling sodium coolant, fuel failure, and re-establishment of flow after reactor scram. Acoustic data from this and previous experiments indicate that singular larger voids are formed in the sodium boiling process. These singular larger voids were observed to collapse upon contacting the subcooled sodium or structures, at repetition rates comparable to formation rates of sodium vapor bubbles. The acoustic observations were made in the presence of sodium exposed to argon cover gas at LMFBR operating conditions, and in the presence of fission gas from fuel pin failure. 6 refs

  15. PROMILLE database as a part of JNC reactor physics analytical system for BFS-2 fast critical facility experiments analysis

    International Nuclear Information System (INIS)

    The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS Visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-1i1 uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core. (author)

  16. Benchmark analysis of reactivity experiment in the TRIGA Mark 2 reactor using a continuous energy Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    A good model on experimental data (criticality, control rod worth, and fuel element worth distributions) is encouraged to provide from the Musashi-TRIGA Mark 2 reactor. In the previous paper, as the keff values for different fuel loading patterns had been provided ranging from the minimum core to the full one, the data would be candidate for an ICSBEP evaluation. Evaluation of the control rod worth and fuel element worth distributions presented in this paper could be an excellent benchmark data applicable for validation of calculation technique used in the field of modern research reactor. As a result of simulation on the TRIGA-2 benchmark experiment, which was performed by three-dimensional continuous-energy Monte Carlo code (MCNP4A), it was found that the MCNP calculated values of control rod worth were consisted to the experimental data for both rod-drop and period methods. And for the fuel and the graphite element worth distributions, the MCNP calculated values agreed well with the measured ones though consideration of real control rod positions was needed for calculating fuel element reactivity positioned in inner ring. (G.K.)

  17. Integrated plant safety assessment: Systematic Evaluation Program. LaCrosse Boiling Water Reactor, Dairyland Power Cooperative, Docket No. 50-409

    International Nuclear Information System (INIS)

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addresed. Equipment and procedural changes have been identified as a result of the review

  18. Integrated plant safety assessment. Systematic Evaluation Program. La Crosse Boiling Water Reactor. Dairyland Power Cooperative, Docket No. 50-409. Final report

    International Nuclear Information System (INIS)

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review

  19. Investigations on the extremely low retention of 131I by an iodine filter of a boiling water reactor

    International Nuclear Information System (INIS)

    An extremely low retention was observed of the I-131 contained in the exhaust air, by an iodine filter of a boiling water reactor. After filling the filter with fresh KI impregnated activated carbon (8-12 mesh), the decontamination factor dropped to about 1 within a few days. The extremely low retention of the I-131 was due to the occurrence of unidentified I-131 species in high proportions. By increasing the residence time to about 1 s and using a KI impregnated activated carbon of a smaller size, a somewhat higher retention can be achieved

  20. Acoustic Analysis for a Steam Dome and Pipings of a 1,100 MWe-Class Boiling Water Reactor

    International Nuclear Information System (INIS)

    For the integrity evaluation of steam dryers in up-rated nuclear power plants, we have applied acoustic analysis to a nuclear power plant steam dome and main steam pipings. We have selected a 1,100 MWe-class boiling water reactor as a subject of the analysis. We have constructed a three-dimensional finite element model, and conducted acoustic analyses. The analysis result suggested that the origin of steam pressure pulsation in high frequency range was due to vortex shedding at standpipes. (authors)

  1. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  2. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  3. Development of a fully-consistent reduced order model to study instabilities in boiling water reactors

    International Nuclear Information System (INIS)

    A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to other existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or Cmn*V,D - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good

  4. Experimental validation of the subchannel two-fluid model code SABENA with out-of-pile sodium boiling experiments

    International Nuclear Information System (INIS)

    Following a description of a two-fluid eight-equation system for sodium boiling and two-phase flow analysis, a discussion is made on the selected constitutive equations. A significance of the droplet entrainment effects is illustrated based on a one-dimensional sodium boiling experiment. Also the appropriateness of using the one-dimensional two-phase flow correlations is discussed when they are applied to the two-fluid subchannel method which is designed to analyze the three-dimensional phenomena. The eight equation two-fluid subchannel code SABENA with the droplet transport model was applied to the KfK 37-pin bundle loss-of-flow sodium boiling experiments with different power profiles. The computational results were in good agreement with the experiments and explained the three-dimensional sodium boiling phenomena involved in the loss-of-flow transients in pin bundles. (orig.)

