WorldWideScience

Sample records for boiling water reactor

  1. 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

  2. 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...

  3. 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.)

  4. 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)

  5. 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)

  6. 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.

  7. 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

  8. 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.)

  9. 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

  10. 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...

  11. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  12. 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)...

  13. 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...

  14. 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)....

  15. 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)

  16. 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)

  17. 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)

  18. 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.

  19. 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.

  20. 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)

  1. 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)

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. 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

  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. 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.

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

    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.

  17. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.''...

  18. 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

  19. 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

  20. 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

  1. 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 Regarding an Exemption Request AGENCY: Nuclear Regulatory Commission. ACTION: Environmental assessment and... Waste Management and Environmental Protection, Office of Federal and State Materials and...

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. Construction of the advanced boiling water reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan). Nuclear Power Plant Construction Dept.

    1996-07-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7.

  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. 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)

  12. 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

  13. 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

  14. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... revised 10 CFR 73.55 through the issuance of a final rule on March 27, 2009 (74 FR 13926). Section 73.55... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI...

  15. 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 ...

  16. 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

  17. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gavilian-Moreno, Carlos [Iberdrola Generacion, S.A., Cofrentes Nuclear Power Plant, Project Engineering Department, Paraje le Plano S/N, Valencia (Spain); Espinosa-Paredes, Gilberto [Area de ingeniera en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Mexico city (Mexico)

    2016-04-15

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  18. 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 ...

  19. 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.

  20. 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)

  1. 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

  2. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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

  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. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Andrey [Paul Scherrer Institut, Villigen (Switzerland); Degueldre, Claude, E-mail: claude.degueldre@psi.ch [Paul Scherrer Institut, Villigen (Switzerland); Kaufmann, Wilfried [Kernkraftwerk Leibstadt, Leibstadt (Switzerland)

    2013-01-15

    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 applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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)

  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. 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

  2. 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

  3. 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

  4. Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.; Gaitonde, S.M.

    1982-09-01

    The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

  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. 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

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

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available 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 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.

  8. 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.

  9. 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)

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Garg, A. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India); Sastry, P.S. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India); Pandey, M. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India)]. E-mail: manmohan@iitg.ac.in; Dixit, U.S. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039, Assam (India); Gupta, S.K. [Atomic Energy Regulatory Board, Mumbai 400085 (India)

    2007-02-15

    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.

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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.

  7. 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.)

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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...

  15. 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.

  16. 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

  17. Strain-induced corrosion cracking behaviour of low-alloy steels under boiling water reactor conditions

    Science.gov (United States)

    Seifert, H. P.; Ritter, S.

    2008-09-01

    The strain-induced corrosion cracking (SICC) behaviour of different low-alloy reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP ⩾+50 mV SHE, ⩾0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness <350 HV5) and increased (once initiated) with increasing loading rates and with increasing temperature with a possible maximum/plateau at 250 °C. A minimum KI value of 25 MPa m 1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate d KI/d t, but were not dependent on the actual KI values up to 60 MPa m 1/2. A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200-250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.

  18. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. A bifurcation analysis of boiling water reactor on large domain of parametric spaces

    Science.gov (United States)

    Pandey, Vikas; Singh, Suneet

    2016-09-01

    The boiling water reactors (BWRs) are inherently nonlinear physical system, as any other physical system. The reactivity feedback, which is caused by both moderator density and temperature, allows several effects reflecting the nonlinear behavior of the system. Stability analyses of BWR is done with a simplified, reduced order model, which couples point reactor kinetics with thermal hydraulics of the reactor core. The linear stability analysis of the BWR for steady states shows that at a critical value of bifurcation parameter (i.e. feedback gain), Hopf bifurcation occurs. These stable and unstable domains of parametric spaces cannot be predicted by linear stability analysis because the stability of system does not include only stability of the steady states. The stability of other dynamics of the system such as limit cycles must be included in study of stability. The nonlinear stability analysis (i.e. bifurcation analysis) becomes an indispensable component of stability analysis in this scenario. Hopf bifurcation, which occur with one free parameter, is studied here and it formulates birth of limit cycles. The excitation of these limit cycles makes the system bistable in the case of subcritical bifurcation whereas stable limit cycles continues in an unstable region for supercritical bifurcation. The distinction between subcritical and supercritical Hopf is done by two parameter analysis (i.e. codimension-2 bifurcation). In this scenario, Generalized Hopf bifurcation (GH) takes place, which separates sub and supercritical Hopf bifurcation. The various types of bifurcation such as limit point bifurcation of limit cycle (LPC), period doubling bifurcation of limit cycles (PD) and Neimark-Sacker bifurcation of limit cycles (NS) have been identified with the Floquet multipliers. The LPC manifests itself as the region of bistability whereas chaotic region exist because of cascading of PD. This region of bistability and chaotic solutions are drawn on the various

  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. 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.

  14. 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

  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. 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

  17. 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

  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. 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.

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16

    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

  12. 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

  13. 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.

  14. 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)

  15. Comparing Simulation Results with Traditional PRA Model on a Boiling Water Reactor Station Blackout Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

    A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: “W” process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four

  16. 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

  17. 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

  18. The chemistry of feedwater for boiling-water and pressurized-water reactors

    International Nuclear Information System (INIS)

    In a nuclear power plant the purity of the feedwater depends largely on whether a condensate polishing plant is provided, whether the loop is conditioned and on the presence of corrosion products originating in the materials from which the loop is made. The feedwater specification depends on the type of steam generator used. The article defines the characteristic parameters of a condensate polishing plant (CPP), such as the 'degree of polishing' and 'practical exchange capacity of the resins' and indicates how they can be determined. In pressurized-water reactors (PWR) the feedwater is normally conditioned with hydrazine. Measurements are quoted to demonstrate that, in contrast to conventional plants, the point of injection is immaterial as regards the copper content of the feedwater. Moreover, the iron content of the feedwater of a PWR can be reduced by using cyclic amines. The feedwater chemistry of a BWR is discussed by referring to oxygen, iron and copper measurements. The authors show that in loops in which the feed-heater condensate is pumped forwards and where a feedwater tank is provided, the stipulated purity of the feedwater can be attained by suitable measures (such as mechanical filtration, prevention of erosion-corrosion, and so on). (Auth.)

  19. 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.

  20. 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

  1. 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.

  2. 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.)

  3. 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.

  4. 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

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. Hybrid analysis of the simplified boiling water reactor using RAMONA-4B and CASMO-3 computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Vivas, G.F.C.; Hassan, Y.A. [Texas A and M Univ., College Station, TX (United States). Dept. of Nuclear Engineering

    1999-09-01

    An analysis of the simplified boiling water reactor (SBWR) is carried out using the reactor analysis computer program ROMONA-4B in an operational transient scenario, a turbine trip with failure of all the bypass valves. The SBWR model represents the vessel`s internal components, such as flow areas, diameters, and volumes. The one-quarter-core neutron parameters are calculated with the CASMO-3 transport theory lattice physics computer program. The three-dimensional representation of the reactor core uses some standard fuel design parameters, such as a wide central water rod, 8 x 8 lattice, gadolinium rods, etc. The thermal-hydraulic equations are solved with the RAMONA-4B computer program in a closed loop inside the reactor vessel and in 184 parallel channels (including bypass) in the core. Finally, the two-phase coolant and neutronic parameters are calculated in steady state and during the turbine trip transient. The results obtained compare favorably with the standard safety analysis report data.

  11. 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

  12. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  13. 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.

  14. 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.

  15. A compilation of boiling water reactor operational experience for the United Kingdom's Office for Nuclear Regulations Advanced Boiling Water Reactor generic design assessment.

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, Timothy A.; Liao, Huafei

    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 Kingdoms Health and Safety Executive Office for Nuclear Regulations (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.

  16. 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)

  17. Source term attenuation by water in the Mark I boiling water reactor drywell

    International Nuclear Information System (INIS)

    Mechanistic models of aerosol decontamination by an overlying water pool during core debris/concrete interactions and spray removal of aerosols from a Mark I drywell atmosphere are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are identified. Ranges for values of parameters that characterize these uncertain features of the models are established. Probability density functions for values within these ranges are assigned according to a set of rules. A Monte Carlo uncertainty analysis of the decontamination factor produced by water pools 30 and 50 cm deep and subcooled 0--70 K is performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0.001 cm3 H2O/cm2-s and decontamination factors of 1.1, 2, 3.3, 10, 100, and 1000

  18. 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.

  19. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Science.gov (United States)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  20. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  1. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    International Nuclear Information System (INIS)

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  2. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    Science.gov (United States)

    Kickhofel, J. L.; Zboray, R.; Damsohn, M.; Kaestner, A.; Lehmann, E. H.; Prasser, H.-M.

    2011-09-01

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  3. Azcaxalli: A system based on Ant Colony Optimization algorithms, applied to fuel reloads design in a Boiling Water Reactor

    International Nuclear Information System (INIS)

    This paper presents some results of the implementation of several optimization algorithms based on ant colonies, applied to the fuel reload design in a Boiling Water Reactor. The system called Azcaxalli is constructed with the following algorithms: Ant Colony System, Ant System, Best-Worst Ant System and MAX-MIN Ant System. Azcaxalli starts with a random fuel reload. Ants move into reactor core channels according to the State Transition Rule in order to select two fuel assemblies into a 1/8 part of the reactor core and change positions between them. This rule takes into account pheromone trails and acquired knowledge. Acquired knowledge is obtained from load cycle values of fuel assemblies. Azcaxalli claim is to work in order to maximize the cycle length taking into account several safety parameters. Azcaxalli's objective function involves thermal limits at the end of the cycle, cold shutdown margin at the beginning of the cycle and the neutron effective multiplication factor for a given cycle exposure. Those parameters are calculated by CM-PRESTO code. Through the Haling Principle is possible to calculate the end of the cycle. This system was applied to an equilibrium cycle of 18 months of Laguna Verde Nuclear Power Plant in Mexico. The results show that the system obtains fuel reloads with higher cycle lengths than the original fuel reload. Azcaxalli results are compared with genetic algorithms, tabu search and neural networks results.

  4. Azcaxalli: A system based on Ant Colony Optimization algorithms, applied to fuel reloads design in a Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel-Estrada, Jaime, E-mail: jaime.esquivel@fi.uaemex.m [Facultad de Ingenieria, Universidad Autonoma del Estado de Mexico, Cerro de Coatepec S/N, Toluca de Lerdo, Estado de Mexico 50000 (Mexico); Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Ortiz-Servin, Juan Jose, E-mail: juanjose.ortiz@inin.gob.m [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Castillo, Jose Alejandro; Perusquia, Raul [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico)

    2011-01-15

    This paper presents some results of the implementation of several optimization algorithms based on ant colonies, applied to the fuel reload design in a Boiling Water Reactor. The system called Azcaxalli is constructed with the following algorithms: Ant Colony System, Ant System, Best-Worst Ant System and MAX-MIN Ant System. Azcaxalli starts with a random fuel reload. Ants move into reactor core channels according to the State Transition Rule in order to select two fuel assemblies into a 1/8 part of the reactor core and change positions between them. This rule takes into account pheromone trails and acquired knowledge. Acquired knowledge is obtained from load cycle values of fuel assemblies. Azcaxalli claim is to work in order to maximize the cycle length taking into account several safety parameters. Azcaxalli's objective function involves thermal limits at the end of the cycle, cold shutdown margin at the beginning of the cycle and the neutron effective multiplication factor for a given cycle exposure. Those parameters are calculated by CM-PRESTO code. Through the Haling Principle is possible to calculate the end of the cycle. This system was applied to an equilibrium cycle of 18 months of Laguna Verde Nuclear Power Plant in Mexico. The results show that the system obtains fuel reloads with higher cycle lengths than the original fuel reload. Azcaxalli results are compared with genetic algorithms, tabu search and neural networks results.

  5. 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)

  6. 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

  7. 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

  8. Analysis of cracked core spray injection line piping from the Quad Cities Units 1 and 2 boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.

    1983-12-01

    Elbow assemblies and adjacent piping from the loops A and B core spray injection lines of Quad Cities Units 1 and 2 Boiling Water Reactors have been examined in order to determine the nature and causes of coolant leakages and flaw indications detected during hydrostatic tests and subsequent ultrasonic inspections. The elbow assemblies were found to contain multiple intergranular cracks in the weld heat-affected zones. The cracking was predominantly axial in orientation in the forged elbow and wedge components, whereas mixed axial and circumferential cracking was seen in the wrought piping pieces. In at least two instances, axial cracks completely penetrated the circumferential weld joining adjacent components. Based upon the observations made in the present study, the failures were attributed to intergranular stress corrosion cracking caused by the weld-induced sensitized microstructure and residual stresses present; dissolved oxygen in the reactor coolant apparently served as the corrosive species. The predominantly axial orientation of the cracks present in the forged components is believed to be related to the banded microstructure present in these components. The metallographic studies reported are supplemented by x-radiography, chemical analysis and mechanical test results, determinations of the degree of sensitization present, and measurements of weld metal delta ferrite content.

  9. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

  10. SWR 1000: an advanced boiling water reactor with passive safety features

    International Nuclear Information System (INIS)

    The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

  11. Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

    Energy Technology Data Exchange (ETDEWEB)

    Yan Jin, E-mail: jinyan10@gmail.co [GE-Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, M/L-30, NC 28402 (United States); Bolger, Francis [GE-Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, M/L-30, NC 28402 (United States)

    2010-07-15

    In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.

  12. Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

    International Nuclear Information System (INIS)

    In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.

  13. 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.

  14. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    Energy Technology Data Exchange (ETDEWEB)

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  15. 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

  16. 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

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

    CERN Document Server

    Marcel, CP

    2007-01-01

    In the design of novel nuclear reactors active systems are replaced by passive ones in order to reduce the risk of failure. For that reason natural circulation is being considered as the primary cooling mechanism in next generation nuclear reactor designs

  18. 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

  19. 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)

  20. Contribution to the multidimensional modelling of convective high pressure boiling flows for pressurised water reactors

    International Nuclear Information System (INIS)

    This study is a contribution to the modelling of multidimensional high pressure boiling flows relative to PWR. Numerical simulation of such two-phase flows is considered to be an interesting way for the DNB understanding. The first part of this study exposes a two-dimensional steady state two-phase flows model able to predict velocity and temperature profiles in tube. The mixture balanced equations are used with the eddy diffusivity concept to close the turbulent transport terms. The second part is devoted to the development of the model in the general two dimensional case. Contrary to the steady state model, this model is independent of experimental data and implies the use of an original local homogeneous relaxation model (HRM). The results obtained from the comparison with the data bank DEBORA reveals that in a mixture approach two sub models are sufficient to obtain a physical good description of turbulent boiling flows. Some limitations appear at conditions close to DNB conditions. The turbulent closures and the relaxation time in the HRM model have been clearly identified as the most important and sensitive parameters in the model. (author)

  1. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... March 27, 2009 (74 FR 13926). The revised regulation stated that it was applicable to all Part 50... April 30, 1987, and reactor defueling was completed on June 11, 1987. Pursuant to Title 10 of the...

  2. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    International Nuclear Information System (INIS)

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff's review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff's review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design

  3. 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)

  4. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Mexico, D.F. (Mexico); Francois, Juan Luis [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico)]. E-mail: jlfl@fi-b.unam.mx; Martin-del-Campo, Cecilia [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana, Avenida San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico)

    2005-04-15

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS 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. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the {sup 233}U build-up. 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 {sup 235}U 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 fuel assembly.

  5. 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

  6. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... March 27, 2009 (74 FR 13926). The revised regulation stated that it was applicable to all Part 50... published in the Federal Register on May 8, 2012 (77 FR 27097). Based upon the environmental assessment, the... April 30, 1987, and reactor defueling was completed on June 11, 1987. The decommissioning plan...

  7. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

    ...) inside the reactor building, and (3) well away from the power block. Locations inside the primary..., 2009 (74 FR 62829). On June 12, 2009 (74 FR 28112), the NRC amended its regulations to require.... Rather, the AIA rule's goal is to enhance the facility's inherent robustness at the design stage. The...

  8. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. 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.

  9. 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)

  10. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. 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

  11. 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

  12. A one dimensional model for the Laguna Verde boiling water reactor

    International Nuclear Information System (INIS)

    This paper presents a BWR model that includes the reactor vessel, two independent recirculation loops, the steam line, the reactor protection system and the pressure and feedwater control systems. The model has one-dimensional core thermalhydraulics and can handle reverse flow in the recirculation loops. Its point-kinetics core approximation has provided excellent results for all the steady state and operational transient conditions analyzed. Other features include a decay heat module, a heat transfer module with one node for the fuel and another for the cladding, some emergency systems, and all the valves associated with the steam lines. The modelling takes into account all the necessary details, while maintaining simplicity in the mathematical aspects: resulting in a flexible code with friendly interactive capabilities. Some specific transients run for the Laguna Verde Nuclear Power Plant are presented

  13. Pellet-clad interaction observations in boiling water reactor fuel elements

    International Nuclear Information System (INIS)

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  14. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  15. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    International Nuclear Information System (INIS)

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  16. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M and calculated (C results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE/C ratios of 1.10 for both neutron (E >1.0 MeV flux and iron atom displacement rate.

  17. Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

    International Nuclear Information System (INIS)

    This report documents benchmark evaluations to test thorium lattice predictive methods and neutron cross sections against available data and summarizes specific evaluations of the once-through thorium cycle when applied to the Browns Ferry-3 BWR. It was concluded that appreciable uncertainties in thorium cycle nuclear data cloud the ability to reliably predict the fuel cycle performance and that power reactor irradiations of ThO2 rods in BWRs are desirable to resolve uncertainties. Benchmark evaluations indicated that the ENDF/B-IV data used in the evaluations should cause an underprediction of U-233/ThO2 fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative

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

    International Nuclear Information System (INIS)

    Core management system by data communication has been designed and proposed for more efficient operation of BWR plants by faster transmission and centralized management of information system comprises three kinds of computers: process computer for monitoring purpose at reactor site, center computer for administration purpose at head office and large scientific computer for planning and evaluation purpose. The process and the large computers are connected to the center computer by data transmission line. To demonstrate the feasibility of such a system, 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 operating history of a commercial BWR shows a great deal of merits. Quick response and considerably large amount of reduction of manpower can be expected by data communication and minimized intervention of human labor. Visual display is also found to be very useful to understand the core characteristics

  19. 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

  20. Reduced scale simulations of boiling water reactor pool swell: some limitations to the scaling laws

    International Nuclear Information System (INIS)

    Several potential sources of misscaling in reduced scale experimental tests have been systematically investigated. Increases in the enthalpy in-flux during pool swell increase resultant uploads; slight boundary flexibility due to small air bubbles attached to the pool walls or true fluid structure interaction can increase peak pool boundary loads; the presence of water vapor in the wetwell airspace can either increase or decrease pool swell uploads, depending on the vapor fraction initially present. 14 refs

  1. On recriticality during reflooding of a degraded boiling water reactor core

    International Nuclear Information System (INIS)

    In-vessel core melt progression in Nordic BWRs has been studied as a part of the RAK-2 project within the Nordic Nuclear Safety Programme 1994-1997. A part of this study was the evaluation of possibility and consequences of recriticality in a re-flooded, degraded BWR core. The objective of the study was to examine, if a BWR core in a Nordic nuclear power plant can reach critical state in a severe accident, when the core is re-flooded with un-borated water from the emergency core cooling system and what is the possible power augmentation related to recriticality. The containment response to elevated power level and consequent enhanced steam production was evaluated. The first sub-task was to upgrade the existing neutronics/thermal hydraulic models to a level needed for a study of recriticality. Three different codes were applied for the task: RECRIT, SIMULATE-3K and APROS. Preliminary calculations were performed with the three codes. The results of present studies showed that reflodding of a partly control rod free core gives a recriticality power peak of a substantial amplitude, but with a short duration due to the Doppler feedback. The energy addition is small and contributes very little to heat-up of the fuel. However, with continued reflodding the fission power increases again and tend to stabilize on a level that can be ten per cent or more of the nominal power, the level being higher with higher reflooding flow rate. A scoping study on TVO BWR containment response to a presumed recriticality accident with a long-term power level being 20% of the nominal power was performed. The results indicated that containment venting system would not be sufficient to prevent containment overpressurization and containment failure would occur about 3-4 h after start of core reflooding. In the case of station blackout with operating ADS the present boron system would be sufficient to terminate the criticality even prior to containment failure, but in case of feedwater LOCA and

  2. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    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 O{sub 2}; 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

  3. Analysis of mixed oxide fuel behavior under reduced moderation boiling water reactor conditions with FRAPCON-EP

    International Nuclear Information System (INIS)

    FRAPCON-EP models have been extended to better represent mixed oxide steady state fuel behavior under the Reduced moderation Boiling Water Reactor (RBWR) conditions. RBWR fuel is designed to operate with higher peak burnup, linear heat rate, and fast neutron fluence compared to typical LWRs. Therefore, assessment of fuel behavior is a critical task for its core performance. The fuel pellet radial power profile is calculated based on plutonium radial variation and edge peaking due to resonance absorption of neutrons. It is found that the edge power peak is much smaller than in typical LWRs due to the harder neutron spectrum. The oxygen potential directly affects fuel thermal conductivity and fission gas diffusivity. Plutonium migration towards the high temperature may potentially lead to power peaks at the central radial locations. The selected fuel thermal conductivity model for mixed oxides accounts for the oxygen-to-metal ratio variation, burnup effects due to fission product precipitates, radiation damage and porosity. In addition, Zircaloy-2 cladding corrosion/hydrogen pickup models in FRAPCON-3 have been updated to reflect accelerated corrosion/hydriding, due mainly to secondary particle precipitate dissolution. Based on experimental data, acceleration is assumed to occur above 10+26 n/m2 of fast neutron fluence (>1 MeV). Analysis of RBWR fuel was made together with neutron dose calculation using the reference power history. The neutron transport analysis shows that RBWR fuel fast fluence-to-volumetric heat generation ratio is approximately 80 % more than in typical LWRs. Initially, an analysis was performed with traditional Zircaloy-2 and reference mixed oxide fuel pellet with 95 % theoretical density. It was found that accelerated corrosion/hydriding may result at peak burnups as low as 30 MWd/kg. Furthermore, excessive fuel swelling may result in significant cladding strain and axial irradiation growth, which may lead to creep induced fracture as well as

  4. 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

  5. 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

  6. 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

  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. 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...

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1999-02-01

    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.

  11. 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

  12. 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

  13. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  14. 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

    ... at the Fukushima Dai-ichi nuclear power plant following the March 2011 earthquake and tsunami... communication between the ] drywell and the wetwell volume above the water in the suppression pool. This..., ``Written Communications.'' The Director, Office of Nuclear Reactor Regulation may, in writing, relax...

  15. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    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)

  16. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1; Dekontamination des RDB inkl. der Einbauten wie Dampftrockner und Wasserabscheider sowie der angeschlossenen Hilfssysteme im deutschen Siedewasserreaktor ISAR 1

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Michael; Sempere Belda, Luis; Basu, Ashim; Topf, Christian [AREVA GmbH, Erlangen (Germany). Abt. Chemistry Services; Erbacher, Thomas; Hiermer, Thomas; Schnurr, Bernhard; Appeldorn, Thomas van [E.ON Kernkraft GmbH, Kernkraftwerk ISAR, Essenbach (Germany). Abt. Maschinentechnik; Volkmann, Christian [ESG Engineering Services GmbH, Greifswald (Germany)

    2015-12-15

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17{sup th}, 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.