  5. Experiment and RELAP5 analysis for the downcomer boiling of APR1400 under LBLOCA

    International Nuclear Information System (INIS)

    Experiment has been carried out to investigate the boiling phenomena in the downcomer and RELAP5/MOD3.2 has been assessed with the present experimental data. The heated wall with a thickness of 8.2 cm and a height of 27 cm is used, which is made of the same material as the prototype (APR1400) with chrome coating against rusting. From the experiment, we visually observed strong liquid recirculation and vapor jetting near the heated wall due to the axial migration of voids only in the thin layer of the heated wall but little bubble migration to the bulk region. The size of the thin layer is about 4 cm, which is used for the determination of the radial nodal size in the radial double-node schemes. The RELAP5 calculations using three different nodal schemes are compared with experimental data in aspects of water level, void fraction, wall temperatures and phase velocities. The radial single-node scheme produces no liquid recirculation, resulting in the sudden level drop due to a sudden increase in void fraction. The double-nodal scheme with the top-bottom radial connections yields the strong circulation, eliminating the sudden level drop. As a result, the scheme produces better results than the radial single-nodal scheme and the double-nodal scheme with all radial connections. Based on the information from the measurement of local liquid velocity profile and visual observation, the drift velocity model is developed to apply into the downcomer with a large gap and a vertical heated wall. The proposed drift velocity model has been implemented into RELAP5 and verified with the experimental results. (author)

  6. Characterization and modeling of the thermal hydraulic and chemical environment of fuel claddings of PWR reactors during boiling

    International Nuclear Information System (INIS)

    In pressurised water reactors (PWR), nucleate boiling can strongly influence the oxidation rate of the fuel cladding. To improve our understanding of the effect of the boiling phenomenon on corrosion kinetics, information about the chemical and thermal hydraulic boundary conditions at the heating rod surface is needed. Moreover, very few data are available in the range of thermal hydraulic parameters of PWR cores (15,5 MPa and 340 deg C) concerning the two-phase flow pattern close to the fuel cladding. A visualization device has been adapted on an out-of-pile loop Reggae to obtain both qualitative and quantitative data. These observations provide a direct access to the geometrical properties of the vapor inclusions, the onset of nucleate boiling and the gas velocity and trajectory. An image processing method has been validated to measure both void fraction and interfacial area concentration in a bubbly two-phase flow. Thus, the visualization device proves to be a suitable and accurate instrumentation to characterize nucleate boiling in PWR conditions. The experimental results analysis indicates that a local approach is needed for the modelling of the fuel rod chemical environment. To simulate the chemical additives enrichment, a new model is proposed where the vapor bubbles are now considered as physical obstacles for the liquid access to the rod surface. The influence of the two-phase flow pattern appears to be of major importance for the enrichment phenomenon. This study clearly demonstrates the existence of strong interactions between the two-phase flow pattern, the rod surface condition, the corrosion process and the water chemistry. (author)

  7. Nucleate Boiling from Smooth and Rough Surfaces--Part 2: Analysis of Surface Roughness Effects on Nucleate Boiling

    OpenAIRE

    McHale, John P.; Garimella, Suresh V.

    2013-01-01

    The effect of surface roughness on nucleate boiling heat transfer is not clearly understood. This study is devised to conduct detailed heat transfer and bubble measurements during boiling on a heater surface with controlled roughness. This second of two companion papers presents an analysis of heat transfer and bubble ebullition in nucleate boiling with new measures of surface roughness: area ratio, surface mean normal angle, and maximum idealized surface curvature. An additional length scale...

  8. LWR-PROTEUS Verification of Reaction Rate Distributions in Modern 10 x 10 Boiling Water Reactor Fuel

    International Nuclear Information System (INIS)