  17. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  18. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  19. Hydraulic performance of pump suction inlets for emergency core cooling systems in boiling water reactors. Containment sump reliability studies. Generic task A-43

    International Nuclear Information System (INIS)

    This document reports on the hydraulic performance of two representative Boiling Water Reactor (BWR) Residual Heat Removal (RHR) suction inlet configurations; namely, those of the Mark I, and Mark II and Mark III designs. Key parameters of interest were air-ingestion levels, vortex types, suction pipe swirl, and the RHR inlet pressure loss coefficient. Tests were conducted with nearly uniform and non-uniform approach flows to the inlets. Flows and submergences were in the range of from 2000 to 12,000 gpm per pipe and 2 to 5 ft, respectively, giving a Froude number range of 0.17 to 1.06. Zero air-withdrawal was measured for both configurations for Froude number equal to or less than 0.8 even under non-unifrom approach flows; likewise, no air-core vortices were observed for the same flow conditions

  20. 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

  1. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  2. Boiling water reactors with Uranium-Plutonium mixed oxide fuel. Report 1: Accuracy of the nuclide concentrations calculated by CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1999-07-01

    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. These CMS (Core Management System) programs have been extensively compared with both measurements and reference codes. Nevertheless some data are proprietary in particular the comparison of the calculated nuclide concentrations versus experiments (because of the cost of this kind of experimental study). This is why this report describes such a comparative investigation carried out with a General Electric 7x7 BWR bundle. Unfortunately, since some core history parameters were unknown, a lot of hypotheses have been adopted. This invokes sometimes a significant discrepancy in the results without being able to determine the origin of the differences between calculations and experiments. Yet one can assess that, except for four nuclides - Plutonium-238, Curium-243, Curium-244 and Cesium-135 - for which the approximate power history (history effect) can be invoked, the accuracy of the calculated nuclide concentrations is rather good if one takes the numerous approximations into account.

  3. On the shape of stress corrosion cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant piping at 288 °C

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang-Kwon; Kramer, Daniel [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Macdonald, Digby D., E-mail: macdonald@berkeley.edu [Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720 (United States)

    2014-11-15

    Evolution of the shape of surface cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant circuit piping at the reactor operating temperature of 288 °C is explored as a function of various environmental variables, such as electrochemical potential (ECP), solution conductivity, flow velocity, and multiplier for the oxygen reduction reaction (ORR) standard exchange current density (SECD), using the coupled environment fracture model (CEFM). For this work, the CEFM was upgraded by incorporating Shoji’s model for calculating the crack tip strain rate and more advanced expressions were used for estimating the stress intensity factor for semi-elliptical surface cracks. This revised CEFM accurately predicts the dependence of the crack growth rate on stress intensity factor and offers an alternative explanation for the development of semi-elliptical cracks than that provided by fracture mechanics alone. The evolution of surface crack semi-elliptical shape depends strongly upon various environmental variables identified above, and the CEFM predicts that the minor axis of the ellipse should be oriented perpendicular to the surface, in agreement with observation. The development of the observed semi-elliptical cracks with the minor axis perpendicular to the surface is therefore attributed to the dependence of the crack growth rate on the electrochemical crack length.

  4. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    International Nuclear Information System (INIS)

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs

  5. Technology, safety and costs of decommissioning a refernce boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  7. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  8. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    International Nuclear Information System (INIS)

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A ampersand 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met

  9. Comparison of depletion results for a boiling water reactor fuel element with CASMO and SCALE 6.1 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Morera, D.; Miro, R.; Barrachina, T.; Verdu, G., E-mail: cmesado@isirym.upv.es, E-mail: dmorera@isirym.upv.es, E-mail: rmiro@isirym.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@isirym.upv.es [Universitat Politecnica de Valencia (UPV), Valencia (Spain). Institute for the Industrial, Radiophysical and Environmental Safety; Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion, S.A.U, Madrid (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S. L., Madrid (Spain); Melara, Jose, E-mail: j.melara@iberdrola.es [Iberdrola Generacion Nuclear, Madrid (Spain)

    2013-07-01

    In this work, the results of depletion calculations with CASMO and SCALE 6.1 (TRITON) are compared. To achieve it, a region of a Boiling Water Reactor (BWR) fuel element is modeled, using both codes. To take into account different operating conditions, the simulations are repeated with different void fraction, ranging from null void fraction to a void fraction closed to one. Special care was used to keep in mind that the homogenization of the materials and the two group approach was the same in both codes. Additionally, KENO-VI and MCDANCOFF modules are used. The k-effective calculated by KENO-VI is used to ensure that the starting point was correct and MCDANCOFF module is used to calculate the Dancoff factors in order to improve the model accuracy. To validate the whole process, a comparison of k{sub eff}, and cross-sections collapsed and homogenized is shown. The results show a very good agreement, with an average error around the 1.75%. Furthermore, an automatic process for translating CASMO data to SCALE input decks was developed. The reason for the translation is the fact that SCALE's TRITON module is a new code very powerful and continuously being developed. Thus, great advantage can be taken from the current and future SCALE features. This is hoped to produce more realistic models, and hence, increase the accuracy of neutronic libraries. (author)

  10. Experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors; Experimentelle und numerische Untersuchung von Gas/Liquid-Phasengrenzflaechen als Referenzwert fuer die hydrostatische Fuellstandsmessung in Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, Stephan

    2013-12-17

    The experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors is considered as relevant for reactor safety research. The experiments allow a quantification of the transition processes in hydrostatic level measurement devices that were up to now only assessed by phenomenological descriptions. Experimental studies covered the topology and stability of water/vapor phase boundaries and the numerical description using CFD codes, including modeling of the surface topology and modeling of the heat and mass transport.

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

    International Nuclear Information System (INIS)

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

  12. 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.

  13. 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

  14. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    International Nuclear Information System (INIS)

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

  15. Categorization of core-damage sequences by containment event tree analysis for boiling water reactor with Mark-II containment

    International Nuclear Information System (INIS)

    In the present study, containment responses to core damage accidents were analyzed for a large spectrum of core damage sequences, which were defined by front-line system event trees, in a BWR with Mark-11 containment by using the Accident Progression Event Tree (APSET) method and their characteristics were examined in terms of mainly probabilistic aspects such as their respective conditional probabilities of containment failure modes and accident termination. This paper showed that various core damage sequences could be categorized into a small number of groups, each of which consisted of the sequences with similar containment response characteristics, as follows: Interfacing system LOCA; ATWS with high pressure injection available; Loss of long-term containment heat removal; Station blackout; Loss of coolant injection with the reactor not depressurized; Loss of coolant injection with the reactor depressurized; Loss of short-term containment heat removal; and Reactor pressure vessel rupture. The above categorization provides a perspective on the potential containment failure modes and the effectiveness of some accident mitigative measures, which could be useful for studying accident management strategies and as well for assisting the analysts in carrying out future CET analyses. (author)

  16. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Wagner, K.C. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  17. An assessment of BWR [boiling water reactor] Mark-II containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs

  18. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    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.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  19. A citation-based assessment of the performance of U.S. boiling water reactors following extended power up-rates

    Science.gov (United States)

    Heidrich, Brenden J.

    Nuclear power plants produce 20 percent of the electricity generated in the U.S. Nuclear generated electricity is increasingly valuable to a utility because it can be produced at a low marginal cost and it does not release any carbon dioxide. It can also be a hedge against uncertain fossil fuel prices. The construction of new nuclear power plants in the U.S. is cautiously moving forward, restrained by high capital costs. Since 1998, nuclear utilities have been increasing the power output of their reactors by implementing extended power up-rates. Power increases of up to 20 percent are allowed under this process. The equivalent of nine large power plants has been added via extended power up-rates. These up-rates require the replacement of large capital equipment and are often performed in concert with other plant life extension activities such as license renewals. This dissertation examines the effect of these extended power up-rates on the safety performance of U.S. boiling water reactors. Licensing event reports are submitted by the utilities to the Nuclear Regulatory Commission, the federal nuclear regulator, for a wide range of abnormal events. Two methods are used to examine the effect of extended power up-rates on the frequency of abnormal events at the reactors. The Crow/AMSAA model, a univariate technique is used to determine if the implementation of an extended power up-rate affects the rate of abnormal events. The method has a long history in the aerospace industry and in the military. At a 95-percent confidence level, the rate of events requiring the submission of a licensing event report decreases following the implementation of an extended power up-rate. It is hypothesized that the improvement in performance is tied to the equipment replacement and refurbishment that is performed as part of the up-rate process. The reactor performance is also analyzed using the proportional hazards model. This technique allows for the estimation of the effects of

  20. 21 CFR 872.6710 - Boiling water sterilizer.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling...

  1. New strategies of reloads design and models of control bars in boiling water reactors; Nuevas estrategias de diseno de recargas y de patrones de barras de control en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J. A.; Ortiz S, J. J.; Perusquia del Cueto, R., E-mail: alejandro.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    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)

  2. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors; Etude experimentale des instabilites hydrodynamiques survenant dans les reacteurs nucleaires a ebullition

    Energy Technology Data Exchange (ETDEWEB)

    Fabreca, S. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-10-01

    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) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible

  3. Qualification of the Darwin code for the studies of the fuel cycle relative to the boiling water reactors; Qualification du formulaire Darwin pour les etudes du cycle du combustible pour les reacteurs a eau bouillante

    Energy Technology Data Exchange (ETDEWEB)

    Allais, V

    1998-03-01

    This thesis was carried out in the framework of fuel cycles studies in partnership with COGEMA; the aim is to determine physics parameters characterising Boiling Reactor Assemblies. Those reactors Firstly distinguish themselves from Pressurised Water Reactor by the boiling of the moderator in the core and secondary by the strong neutronics heterogeneity due to complex design. The diphasic mixture formed is characterised by the void fraction parameter. The loss of information, and neutronic studies characteristics of Boiling Water Reactors led us to make preliminary studies having in view to quantify the void fraction impact on the isotopics evolution. Studies on neutronics influence of assemblies and control rods from the immediate environment allows to define the cluster size to describe. The radial description optimisation with APOLLO-2 is necessary to improve the calculation performance and to reduce the errors coming from the modelization. The following points were studied: pellet radial discretization, clustering of cells characterized by a similar behaviour, options in flux spatial calculation (interface current formalism), self-shielding optimisation (specific to each isotopes). The three dimensional modelization with CRONOS-2 and the simplified accounting of the thermohydraulics / neutronics coupling done by a procedure developed and written during this thesis, allow an evaluation of axial distribution of void fraction, power and burn-up during the irradiation. The comparison with experimental analytic results of complete assembly and pin samples dissolutions allows the qualification of this procedure and confirms the necessity to take into account the void fraction axial variation during the evolution. The application of an automatic coupling with the DARWIN cycle code will allow a precise burnup calculation to be utilized in an industrial procedure. (author)

  4. Water boiling kinetic in rapid decompression

    International Nuclear Information System (INIS)

    This study entering in the frame of a CEA, EDF and Framatome collaboration, has for objective to modelize two-phase flows in case of PWR Loca. The objective is to find, by taking in account the all imbalances, a formulation for the mass transfer at the interface water-vapor by the study of water boiling phenomenon in case of fast decompression such as a primary circuit break. In this accident, the estimation of boiling speeds in an essential parameter for determining the break discharge which conditions the safety systems design

  5. Analysis of the rotation accident of assemblies in boiling water reactors; Analisis del accidente de rotacion de ensambles en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Becerril-Gonzalez M, J. J. [Universidad Autonoma de Yucatan, Av. Industrias no contaminantes por Anillo Periferico Norte s/n, Apdo. Postal 150 Cordemex, Merida, Yucatan (Mexico); Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia de Cueto, R., E-mail: juanjosebecerril_1@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    For this work was analyzed the impact that would cause the load of a rotated fuel assembly in the behaviour of the core in the Cycle 14 of the Unit 1 of the nuclear power plant of Laguna Verde. To carry out this analysis the code Simulate-3 was used, with which was possible to analyze the behavior of the effective multiplication factor and the thermal limits (MAPRAT, MFLPD and MFLCPR). The rotation of fuel assemblies to 90, 180 and 270 grades was analyzed with regard to the design position, with 0, 1, 2 and 3 burnt cycles for these assemblies. The results show that the thermal limits remain inside the allowed values, therefore if this accident type happened the reactor could continue operating in a sure way. (Author)

  6. The safety of light water reactors

    International Nuclear Information System (INIS)

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  7. Optimization of operation schemes in boiling water reactors using neural networks; Optimizacion de esquemas de operacion en reactores de agua en ebullicion usando redes neuronales

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Pelta, D. A., E-mail: juanjose.ortiz@inin.gob.mx [Universidad de Granada, Escuela Superior de Ingenierias, Informatica y Telecomunicacion, C/Daniel Saucedo Aranda s/n, 18071 Granada (Spain)

    2012-10-15

    In previous works were presented the results of a recurrent neural network to find the best combination of several groups of fuel cells, fuel load and control bars patterns. These solution groups to each problem of Fuel Management were previously optimized by diverse optimization techniques. The neural network chooses the partial solutions so the combination of them, correspond to a good configuration of the reactor according to a function objective. The values of the involved variables in this objective function are obtained through the simulation of the combination of partial solutions by means of Simulate-3. In the present work, a multilayer neural network that learned how to predict some results of Simulate-3 was used so was possible to substitute it in the objective function for the neural network and to accelerate the response time of the whole system of this way. The preliminary results shown in this work are encouraging to continue carrying out efforts in this sense and to improve the response quality of the system. (Author)

  8. 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

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

    International Nuclear Information System (INIS)

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

  10. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    Science.gov (United States)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  11. 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 ...

  12. 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].

  13. Water Boiling inside Carbon Nanotubes: Towards Efficient Drug Release

    OpenAIRE

    Chaban, Vitaly V.; Prezhdo, Oleg V.

    2012-01-01

    We show using molecular dynamics simulation that spatial confinement of water inside carbon nanotubes (CNT) substantially increases its boiling temperature and that a small temperature growth above the boiling point dramatically raises the inside pressure. Capillary theory successfully predicts the boiling point elevation down to 2 nm, below which large deviations between the theory and atomistic simulation take place. Water behaves qualitatively different inside narrow CNTs, exhibiting trans...

  14. Technique for technological calculation of critical flow of boiling water

    International Nuclear Information System (INIS)

    Average values of friction factor and mach number for a critical flow of boiling water are determined on the basis of computerized processing of experimental data. Empirical formula, relating these values, which can be used for technological calculations of critical conditions of boiling water flow through transport pipelines, is derived

  15. Microbiological Effectiveness of Disinfecting Water by Boiling in Rural Guatemala

    OpenAIRE

    Rosa, Ghislaine; Miller, Laura; Clasen, Thomas

    2010-01-01

    Boiling is the most common means of treating water in the home and the benchmark against which alternative point-of-use water treatment options must be compared. In a 5-week study in rural Guatemala among 45 households who claimed they always or almost always boiled their drinking water, boiling was associated with a 86.2% reduction in geometric mean thermotolerant coliforms (TTC) (N = 206, P < 0.0001). Despite consistent levels of fecal contamination in source water, 71.2% of stored water sa...

  16. Leukemia in the proximity of a German boiling water nuclear reactor: Evidence of population exposure by chromosome studies and environmental radioactivity

    International Nuclear Information System (INIS)

    The detection of an exceptional elevation of leukemia in children appearing 5 years after the start-up of the nuclear power plant Kruemmel in 1983, accompanied by a significant increase of leukemia cases in adults gave rise for investigations of radiation exposures of the population living near to the plant. The rate of dicentric chromosomes in peripheral blood lymphocytes of 7 parents of leukemia children and 14 other inhabitants in the proximity of the plant was significantly elevated and showed ongoing exposures over the years of operation. This finding gives rise to the hypothesis that chronic leakages by the reactor had occurred. This assumption is supported by the identification of artificial radioactivity in air, rain water, soil, and vegetation registered by the regular environmental monitoring programme of the nuclear power plant. Calculations of the corresponding source terms show that the originating emissions must have been well above authorized annual limits. The bone marrow dose is supposed to be originated mainly by incorporating of bone-seeking β- and α-emitters. (author)

  17. Transient CHF enhancement of saturated pool boiling of water using a honeycomb porous media

    International Nuclear Information System (INIS)

    Several studies have been performed to make clear the transient boiling heat transfer during the exponential heat generation which is occurred in reactivity accident of a nuclear reactor. These researches have been focused on the mechanism of the phenomena mainly, not on the enhancement of the transient boiling heat transfer. In a previous study, we proposed a method of CHF enhancement under steady-state conditions using honeycomb porous plate. The CHF was shown experimentally to be enhanced to more than twice that of a plain surface using honeycomb porous plate. The enhancement is considered to result from the capillary supply of liquid onto the heated surface and the release of generated vapor through the channels. In the present paper, enhancement of the transient critical heat flux in pool boiling by the attachment of a honeycomb-structured porous plate on a heated wire is investigated experimentally using water under saturated boiling conditions. (author)

  18. 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.

  19. 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.

  20. Advances in the development and validation of CFD-BWR, a two-phase computational fluid dynamics model for the simulation of flow and heat transfer in boiling water reactors

    International Nuclear Information System (INIS)

    This paper presents recent advances in the validation of an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel bundles. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. We have described the model development strategy that has been adopted by the development team for the prediction of boiling flow regimes in a BWR fuel bundle. This strategy includes the use of local flow topology maps and flow topology specific phenomenological models. The paper reviews the key boiling phenomenological models and focuses on recent results of experiment analyses for the validation of two-phase BWR phenomena models including cladding-to-coolant heat transfer and Critical Heat Flux experiments and the BWR Full-size Assembly Boiling Test (BFBT). The two-phase flow models implemented in the CFD-BWR code can be grouped into three broad categories: models describing the vapor generation at the heated cladding surface, models describing the interactions between the vapor and the liquid coolant, and models describing the heat transfer between the fuel pin and the two-phase coolant. These models have been described and will be briefly reviewed. The boiling model used in the second generation of the CFD-BWR code includes a local flow topology map which allows the cell-by-cell selection of the local flow topology. Local flow topologies can range from a bubbly flow topology where the continuous phase is liquid, to a transition flow topology, to a droplet flow topology where the continuous phase is vapor, depending primarily on the local void fraction. The models describing the cladding-to-coolant heat transfer and the interplay between these models and the local flow topology are important in Critical Heat Flux (CHF) analyses, and will

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

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

  4. A Matrix Method of Analyzing the Thermodynamic System of Advance Boiling Water Reactor Nuclear Power Unit%先进型沸水堆核电机组热经济性矩阵分析方法

    Institute of Scientific and Technical Information of China (English)

    冉鹏; 李庚生; 廖丹; 朱伟平

    2010-01-01

    根据先进型沸水堆(advance boiling water reactor,ABWR)核电机组热力系统的结构特点,基于热力系统等效热降分析方法和矩阵方法,确定其主、辅系统的划分原则以及辅助汽水成分划分原则,对先进型沸水堆各种汽水成分进行归并处理,构建表达规则的先进型沸水堆核电机组汽水分布方程填写规则,推导出适合先进型沸水堆核电机组热力系统热经济性分析的通用矩阵方法,并给出该类型核电机组辅助汽水成分对热经济性影响的表达方式.该矩阵全面反映了先进犁沸水堆核电机组热力系统主系统和各种辅助系统对机组热经济性的影响状况,每个子矩阵物理意义明确、规律性强,可使先进型沸水堆核电机组热力系统的整体计算和局部分析变得清晰、简单,适合于计算机程序化,并通过实例对该方法进行了验证.

  5. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 1. Investigation and evaluation of stress corrosion cracking in piping of boiling water reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1984-08-01

    IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry.

  6. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 1. Investigation and evaluation of stress corrosion cracking in piping of boiling water reactor plants

    International Nuclear Information System (INIS)

    IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry

  7. Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T.; Dinh, T.N.; Bui, V.A.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Energy Systems Technology

    1995-07-01

    Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs.

  8. 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

  9. On the dynamics of bubbles in boiling water

    International Nuclear Information System (INIS)

    Research highlights: → We devote this work to investigate the bubbles dynamics in boiling water. → A simple experiment of laser scattering was designed to obtain dynamical features. → Correlations and non-exponential distributions were found. → A simple model was able to describe several aspects of the system. - Abstract: We investigate the dynamics of many interacting bubbles in boiling water by using a laser scattering experiment. Specifically, we analyze the temporal variations of a laser intensity signal which passed through a sample of boiling water. Our empirical results indicate that the return interval distribution of the laser signal does not follow an exponential distribution; contrariwise, a heavy-tailed distribution has been found. Additionally, we compare the experimental results with those obtained from a minimalist phenomenological model, finding a good agreement.

  10. Safety analysis for boiling water reactors

    International Nuclear Information System (INIS)

    This report is the translation of GRS-95 'Sicherheitsanalyse fuer Siedewasserreaktoren - Zusammenfassende Darstellung'. Recent analysis results -concerning the chapters on accident management, fire and earthquake - that were not included in the German text have been added to this translation. In cases of doubt, GRS-102 (main volume) is the factually correct version. (orig.)

  11. Boiling water reactor uranium utilization improvement potential

    International Nuclear Information System (INIS)

    This report documents the results of design and operational simulation studies to assess the potential for reduction of BWR uranium requirements. The impact of the improvements on separative work requirements and other fuel cycle requirements also were evaluated. The emphasis was on analysis of the improvement potential for once-through cycles, although plutonium recycle also was evaluated. The improvement potential was analyzed for several design alternatives including axial and radial natural uranium blankets, low-leakage refueling patterns, initial core enrichment distribution optimization, reinsert of initial core discharge fuel, preplanned end-of-cycle power coastdown and feedwater temperature reduction, increased discharge burnup, high enrichment discharge fuel rod reassembly and reinsert, lattice and fuel bundle design optimization, coolant density spectral shift with flow control, reduced burnable absorber residual, boric acid for cold shutdown, six-month subcycle refueling, and applications of a once-through thorium cycle design and plutonium recycle

  12. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.)

  13. Electrochemical study of aluminum corrosion in boiling high purity water

    Science.gov (United States)

    Draley, J. E.; Legault, R. A.

    1969-01-01

    Electrochemical study of aluminum corrosion in boiling high-purity water includes an equation relating current and electrochemical potential derived on the basis of a physical model of the corrosion process. The work involved an examination of the cathodic polarization behavior of 1100 aluminum during aqueous oxidation.

  14. 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.

  15. 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...

  16. 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...

  17. 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)

  18. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  19. 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

  20. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  1. Flow boiling of water on nanocoated surfaces in a microchannel

    CERN Document Server

    Phan, Hai Trieu; Marty, Philippe; Colasson, Stéphane; Gavillet, Jérôme

    2010-01-01

    Experiments were performed to study the effects of surface wettability on flow boiling of water at atmospheric pressure. The test channel is a single rectangular channel 0.5 mm high, 5 mm wide and 180 mm long. The mass flux was set at 100 kg/m2 s and the base heat flux varied from 30 to 80 kW/m2. Water enters the test channel under subcooled conditions. The samples are silicone oxide (SiOx), titanium (Ti), diamond-like carbon (DLC) and carbon-doped silicon oxide (SiOC) surfaces with static contact angles of 26{\\deg}, 49{\\deg}, 63{\\deg} and 103{\\deg}, respectively. The results show significant impacts of surface wettability on heat transfer coefficient.

  2. Oxidation Effect on Pool Boiling Heat Transfer in Atmospheric Saturated Water

    Energy Technology Data Exchange (ETDEWEB)

    Son, Hong Hyun; Jeong, Uiju; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of)

    2014-10-15

    During the hypothesized severe accidents, however, the modified nature of the oxidized outer surface of RPV may act as a significant heat transfer variable to achieve In-Vessel Retention through External Reactor Vessel Cooling (IVR-ERVC) strategy, which is the one of important mitigation strategies of severe accident to delay occurrence of critical heat flux (CHF). As well understood, the CHF is mainly affected by the two distinctive conditions classified to thermal hydraulic behavior of fluid system and surface characteristics. In this regard, a CHF test considering oxidation effect on the pool boiling heat transfer of the RPV outer surface has been proposed to evaluate realistic thermal margin of IVR-ERVC strategy. In this study, pool boiling heat transfer experiment was conducted under the condition of atmospheric saturated water. Oxidized surface characteristics were quantitatively evaluated with measurement of contact angle and roughness. In this study, oxide layer formation on the heated surface was investigated and experimentally simulated to find out its effect on the pool boiling CHF. Several SS316L substrates were oxidized in the corrosive environment under the condition of high temperature with different oxidation periods. Local pitting corrosion was observed on the heating surface in 5 days of short-term oxidation but a fully oxidized surface with somewhat uniform thickness, 1. Pool boiling heat transfer tests with the bare and oxidized heaters were conducted and major findings are summarized as follows: 1. Wettability in terms of the receding angle of the oxidized surface is enhanced regardless of the oxidation period. 2. Average roughness between the oxidized surfaces is almost the same in the range of nano-scale. 3. Effect of wettability and surface roughness on the CHF was negligible in the locally oxidized surface, which may be attributed to the presence of the disconnected porous channel. Unlike the local oxidation, fully oxidized surface shows

  3. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  4. 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

  5. 75 FR 26967 - Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

    Science.gov (United States)

    2010-05-13

    ... HUMAN SERVICES Food and Drug Administration Guidance for Industry: Use of Water by Food Manufacturers in... entitled ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water... ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory.''...