    HELIOS, CASMO-4, and MCNP4B calculations of reaction rate distributions in a modern, fresh 10 x 10 boiling water reactor fuel element have been validated using the experimental results of the LWR-PROTEUS Phase I project corresponding to full-density water moderation conditions (core 1B). The reaction rate distributions measured with a special gamma-scanning machine employing twin germanium detectors consisted of total fission Ftot and 238U-capture C8. The average statistical errors for the gamma scans were better than 0.5% for Ftot and 0.9% for C8. The rod-by-rod measurements were performed on 60 different fuel rods selected from the central part of a test zone consisting of actual, fresh SVEA-96+ fuel elements, thus gaining in realism by departing from conventional fuel rod mockups. In the case of Ftot, the root-mean-square (rms) of the rod-by-rod distribution of differences between calculational and experimental (C-E) values has been found to be 1.1% for HELIOS and for CASMO-4, and 1.3% for MCNP4B. For C8, the rms values of the (C-E) distributions are 1.0, 1.3, and 1.4% as obtained with HELIOS, CASMO-4, and MCNP4B, respectively. The effects of using different data libraries (ENDF/B-V, ENDF/B-VI, and JEF-2.2) with MCNP4B were also studied and have been found to be small

  9. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  10. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    International Nuclear Information System (INIS)

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  11. Water quality control in primary cooling system of crud concentration suppressed boiling water reactor, (1)

    International Nuclear Information System (INIS)

    The No.2 Unit of Fukushima-Daini Nuclear Power Plant (2F-2; 1,100 MWe) was commercially operated for 10,320 effective full power hours (EFPH) as its first fuel cycle. The basic design concept of the 2F-2 incorporated the following two features : (1) Application of procedures for reducing shutdown dose rate based on the Japanese Improvement and Standardization Program, (2) Low crud generation to minimize radioactive waste by careful material selection for the primary system. Thus, it was possible to keep the average Fe concentration in the condensate water at less than 6 ppb during the first fuel cycle. As a result of this low value, the average life of powdered resin precoated prefilters was extended to about a month, and the average chemical regeneration period of the deep bed demineralizers was extended to more than one year. The water chemistry of the 2F-2 was characterized by low 60Co and high 58Co radioactivities in the reactor water, which resulted in a low shutdown dose rate determined mainly by 58Co depositing on the primary piping. For example, average dose rate around the primary piping just after reactor shutdown was about 70 mR/h, about 75 % of which was from 58Co depositing on the pipe inner surfaces. The contribution of 60Co was about 25 %. (author)

  12. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  13. Header feedwater supply and power distribution stability in channel boiling water cooled reactors

    International Nuclear Information System (INIS)

    Boundaries of radial-azimuthal instability of the reactor neutron field during the supply of all feedwater and a part of it (25%) to downtake pipes of the separating drum (75% of feedwater come to distributive group headers) are found out for NPP with a RBMK type reactor. Results of computer calculation of the transient process at NPP caused by 2% step increase of nominal pressure in a head collector of a feedwater electric pump are also presented for comparison of the above methods of feed-water supply. Calculation is carried out according to the OKA program with provision for the control system of the reactor total power. It is shown that the boundary of ''mean period'' instability does not change but the reserve in respect to the ''fast'' space instability slightly increases when header feedwater supply at NPP from RBMK is used. It is noted that requirements to the pressure regulator system quick action in a separating drum are increased when the header feedwater supply is used. This fact is explained by the fact that considerable pressure drop in a separating drum occurs during some accidents (for example, at false operation of the emergensy protective system)

  14. Measurement of neutrino mixing angle θ13 and mass difference Δ mee2 from reactor antineutrino disappearance in the RENO experiment

    Science.gov (United States)

    Kim, Soo-Bong

    2016-07-01

    RENO (Reactor Experiment for Neutrino Oscillation) made a definitive measurement of the smallest neutrino mixing angle θ13 in 2012, based on the disappearance of reactor electron antineutrinos. The experiment has obtained a more precise value of the mixing angle and the first result on neutrino mass difference Δ mee2 from an energy and baseline dependent reactor neutrino disappearance using ∼500 days of data. Based on the ratio of inverse-beta-decay (IBD) prompt spectra measured in two identical far and near detectors, we obtain sin2 ⁡ (2θ13) = 0.082 ± 0.009 (stat .) ± 0.006 (syst .) and | Δ mee2 | = [2.62-0.23+0.21 (stat .)-0.13+0.12 (syst .) ] ×10-3 eV2. An excess of reactor antineutrinos near 5 MeV is observed in the measured prompt spectrum with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. A successful measurement of θ13 is also made in an IBD event sample with a delayed signal of neutron capture on hydrogen. A precise value of θ13 would provide important information on determination of the leptonic CP phase if combined with a result of an accelerator neutrino beam experiment.