  6. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  7. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  8. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  9. Calculations of the effect of boiling water on bitumen production

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.; Kantzas, A. [Calgary Univ., AB (Canada). Dept. of Chemical and Petroleum Engineering]|[Calgary Univ., AB (Canada). Tomographic Imaging and Porous Media Laboratory; McGee, B. [E-T Energy Limited, Calgary, AB (Canada)

    2006-07-01

    Alberta's vast resources of heavy oil and bitumen are playing an increasing role as a main resource for crude oil. Thermal recovery methods for heavy oil and bitumen include steam injection and steam flooding in which thermal energy is given to the oil to reduce its viscosity and allow it to flow towards a production spot. A viable alternative to steam injection is the electromagnetic heating method for heavy oil and bitumen reservoirs. Electromagnetic heating transfers heat to heavy oil reservoirs based on electromagnetic energy and can be used in situations where steam injection may not work well. The process can also be used to preheat the reservoir before steam injection. This study examined the possible displacement mechanisms of such processes with particular focus on the physics of boiling water in porous media as a potential displacement agent for heavy oil and bitumen. It is very possible that water could vaporize while being electrically heated and the vaporized water could push more heavy oil or bitumen out of reservoir. As such, higher oil recovery could be expected due to water vaporization. The role of water vaporization during electrical heating process was examined and a methodology to estimate the magnitude of incremental oil recovery was developed based on simple conceptual models with numerical simulators and illustrative experiments. The primary contributors of this process appear to be a combination of drainage, imbibition, viscosity reduction and gas expansion. The study showed that the expansion of water into steam could very efficiently flush oil out of pore spaces. It was concluded that water vaporization inside the reservoir can be an additional driving force for heavy oil or bitumen production, and that this alternative to steam injection can offer energy savings for the recovery process. 10 refs., 4 tabs., 6 figs., 1 appendix.

  10. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

  11. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  12. 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).

  13. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  14. Generation of shockwave and vortex structures at the outflow of a boiling water jet

    Science.gov (United States)

    Alekseev, M. V.; Lezhnin, S. I.; Pribaturin, N. A.; Sorokin, A. L.

    2014-12-01

    Results of numerical simulation for shock waves and generation of vortex structures during unsteady outflow of boiling liquid jet are presented. The features of evolution of shock waves and vortex structures formation during unsteady outflow of boiling water are compared with corresponding structures during unsteady gas outflow.

  15. Risks of nuclear energy technology safety concepts of light water reactors

    CERN Document Server

    Kessler, Günter; Schlüter, Franz-Hermann

    2014-01-01

    The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on?reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: ? A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Ch

  16. Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, K. [Kobe Univ. of Mercantile Marine (Japan); Shiotsu, M.; Sakurai, A. [Kyoto Univ. (Japan)

    1995-09-01

    Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q{sub max}, on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q{sub o}e{sup t/T}, with periods, {tau}, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q{sub max}. Two main mechanisms of q{sub max} exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q{sub max} for long period range belonging to the former mechanism becomes longer and the q{sub max}mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q{sub max} for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling.

  17. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  18. Modeling and measurement of boiling point elevation during water vaporization from aqueous urea for SCR applications

    Energy Technology Data Exchange (ETDEWEB)

    Dan, Ho Jin; Lee, Joon Sik [Seoul National University, Seoul (Korea, Republic of)

    2016-03-15

    Understanding of water vaporization is the first step to anticipate the conversion process of urea into ammonia in the exhaust stream. As aqueous urea is a mixture and the urea in the mixture acts as a non-volatile solute, its colligative properties should be considered during water vaporization. The elevation of boiling point for urea water solution is measured with respect to urea mole fraction. With the boiling-point elevation relation, a model for water vaporization is proposed underlining the correction of the heat of vaporization of water in the urea water mixture due to the enthalpy of urea dissolution in water. The model is verified by the experiments of water vaporization as well. Finally, the water vaporization model is applied to the water vaporization of aqueous urea droplets. It is shown that urea decomposition can begin before water evaporation finishes due to the boiling-point elevation.

  19. Environmentally assisted cracking in Light Water Reactors

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  20. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Science.gov (United States)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  1. Experimental study on a new solar boiling water system with holistic track solar funnel concentrator

    International Nuclear Information System (INIS)

    A new solar boiling water system with conventional vacuum-tube solar collector as primary heater and the holistic solar funnel concentrator as secondary heater had been designed. In this paper, the system was measured out door and its performance was analyzed. The configuration and operation principle of the system are described. Variations of the boiled water yield, the temperature of the stove and the solar irradiance with local time have been measured. Main factors affecting the system performance have been analyzed. The experimental results indicate that the system produced large amount of boiled water. And the performance of the system has been found closely related to the solar radiance. When the solar radiance is above 600 W/m2, the boiled water yield rate of the system has reached 20 kg/h and its total energy efficiency has exceeded 40%.

  2. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1975

    International Nuclear Information System (INIS)

    The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1975. The report includes 1169 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Seven of the unique events that occurred during the year are reviewed in detail

  3. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1975

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.L.; Gallaher, R.B.

    1976-07-01

    The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1975. The report includes 1169 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Seven of the unique events that occurred during the year are reviewed in detail.

  4. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1976

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.L.; Gallaher, R.B.

    1977-08-02

    This bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1976. The report includes 1,253 abstracts that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Three of the unique events that occurred during the year are reviewed in detail.

  5. Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980

    Energy Technology Data Exchange (ETDEWEB)

    McCormack, K.E.; Gallaher, R.B.

    1982-03-01

    This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

  6. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1976

    International Nuclear Information System (INIS)

    This bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1976. The report includes 1,253 abstracts that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Three of the unique events that occurred during the year are reviewed in detail

  7. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  8. Critical heat flux of an impinging water jet on a heated surface with boiling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.S. [Andong Institute of Informaion Technology, Andong (Korea); Kim, H.D. [Andong National University, Andong (Korea); Choi, K.W. [Incheon University, Incheon (Korea)

    2000-04-01

    The purpose of this paper is to investigate a critical heat flux(CHF) during forced convective subcooled and saturated boiling in free water jet system impinged on a rectangular heated surface. The surface is supplied with subcooled or saturated water through a rectangular jet. Experimental parameters studied are a width of heated surface, a height of supplementary water and a degree of subcooling. Incipient boiling point is observed in the temperature of 6{approx}8 deg.C of superheat of test specimen. CHF depends on jet velocity for various boiling-involved coolant system. CHF also is proportional to the nozzle exit velocity to the power of n, where n is 0.55 and 0.8 for subcooled and saturated boiling, respectively. CHF is enhanced with a higher jet velocity, higher degree of subcooling and smaller width of a heated surface. (author). 18 refs., 13 figs., 1 tab.

  9. Pressure measurements in boiling particle beds with water at 1 bar

    International Nuclear Information System (INIS)

    Pressures have been measured at the top and bottom of uniformly heated beds of uniform spherical particles with water boiling at atmospheric pressure. Particle sizes used vary from 0.22 to 5 mm diameter and bed heights from 50 to 150 mm. The pressures have been recorded at power levels up to dry-out. The results show how much liquid remains in a boiling bed at different power levels and how the liquid/vapour phase pressure losses vary. The results give a valuable insight into the working of a boiling bed. (author)

  10. 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.)

  11. 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)

  12. 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)

  13. Effects of Boiling Water Temperature on Biofilm Formation in PTI Community Potable Water

    Directory of Open Access Journals (Sweden)

    E. A. Fadairo

    2015-04-01

    Full Text Available This study investigated the effects of boiling temperature and associated physico-chemical parameters on the Petroleum Training Institute potable water and the possibility of biofilm formation in its delivery systems. A total of 25 potable water samples were used for this study. The environmental parameters investigated were pH, conductivity, total dissolved solids (TDS, total suspended solid (TSS, dissolved oxygen (DO1, / DO5, salinity, resistivity, total coliform bacteria (as an indicator of possible biofilm presence in the distribution system and biofilm . An overall prevalence of <1 of the total coliform bacteria was observed in the plus-boiling and minus-boiling potable water sample, except for the female hostel which showed moderate stain for the qualitative biofilm test. For the minus-boiling water sample, pH values were between 5.04±0.47 to 6.82±0.48; Total suspended solids ranged between 0.09±0.05-0.17±0.02, total dissolved solid ranged between 4.07±0.73 to 5.58±0.70, conductivity values ranged between 8.02±0.90 to 11.54±1.67, dissolved oxygen ranged between 1.97±0.26 to 3.12 ±0.13, the DO5 ranged between 1.91±0.32 to 2.72± 0.29 while resistivity ranged between 7.79±0.13 to 10.88±0.18. Values for the Plus-boiling and filtered samples showed a pH range of 6.02±0.26 to 6.95±0.26; conductivity 7.21±0.10 to 9.88±0.67; DO ranged between 1.01±0.14 to 2.08±0.35, DO 5 was 1.02±0.02 to 2.01±0.38, TSS and TDS ranged between 0.02±0.001, 3.74±0.62 to 0.03±0.002 and 4.95±0.42 respectively while resistivity ranged between 1.02±0.11 to 1.98±0.16. For all parameters analyzed, values obtained falls within the WHO limit for potable water except for the qualitative biofilm test on FSH minus-boiling water sample which gave moderate stain with 0.1% crystal violet stain and the pH values which fall below WHO acceptable limits. Boiling and filtration of potable water irrespective of the source is campaigned from this study in order

  14. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  15. Experimental study of the characteristics of pool boiling CHF enhancement using water-based magnetic fluid

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Hyuk; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2011-05-15

    Nucleate boiling is a very effective heat transfer mechanism. However, there exists a critical value of heat flux at which nucleate boiling transitions to film boiling shows very poor heat transfer behavior. Critical heat flux(CHF) is a main constraint to the design process because it can generate damages or deformations of material. There have been many efforts to improve the CHF by using nanofluids by researchers. This paper will describe the effects of magnetic fluid on CHF enhancement of pool boiling. We compared the CHF values of pool boiling experiment between magnetic fluid and other nanofluids with several volume concentrations to evaluate the degree of CHF enhancement. SEM(Scanning Electron Microscope) images were obtained to explain CHF enhancement through the effect of the deposited nanoparticles, which can change the surface wettability, during the pool boiling experiment. Lastly, Finally, in order to investigate the effect of magnetic field in the water-based magnetic fluid, magnetic field was analytically calculated by using Biot-Savart law. Using these results, we discussed the CHF enhancement of magnetite-water nanofluids in detailed

  16. Integrated modular water reactor: IMR

    International Nuclear Information System (INIS)

    The Mitsubishi Heavy Industries, Ltd. Has investigated on a concept on small scale reactor with economical efficiency comparable with large scale one. Aims of development on the integrated modular water reactor (IMR) of a small scale reactor plant concept consist in large construction cost reduction through adoption of technique specific to the small scale reactor and integrated production of plural units and in establishment of high safety target without reality in a large scale reactor to realize reduction of operation and maintenance costs by this reduction to simplification of operation and maintenance. Its concrete developmental targets are to make an integrated reactor with vessel size actually producible and the largest output, to remove feasibility of coolant loss accident (LOCA), to remove an accident with feasibility related to fuel fracture, to remove feasibility of nuclear reactor coolant to leak out from a storage vessel, to secure safety of plant without necessity of human and physical assistances from other plants at all on an accident, to make numbers of operators per unit output equal to those of large scale reactor, and to make working amounts at maintenance per unit output equal to large scale reactor by simplification of apparatus practice of rotation on main apparatus such as SG, and so on. Here were described on design concept and plan to realization. (G.K.)

  17. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de inyeccion de agua de refrigeracion a baja presion (LPCI) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C., E-mail: renedelgado2015@hotmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

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

  18. Pool boiling of water on nano-structured micro wires at sub-atmospheric conditions

    Science.gov (United States)

    Arya, Mahendra; Khandekar, Sameer; Pratap, Dheeraj; Ramakrishna, S. Anantha

    2015-10-01

    Past decades have seen active research in enhancement of boiling heat transfer by surface modifications. Favorable surface modifications are expected to enhance boiling efficiency. Several interrelated mechanisms such as capillarity, surface energy alteration, wettability, cavity geometry, wetting transitions, geometrical features of surface morphology, etc., are responsible for change in the boiling behavior of modified surfaces. Not much work is available on pool boiling at low pressures on microscale/nanoscale geometries; low pressure boiling is attractive in many applications wherein low operating temperatures are desired for a particular working fluid. In this background, an experimental setup was designed and developed to investigate the pool boiling performance of water on (a) plain aluminum micro wire (99.999 % pure) and, (b) nano-porous alumina structured aluminum micro wire, both having diameter of 250 µm, under sub-atmospheric pressure. Nano-structuring on the plain wire surface was achieved via anodization. Two samples, A and B of anodized wires, differing by the degree of anodization were tested. The heater length scale (wire diameter) was much smaller than the capillary length scale. Pool boiling characteristics of water were investigated at three different sub-atmospheric pressures of 73, 123 and 199 mbar (corresponding to T sat = 40, 50 and 60 °C). First, the boiling characteristics of plain wire were measured. It was noticed that at sub-atmospheric pressures, boiling heat transfer performance for plain wire was quite low due to the increased bubble sizes and low nucleation site density. Subsequently, boiling performance of nano-structured wires (both Sample A and Sample B) was compared with plain wire and it was noted that boiling heat transfer for the former was considerably enhanced as compared to the plain wire. This enhancement is attributed to increased nucleation site density, change in wettability and possibly due to enhanced pore scale

  19. Pool boiling of water on nano-structured micro wires at sub-atmospheric conditions

    Science.gov (United States)

    Arya, Mahendra; Khandekar, Sameer; Pratap, Dheeraj; Ramakrishna, S. Anantha

    2016-09-01

    Past decades have seen active research in enhancement of boiling heat transfer by surface modifications. Favorable surface modifications are expected to enhance boiling efficiency. Several interrelated mechanisms such as capillarity, surface energy alteration, wettability, cavity geometry, wetting transitions, geometrical features of surface morphology, etc., are responsible for change in the boiling behavior of modified surfaces. Not much work is available on pool boiling at low pressures on microscale/nanoscale geometries; low pressure boiling is attractive in many applications wherein low operating temperatures are desired for a particular working fluid. In this background, an experimental setup was designed and developed to investigate the pool boiling performance of water on (a) plain aluminum micro wire (99.999 % pure) and, (b) nano-porous alumina structured aluminum micro wire, both having diameter of 250 µm, under sub-atmospheric pressure. Nano-structuring on the plain wire surface was achieved via anodization. Two samples, A and B of anodized wires, differing by the degree of anodization were tested. The heater length scale (wire diameter) was much smaller than the capillary length scale. Pool boiling characteristics of water were investigated at three different sub-atmospheric pressures of 73, 123 and 199 mbar (corresponding to T sat = 40, 50 and 60 °C). First, the boiling characteristics of plain wire were measured. It was noticed that at sub-atmospheric pressures, boiling heat transfer performance for plain wire was quite low due to the increased bubble sizes and low nucleation site density. Subsequently, boiling performance of nano-structured wires (both Sample A and Sample B) was compared with plain wire and it was noted that boiling heat transfer for the former was considerably enhanced as compared to the plain wire. This enhancement is attributed to increased nucleation site density, change in wettability and possibly due to enhanced pore scale

  20. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - appendices. Final report

    International Nuclear Information System (INIS)

    The NRC staff is in need of decommissioning bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System's Washington Nuclear Plant Two (WNP-2) located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear structures on the site and to restore the site to a open-quotes green fieldclose quotes condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined

  1. Measurement of void fraction in flow boiling of ZnO–water nanofluids using image processing technique

    Energy Technology Data Exchange (ETDEWEB)

    Rana, K.B., E-mail: kunj.216@gmail.com [Department of Mechanical Engineering, Malaviya National Institute of Technology, Jaipur (India); Agrawal, G.D.; Mathur, J. [Department of Mechanical Engineering, Malaviya National Institute of Technology, Jaipur (India); Puli, U. [Faculty of Mechanical Engineering, Department of Technical Education, Government of Andhra Pradesh, Hyderabad (India)

    2014-04-01

    Highlights: • Void fraction during flow boiling of nanofluids measured using optical techniques. • Bubble behavior of nanofluids was investigated and compared with water. • Nanofluids showed lower void fraction as compared to water. • Void fraction decreases with increasing nanoparticle concentration and flow rate. • Void fraction increases with heat flux and axial location of heated length. - Abstract: In recent years, nanofluids have been an active area of research in many engineering applications, especially for nuclear reactor safety systems due to their enhanced thermal properties as a coolant. In this study, experiments were performed in subcooled flow boiling of water and ZnO–water nanofluids with different nanoparticle concentrations (0.001–0.01 vol.%) in horizontal annulus at heat fluxes varying from 100 to 550 kW/m{sup 2} and flow rates from 0.1 to 0.175 lps at 1 bar inlet pressure and constant subcooling of 20 °C to determine the void fraction by image processing technique. Parametric effects of nanoparticle volume fraction, heat flux, flow rate and axial location of heater rod on void fraction were studied. Bubble images during flow boiling were captured with high speed visualization and analyzed by National Instruments IMAQ Vision Builder 6.1 image processing software. Results show that void fraction decreases up to 86% with the use of nanofluid in place of water and it also decreases with increasing nanoparticle concentration and flow rate, whereas increase in heat flux and axial location of heater rod have opposite effect.

  2. R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

    Energy Technology Data Exchange (ETDEWEB)

    Vanderhaegen, M. [CEA, Nuclear Energy Directorate DEN, Laboratory of Instrumentation and Technological Test, Cadarache, 13108 Saint-Paul-lez-Durance (France); Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris (France); Paumel, K. [CEA, Nuclear Energy Directorate DEN, Laboratory of Instrumentation and Technological Test, Cadarache, 13108 Saint-Paul-lez-Durance (France); Seiler, J. M. [CEA, Nuclear Energy Directorate DEN, Laboratory of Physical Chemistry and Multiphase Thermalhydraulics, 38054 Grenoble (France); Tourin, A. [Laboratory of Waves and Acoustics, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris (France); Jeannot, J. P. [CEA, Nuclear Energy Directorate DEN, Laboratory of Instrumentation and Technological Test, Cadarache, 13108 Saint-Paul-lez-Durance (France); Rodriguez, G. [CEA, Nuclear Energy Directorate DEN, Cadarache, 13108 Saint-Paul-lez-Durance (France)

    2011-07-01

    In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boiling intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)

  3. Light water reactor safety research

    International Nuclear Information System (INIS)

    As the technology of light water reactors (LWR) was being commercialized, the German Federal Government funded the reactor safety research program, which was conducted by national research centers, universities, and industry, and which led to the establishment, in early 1972, of the Nuclear Safety Project in Karlsruhe. In the seventies, the PNS project mainly studied the loss-of-coolant accident. Numerous experiments were run and computer codes developed for this purpose. In the eighties, the Karlsruhe Nuclear Research Center contributed to the German Risk Study, investigating especially core meltdown accidents under the impact of the events at Three Mile Island-2 and Chernobyl-4. Safety research in the nineties is concentrated on the requirements of future reactor generations, such as the European Pressurized Water Reactor (EPR) or potential approaches which, at the present time, are discernible only as tentative theoretical designs. (orig.)

  4. Water shielding nuclear reactor container

    International Nuclear Information System (INIS)

    The reactor container of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevated inner pressure and keeping airtightness, and shielding water is filled inside from a water injection port. It is endurable to a great inner pressure satisfactorily and keep airtightness by the two spaced relatively thin steel plates. It exhibits radiation shielding effect by filling water substantially the same as that of a conventional reactor container made of iron reinforced concretes. Then, it is no more necessary to use concretes for the construction of the reactor container, which shortens the term of the construction, and saves the construction cost. In addition, a cooling effect for the reactor container is provided. Syphons are disposed contiguously to a water injection port and the top end of the syphon is immersed in an equipment temporarily storage pool, and further, pipelines are connected to the double steel plate walls or the syphons for supplying shielding water to enhance the cooling effect. (N.H.)

  5. Heat Transfer From Electrically Heated Nichrome Wires to Boiling Water at Different Pressures

    Directory of Open Access Journals (Sweden)

    Devi Dayal

    1968-01-01

    Full Text Available Boiling curves for nucleate and film boiling have been drawn for nichrome of three sizes in distilled and degasified water at saturation temperatures under five different sub-atmospheric vapour pressure. It has been observed that (i for the same Q/A (heat transfer, Delta Theta (excess of wire temperature over saturation point of water decreases with pressure in both nucleate and film boiling ranges, (ii Both Q/A max. and Delta Theta/SubC show a rapid decrease with pressure but these variations become more gradual at higher pressures, and (iii Q/A max. and Delta Theta/SubC increase with wire size at all pressures; increase in Delta Theta/SubC however, becomes less conspicuous at higher pressures approaching one atmosphere.

  6. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  7. OECD/NRC Boiling Water Reactor Turbine Trip Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Coddington, P

    2001-11-01

    The meeting was opened by Dr. Paul Coddington from PSI, the Chairman of the Organising and Program Committee of the Workshop. He chaired the Introductory Session 1. The deputy director of the PSI Nuclear Energy Department, Dr. Konstantin Foskolos, welcomed the participants on the behalf of PSI and gave an overview of the nuclear research performed at PSI. The participants introduced themselves and the organisations they represented. 26 participants representing 15 organisations from 8 countries attended the 2nd Workshop. The actual number of participants having submitted results and presentations for the Second Workshop was larger than the number of attendees. The list of participants is provided in Annex 1. Some participants (denoted by stars) had to cancel their trip but they submitted presentations. The agenda, provided as Annex 2, was reviewed and adopted after minor changes. (author)

  8. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-31

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  9. Analysis of the dynamics of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    The March-Leuba lineal reduced model is represented mathematically by a differential equations system, which corresponds to the direct transfer function, punctual kinetics approximation, neutron field dynamics, heat transfer in fuels, and channel dynamics approximation that relates the fuel temperature changes to the reactivity changes by vacuums. The model presents significant differences in one of the equation coefficients. The Pade order approximation used for the equation deduction for the channel has a different behavior to the exponential one for long periods of bubble residence. (Author)

  10. Radiation effects in organic paints of a Boiling Water Reactor

    International Nuclear Information System (INIS)

    The coatings on a BWR are used as a protection for the building and equipments from corrosion and contamination by radionuclides. The purpose of this work is to test this kind of coatings by simulating real absorbed doses in 40 years of use plus a nuclear accident (LOCA). Standards said that irradiation should be made with gamma radiation. In this work it's suggested to irradiate with electrons simulating secondary radiation produced on the interaction gamma-matter, and protons simulating the damage caused by the interaction neutron-matter. It's also suggested a new kind of adhesion test for coatings that gives a quantitative measure all other tests are qualitative. Two types of coatings were tested: Modified Phenolic and Epoxic both kinds had a very satisfactory performance in all the tests. The maximum dose accumulated by the coatings was 450 Mrad and the minimum 50 Mrad. The dose rates were: gamma in between 0.4 Mrad/hr and 1.0 Mrad/hr; protons and electrons between 500 Mrad/hr and 4000 Mrad/hr. Other important fact is that a calibration was made for a polymer to be used as a high dose dosimeter, these new dosimeters can measure doses between 10 Mrad and 100 Mrad not depending on the dose rate. (author)

  11. Standard technical specifications for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on General Electric plants currently being reviewed for an Operating License. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. This revision of the GE-STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  12. Pressure suppression containment system for boiling water reactor

    Science.gov (United States)

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  13. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    International Nuclear Information System (INIS)

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  14. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

  15. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  16. Safety of thermal water reactors

    International Nuclear Information System (INIS)

    This book reports on the latest European research into the safety of thermal water reactors, based on the presentation and evaluation of results obtained from research projects undertaken in different national laboratories of the European Community. Information is included under the following areas of research: 1.) The loss of coolant accident (LOCA) and the functioning and performance of the emergency core cooling system; 2.) The protection of nuclear power plants against external gas cloud explosions; and 3.) The release and distribution of radioactive fission products in the atmosphere following a reactor accident

  17. Reactor water level control system

    International Nuclear Information System (INIS)

    A BWR type reactor comprises a control valve disposed in a reactor water draining pipelines and undergoing an instruction to control the opening degree, an operation board having a setting device for generating the instruction and a control board for giving the instruction generated by the setting device to the control valve. The instruction is supplied from the setting device to the control valve by way of a control circuit to adjust the opening degree of the control valve thereby controlling the water level in the reactor. In addition, a controller generating an instruction independent of the setting device and a signal transmission channel for signal-transmitting the instruction independent of the control circuit are disposed, to connect the controller electrically to the signal transmission. The signal transmission channel and the control circuit are electrically connected to the control valve switchably with each other. Since instruction can be given to the control valve even at a periodical inspection or modification when the setting device and the control circuit can not be used, the reactor water level can be controlled automatically. Then, operator's working efficiency upon inspection can be improved remarkably. (N.H.)