  15. Performance of small reactors at universities for teaching, research, training and service (TRTS): thirty five years' experience with the Dalhousie University SLOWPOKE-2 reactor

    International Nuclear Information System (INIS)

    The Dalhousie University SLOWPOKE-2 Reactor (DUSR) facility, operated during 1976-2011, was the only research reactor in Atlantic Canada as well as the only one associated with a chemistry department in a Canadian university. The most outstanding features of the facility included: a rapid (100 ms) cyclic pneumatic sample transfer system, a permanently installed Cd-site, and a Compton-suppression gamma-ray spectrometer. The usage encompassed fundamental as well as applied studies in various fields using neutron activation analysis (NAA). The facility was used for training undergraduate/graduate students, postdoctoral fellows, technicians, and visiting scientists, and for cooperative projects with other universities, research organizations and industries. (author)

  16. Neutronic evaluation of a non-fertile fuel for the disposition of weapons-grade plutonium in a boiling water reactor

    International Nuclear Information System (INIS)

    A new non-fertile, weapons-grade plutonium oxide fuel concept is developed and evaluated for deep burn applications in a boiling water reactor environment using the General Electric 8x8 Advanced Boiling Water Reactor (ABWR) fuel assembly dimensions and pitch. Detailed infinite lattice fuel burnup results and neutronic performance characteristics are given and although preliminary in nature, clearly demonstrate the fuel's potential as an effective means to expedite the disposition of plutonium in existing light water reactors. The new non-fertile fuel concept is an all oxide composition containing plutonia, zirconia, calcia, and erbia having the following design weight percentages: 8.3; 80.4; 9.7; and 1.6. This fuel composition in an infinite fuel lattice operating at linear heat generation rates of 6.0 or 12.0 kW/ft per rod can remain critical for up to 1,200 and 600 Effective Full Power Days (EFPD), respectively, and achieve a burnup of 7.45 x 1020 f/cc. These burnups correspond to a 71--73% total plutonium isotope destruction and a 91--94% destruction of the 239Pu isotope for the 0--40% moderator steam void condition. Total plutonium destruction greater than 73% is possible with a fuel management scheme that allows subcritical fuel assemblies to be driven by adjacent high reactivity assemblies. The fuel exhibits very favorable neutron characteristics from beginning-of-life (BOL) to end-of-life (EOL). Prompt fuel Doppler coefficient of reactivity are negative, with values ranging between -0.4 to -2.0 pcm/K over the temperature range of 900 to 2,200 K. The ABWR fuel lattice remains in an undermoderated condition for both hot operational and cold startup conditions over the entire fuel burnup lifetime

  17. SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

    International Nuclear Information System (INIS)

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

  18. SWR 1000: A Next-Generation Boiling Water Reactor Ready for Deployment

    International Nuclear Information System (INIS)

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 design is particularly suitable for countries whose power systems do not include any large power plants. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (IandC) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10x10 rod array to a 12x12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free start-up, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving

  19. Status of the RENO Reactor Neutrino Experiment

    International Nuclear Information System (INIS)

    The Reactor Experiment for Neutrino Oscillation (RENO) is a reactor based neutrino oscillation experiment to measure the neutrino mixing angle θ13 using anti-neutrinos emitted from the Yonggwang nuclear power plant in Korea. Its thermal power output is 16.4 Gwth. The experimental setup consists of two identical 16-ton gadolinium-loaded liquid scintillator detectors. The near and far detectors are placed roughly 290 m and 1.4 km from the center of the reactor array, respectively. The near detector is constructed at underground of a 70 m high hill and the far detector at underground of a 260 m high mountain. The construction of experimental halls and access tunnels for both near and far detector sites was completed in early 2009. The experiment is planned to start data-taking from early 2011. An expected number of observed anti-neutrinos is roughly 1300 per day and 100 per day in the far and near detector, respectively. We expect that an estimated systematic uncertainty is less than 0.5%. With three years of data taken, RENO is sensitive to measure sin2(2θ13)>0.02. Sensitivity is ten times better than the current limit obtained by Chooz. In this review, current status of RENO is presented.

  20. Analysis of the peach bottom 2 BWR turbine trip experiment by RELAP 5/3.2 code

    OpenAIRE

    Bousbia-Salah Anis; D'auria Francesko

    2002-01-01

    This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to...