  18. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  19. Co-boiling of NAPLs and water during thermal remediation: experimental and modeling study

    Science.gov (United States)

    Krol, M.; Zhao, C.; Mumford, K. G.; Sleep, B. E.; Kueper, B. H.

    2015-12-01

    The persistence of non-aqueous-phase liquids (NAPLs) in the subsurface has led to the development of several remediation technologies to address this environmental problem. One such group of technologies (in situ thermal treatment) uses heat to volatilize contaminants. Subsurface temperature measurements are often used to monitor progress and optimize contaminant removal. However, when NAPL and water are heated together, gas is created at a temperature lower than the boiling point of either liquid (co-boiling), which can affect temperature observations. To examine the effect of co-boiling on observed temperatures and NAPL mass removal, a series of heated laboratory experiments were performed using single and multi-component NAPLs. The experiments consisted of glass jars filled with a mixture of sand, water, and NAPL mixed to obtain an approximately uniform NAPL distribution within the jar. The experiments were heated from the outside and interior temperatures were measured using a thermocouple. The tests showed that local-scale temperature measurements are unreliable in indicating the end of co-boiling and may not indicate complete mass removal. This is because a well-defined co-boiling plateau does not exist when heating a multi-component NAPL and the temperature is dependent on the proximity of NAPL to the monitoring point. To further investigate temperature distributions and the potential to use gas production as a complementary indicator of NAPL removal, a 2D finite-difference mass transport model was used that incorporated heat transport, latent heat, phase change, and a multicomponent gas phase and used a macroscopic invasion percolation (MIP) model to simulate gas movement. Latent heat was calculated by multiplying specific latent heat, which is an intrinsic property of a substance, by the amount of liquid mass being vaporized and its incorporation into the model allowed for the simulation of co-boiling plateaus (during single component NAPL boiling). The

  20. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  1. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  2. The risks of nuclear energy technology. Safety concepts of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Kern- und Energietechnk (IKET); Kessler, Guenter; Veser, Anke; Schlueter, Franz-Hermann

    2014-11-01

    Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

  3. The risks of nuclear energy technology. Safety concepts of light water reactors

    International Nuclear Information System (INIS)

    Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

  4. Boiling-up of liquid nitrogen jet in water

    Science.gov (United States)

    Nakoryakov, V. E.; Tsoi, A. N.; Mezentsev, I. V.; Meleshkin, A. V.

    2014-06-01

    The hydrodynamic processes occurring at injection of cryogenic liquid into water pool were studied experimentally. Processes accompanying the phase transitions were registered. Data testify the developing pressure burst with an amplitude sufficient for possible formation of gas hydrates when methane is injected as a cryogenic fluid.

  5. 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

  6. 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.

  7. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with ∼300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289 degrees C

  8. 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

  9. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

  10. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  11. Pressure drop, heat transfer, critical heat flux, and flow stability of two-phase flow boiling of water and ethylene glycol/water mixtures - final report for project "Efficent cooling in engines with nucleate boiling."

    Energy Technology Data Exchange (ETDEWEB)

    Yu, W.; France, D. M.; Routbort, J. L. (Energy Systems)

    2011-01-19

    Because of its order-of-magnitude higher heat transfer rates, there is interest in using controllable two-phase nucleate boiling instead of conventional single-phase forced convection in vehicular cooling systems to remove ever increasing heat loads and to eliminate potential hot spots in engines. However, the fundamental understanding of flow boiling mechanisms of a 50/50 ethylene glycol/water mixture under engineering application conditions is still limited. In addition, it is impractical to precisely maintain the volume concentration ratio of the ethylene glycol/water mixture coolant at 50/50. Therefore, any investigation into engine coolant characteristics should include a range of volume concentration ratios around the nominal 50/50 mark. In this study, the forced convective boiling heat transfer of distilled water and ethylene glycol/water mixtures with volume concentration ratios of 40/60, 50/50, and 60/40 in a 2.98-mm-inner-diameter circular tube has been investigated in both the horizontal flow and the vertical flow. The two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux of the test fluids were determined experimentally over a range of the mass flux, the vapor mass quality, and the inlet subcooling through a new boiling data reduction procedure that allowed the analytical calculation of the fluid boiling temperatures along the experimental test section by applying the ideal mixture assumption and the equilibrium assumption along with Raoult's law. Based on the experimental data, predictive methods for the two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux under engine application conditions were developed. The results summarized in this final project report provide the necessary information for designing and implementing nucleate-boiling vehicular cooling systems.

  12. A meta-analysis of public compliance to boil water advisories.

    Science.gov (United States)

    Vedachalam, Sridhar; Spotte-Smith, Kyra T; Riha, Susan J

    2016-05-01

    Water utilities that generally provide continuous and reliable service to their customers may sometimes issue an advisory notification when service is interrupted or water quality is compromised. When the contamination is biological, utilities or the local public health agencies issue a 'boil water advisory' (BWA). The public health effectiveness of a BWA depends strongly on an implicit public understanding and compliance. In this study, a meta-analysis of 11 articles that investigated public compliance to BWA notifications was conducted. Awareness of BWA was moderately high, except in situations involving extreme weather. Reported rates of compliance were generally high, but when rate of awareness and non-compliant behavior such as brushing teeth were factored in, the median effective compliance rate was found to be around 68 percent. This does not include situations where people forgot to boil water for some part of the duration, or ingested contaminated water after the BWA was issued but before they became aware of the notification. The two-thirds compliance rate is thus an over-estimate. Results further suggest that timeliness of receipt, content of the advisory, and number of sources reporting the advisory have a significant impact on public response and compliance. This analysis points to improvements in the phrasing and content of BWA notices that could result in greater compliance, and recommends the use of a standard protocol to limit recall bias and capture the public response accurately. PMID:26938499

  13. Oscillate Boiling

    CERN Document Server

    Li, Fenfang; Nguyen, Dang Minh; Ohl, Claus-Dieter

    2016-01-01

    We report about an intriguing boiling regime occurring for small heaters embedded on the boundary in subcooled water. The microheater is realized by focusing a continuous wave laser beam to about $10\\,\\mu$m in diameter onto a 165\\,nm-thick layer of gold, which is submerged in water. After an initial vaporous explosion a single bubble oscillates continuously and repeatably at several $100\\,$kHz. The microbubble's oscillations are accompanied with bubble pinch-off leading to a stream of gaseous bubbles into the subcooled water. The self-driven bubble oscillation is explained with a thermally kicked oscillator caused by the non-spherical collapses and by surface pinning. Additionally, Marangoni stresses induce a recirculating streaming flow which transports cold liquid towards the microheater reducing diffusion of heat along the substrate and therefore stabilizing the phenomenon to many million cycles. We speculate that this oscillate boiling regime may allow to overcome the heat transfer thresholds observed dur...

  14. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo a baja presion (LPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Membrillo G, O. E.; Chavez M, C., E-mail: garzo1012@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

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

  15. Flow-induced vibration for light water reactors. Program final report

    International Nuclear Information System (INIS)

    The Flow-Induced Vibrations for Light Water Reactors Program was a five-year effort to develop basic knowledge and understanding needed to improve the flow-induced vibration design of light water reactors. Major tasks included analytical and test investigations of the flow-induced vibration of cylinders in isolation and arrays in smooth and turbulent single-vibration testing of reactor components including reactor inlet plenum components, jet pumps, low-pressure coolant injection coupling, and fuel rods. Cases of self-excited limit cycle response were encountered which required design modifications. The fuel rod tests were made in axially flowing water and steam/water mixtures under adiabatic and boiling conditions

  16. Dimensional analysis of boiling heat transfer burnout conditions

    International Nuclear Information System (INIS)

    The first criteria in boiling water systems design, such as boiling water reactors, is that no burnout in the core is allowed to exist under any conditions of the reactor operation either during steady state operation or during any of the several postulated accidental transients, such as sudden interruption of coolant flow in the reactor core (due to pump failure or blockage of fuel channel). The aim of the present work is to obtain a correlation for the critical heat flux based on a theoretical study where the mechanism of burn out and the related hydrodynamic and heat transfer equations are considered. 8 refs

  17. Experimental Research on Water Boiling Heat Transfer on Horizontal Copper Rod Surface at Sub-Atmospheric Pressure

    Directory of Open Access Journals (Sweden)

    Li-Hua Yu

    2015-09-01

    Full Text Available In recent years, water (R718 as a kind of natural refrigerant—which is environmentally-friendly, safe and cheap—has been reconsidered by scholars. The systems of using water as the refrigerant, such as water vapor compression refrigeration and heat pump systems run at sub-atmospheric pressure. So, the research on water boiling heat transfer at sub-atmospheric pressure has been an important issue. There are many research papers on the evaporation of water, but there is a lack of data on the characteristics at sub-atmospheric pressures, especially lower than 3 kPa (the saturation temperature is 24 °C. In this paper, the experimental research on water boiling heat transfer on a horizontal copper rod surface at 1.8–3.3 kPa is presented. Regression equations of the boiling heat transfer coefficient are obtained based on the experimental data, which are convenient for practical application.

  18. Study of Pu consumption in Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    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

  19. Burnout in the boiling of water and freon-113 on tubes with annular fins

    International Nuclear Information System (INIS)

    This paper presents the results of numerical calculations of burnout heat flux associated with the boiling of Freon-113 and water on an annular fin of constant thickness which have been approximated by simple analytical relations. These are used to calculate the critical burnout parameters of tubes with an annular fin assembly. The calculated data may be used for the analysis of tubes with an annular fin assembly over a wide range of variation of the thermophysical properties of the material and geometrical parameters of the fin assembly

  20. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

    OpenAIRE

    Diego Mandelli; Steven Prescott; Curtis Smith; Andrea Alfonsi; Cristian Rabiti; Joshua Cogliati; Robert Kinoshita

    2015-01-01

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to...

  1. Production induced boiling and cold water entry in the Cerro Prieto geothermal reservoir indicated by chemical and physical measurements

    Energy Technology Data Exchange (ETDEWEB)

    Grant, M.A. (DSIR, Wellington, New Zealand); Truesdell, A.H.; Manon, A.

    1981-01-01

    Chemical and physical data suggest that the relatively shallow western part of the Cerro Prieto reservoir is bounded below by low permeability rocks, and above and at the sides by an interface with cooler water. There is no continuous permeability barrier around or immediately above the reservoir. Permeability within the reservoir is dominantly intergranular. Mixture with cooler water rather than boiling is the dominant cooling process in the natural state, and production causes displacement of hot water by cooler water, not by vapor. Local boiling occurs near most wells in response to pressure decreases, but no general vapor zone has formed.

  2. Chemistry control strategies for a supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    The long-term viability of any Generation IV Supercritical Water-cooled Reactor (SCWR) concept depends on the ability of reactor designers and operators to predict and control water chemistry to minimize corrosion and corrosion product transport. Currently, SCWR material testing is being carried out using a limited range of water chemistries to establish the dependencies of the corrosion behavior on parameters such as water temperature and dissolved oxygen concentration. Once a final suite of candidate alloys is identified, more detailed, longer term testing will be needed under water chemistry regimes expected to be used in the SCWR. Prior to these tests, it will be necessary to identify proposed water chemistry regimes for the SCWR, and provide expected ranges for the key parameters. The direct-cycle design proposed for various SCWR concepts take advantage of the extensive operating experience world-wide of fossil-fired SCW power plants. Conceptually, the SCWR replaces the fossil-fired boiler with the reactor core (pressure vessel or pressure tube); the concept is broadly similar to that of a boiling water reactor. Current fossil-fired SCW power plants use either an all-volatile treatment or oxygenated water treatment for the feedwater systems. While the optimal water chemistry for a SCWR is yet to be determined, the monitored parameters are likely to be the same as those in existing fossil-fired and nuclear power plants: pH; conductivity, and concentrations of O2, H2, additives, impurities, corrosion and activation products. Monitoring will be required at many of the same components: feedwater, main 'steam', drains, pump outlets, condenser hotwell, and purification inlets and outlets. This paper outlines the strategy being used to develop a water chemistry regime for a CANDU® SCWR. It describes the key areas identified for chemistry monitoring and control: a) the reactor core, where materials are subjected to irradiation and high temperature, b

  3. Effect of subcooling and wall thickness on pool boiling from downward-facing curved surfaces in water

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, M.S.; Glebov, A.G. [Univ. of New Mexico, Albuquerque, NM (United States)

    1995-09-01

    Quenching experiments were performed to investigate the effects of water subcooling and wall thickness on pool boiling from a downward-facing curved surface. Experiments used three copper sections of the same diameter (50.8 mm) and surface radius (148 mm), but different thickness (12.8, 20 and 30 mm). Local and average pool boiling curves were obtained at saturation and 5 K, 10 K, and 14 K subcooling. Water subcooling increased the maximum heat flux, but decreased the corresponding wall superheat. The minimum film boiling heat flux and the corresponding wall superheat, however, increased with increased subcooling. The maximum and minimum film boiling heat fluxes were independent of wall thickness above 20 mm and Biot Number > 0.8, indicating that boiling curves for the 20 and 30 thick sections were representative of quasi steady-state, but not those for the 12.8 mm thick section. When compared with that for a flat surface section of the same thickness, the data for the 12.8 mm thick section showed significant increases in both the maximum heat flux (from 0.21 to 0.41 MW/m{sup 2}) and the minimum film boiling heat flux (from 2 to 13 kW/m{sup 2}) and about 11.5 K and 60 K increase in the corresponding wall superheats, respectively.

  4. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  5. TA-2 Water Boiler Reactor Decommissioning Project

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, M.E. (ed.); Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

  6. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  7. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  8. Flow boiling heat transfer of ammonia/water mixture in a plate heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Taboas, Francisco [Universidad de Cordoba, Campus de Rabanales, Edificio Leonardo da Vinci, 14014 Cordoba (Spain); Valles, Manel; Bourouis, Mahmoud; Coronas, Alberto [CREVER - Universitat Rovira i Virgili, Av. Paisos Catalans No. 26, 43007 Tarragona (Spain)

    2010-06-15

    The objective of this work is to contribute to the development of plate heat exchangers as desorbers for ammonia/water absorption refrigeration machines driven by waste heat or solar energy. In this study, saturated flow boiling heat transfer and the associated frictional pressure drop of ammonia/water mixture flowing in a vertical plate heat exchanger is experimentally investigated. Experimental data is presented to show the effects of heat flux between 20 and 50 kW m{sup -2}, mass flux between 70 and 140 kg m{sup -2} s{sup -1}, mean vapour quality from 0.0 to 0.22 and pressure between 7 and 15 bar, for ammonia concentration between 0.42 and 0.62. The results show that for the selected operating conditions, the boiling heat transfer coefficient is highly dependent on the mass flux, whereas the influence of heat flux and pressure are negligible mainly at higher vapour qualities. The pressure drop increases with increasing mass flux and quality. However, the pressure drop is independent of the imposed heat flux. (author)

  9. Non-equilibrium aspects of water reactor safety

    International Nuclear Information System (INIS)

    A critical review of the thermal non-equilibrium phase change models applicable to the water reactor safety calculations is presented. Four specific areas are discussed: (1) rapid depressurization (flashing), (2) post-dryout heat transfer, (3) direct contact condensation, and (4) sub-cooled boiling. The models have been judged from a mechanistic viewpoint, and the areas where further work is needed have been pointed out. It is suggested that when a phase change occurs due to a difference in the bulk fluid temperatures, the non-equilibrium phase change process can be viewed as a linear, non-homogeneous thermal relaxation phenomenon. Direct measurements of the actual phase change rates and the interfacial area densities are emphasized. Finally, a recommendation for the volumetric vapor generation rates applicable to the above four specific areas is given

  10. Light water reactor safety research project

    International Nuclear Information System (INIS)

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  11. Experimental studies of boiling heat transfer and dryout in heat generating particulate beds in water at 1 bar

    International Nuclear Information System (INIS)

    Boiling heat transfer and dryout occurring while a liquid permeates a bed of self-heated particulate material are phenomena of relevance to reactor safety since they control the rate of heat removal from beds of core debris. This report presents results from laboratory experiments in which water was the coolant and the particulate material was metal spheres, usually tin-plated iron shot, heated by passing low voltage alternating current laterally through them. The study covered bed depths up to 200 mm, and particle diameters up to 5.0 mm. Values of dryout heat flux obtained for beds of uniform particles are consistent with those obtained elsewhere using different heating methods. Stratified beds in which a layer of fine particles rests upon a bed of coarse particles can reduce the dryout heat flux to below the level appropriate to either particle size alone, and devices which aid the flow of liquid and/or vapour in a bed can greatly increase the dryout heat flux. The data exhibit a high degree of consistency, and thus will prove to be valuable in testing theoretical models. (U.K.)

  12. Nickel Catalyzed Conversion of Cyclohexanol into Cyclohexylamine in Water and Low Boiling Point Solvents

    Directory of Open Access Journals (Sweden)

    Yunfei Qi

    2016-04-01

    Full Text Available Nickel is found to demonstrate high performance in the amination of cyclohexanol into cyclohexylamine in water and two solvents with low boiling points: tetrahydrofuran and cyclohexane. Three catalysts, Raney Ni, Ni/Al2O3 and Ni/C, were investigated and it is found that the base, hydrogen, the solvents and the support will affect the activity of the catalyst. In water, all the three catalysts achieved over 85% conversion and 90% cyclohexylamine selectivity in the presence of base and hydrogen at a high temperature. In tetrahydrofuran and cyclohexane, Ni/Al2O3 exhibits better activity than Ni/C under optimal conditions. Ni/C was stable during recycling in aqueous ammonia, while Ni/Al2O3 was not due to the formation of AlO(OH.

  13. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code; Simulacion del transitorio de falla del controlador de agua de alimentacion en un reactor de agua en ebullicion con el codigo RAMONA-3B

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, J.L.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2005-07-01

    The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

  14. Reactor process water (PW) piping inspections, 1984--1990

    International Nuclear Information System (INIS)

    In July 1983, the NRC ordered the shutdown of five boiling water reactors (BWR's) because of concerns about reliability of ultrasonic examination for detecting intergranular stress corrosion cracking (IGSCC). These concerns arose because of leaking piping at Niagara Mohawk's Nine Mile Point which was attributed to IGSCC. The leaks were detected shortly after completion of ultrasonic examinations of the piping. At that time, the Dupont plant manager at Savannah River (SR) directed that investigations be performed to determine if similar problems could exist in SR reactors. Investigation determined that all conditions believed necessary for the initiation and propagation of IGSCC in austenitic stainless steel exist in SR reactor process water (PW) systems. Sensitized, high carbon, austenitic stainless steel, a high purity water system with high levels of dissolved oxygen, and the residual stresses associated with welding during construction combine to provide the necessary conditions. A periodic UT inspection program is now in place to monitor the condition of the reactor PW piping systems. The program is patterned after NRC NUREG 0313, i.e., welds are placed in categories based on their history. Welds in upgraded or replacement piping are examined on a standard schedule (at least every five years) while welds with evidence of IGSCC, evaluated as acceptable for service, are inspected at every extended outage (15 to 18 months). This includes all welds in PW systems three inches in diameter and above. Welds are replaced when MSCC exceeds the replacement criteria of more than twenty percent of pipe circumference of fifty percent of through-wall depth. In the future, we intend to perform flow sizing with automated UT techniques in addition to manual sizing to provide more information for comparison with future examinations

  15. Heat transfer and critical heat flux of subcooled water flow boiling in a short horizontal tube

    International Nuclear Information System (INIS)

    The steady-state turbulent heat transfer (THT) due to exponentially increasing heat inputs with various exponential periods (Q=Q0exp(t/τ), τ=6.55 to 21.81 s) were systematically measured with the flow velocities, u, of 4.15, 7.05, 10.07 and 13.50 m/s by an experimental water loop flow. Measurements were made on a 6 mm inner diameter, a 59.2 mm effective length and a 0.4 mm thickness of HORIZONTAL Platinum (Pt) circular test tube. The relation between the steady-state turbulent heat transfer and the flow velocity were clarified. The steady state nucleate boiling heat transfer (NBHT) and the steady state critical heat fluxes (CHFs) of the subcooled water flow boiling for HORIZONTAL SUS304 circular test tube were systematically measured with the flow velocities (u=3.94 to 13.86 m/s), the inlet subcoolings (ΔTsub,in=81.30 to 147.94 K), the inlet pressures (Pin=786.29 to 960.93 kPa) and the increasing heat input (Q0 exp(t/τ), τ=8.36 s). The HORIZONTAL SUS304 test tube of inner diameter (d=6 mm), heated length (L=59.4 mm), effective length (Leff=48.4 mm), L/d (=9.9), Leff/d (=8.06) and wall thickness (δ=0.5 mm) with surface roughness (Ra=3.89 μm) was used in this work. The NBHT and the steady state CHFs of the subcooled water flow boiling for the HORIZONTAL SUS304 test tube were clarified at the flow velocities u ranging from 3.94 to 13.86 m/s. The steady-state THT data, the NBHT ones and the steady state CHF ones were compared with the values calculated by authors' THT correlation, their NBHT ones and their transient CHF ones against outlet and inlet subcoolings based on the experimental data for the VERTICAL circular test tubes with the flow velocities u ranging from 4.0 to 42.4 m/s. The influences of test tube orientation on the THT, the NBHT and the subcooled flow boiling CHF are investigated into details and the widely and precisely predictable correlations of the THT, the NBHT and the transient CHFs against outlet and inlet subcoolings in a short

  16. Numerical investigation of water-based nanofluid subcooled flow boiling by three-phase Euler-Euler, Euler-Lagrange approach

    Science.gov (United States)

    Valizadeh, Ziba; Shams, Mehrzad

    2016-08-01

    A numerical scheme for simulating the subcooled flow boiling of water and water-based nanofluids was developed. At first, subcooled flow boiling of water was simulated by the Eulerian multiphase scheme. Then the simulation results were compared with previous experimental data and a good agreement was observed. In the next step, subcooled flow boiling of water-based nanofluid was modeled. In the previous studies in this field, the nanofluid assumed as a homogeneous liquid and the two-phase scheme was used to simulate its boiling. In the present study, a new scheme was used to model the nanofluid boiling. In this scheme, to model the nanofluid flow boiling, three phases, water, vapor and nanoparticles were considered. The Eulerian-Eulerian approach was used for modeling water-vapor interphase and Eulerian-Lagrangian scheme was selected to observe water-nanoparticle interphase behavior. The results from the nanofluid boiling modeling were validated with an experimental investigation. The results of the present work and experimental data were consistent. The addition of 0.0935 % volume fraction of nanoparticles in pure liquid boiling flow increases the vapor volume fraction at the outlet almost by 40.7 %. The results show the three-phase model is a good approach to simulate the nanofluid boiling flow.

  17. Performance of Charcoal Cookstoves for Haiti Part 1: Results from the Water Boiling Test

    Energy Technology Data Exchange (ETDEWEB)

    Booker, Kayje; Han, Tae Won; Granderson, Jessica; Jones, Jennifer; Lsk, Kathleen; Yang, Nina; Gadgil, Ashok

    2011-06-01

    In April 2010, a team of scientists and engineers from Lawrence Berkeley National Lab (LBNL) and UC Berkeley, with support from the Darfur Stoves Project (DSP), undertook a fact-finding mission to Haiti in order to assess needs and opportunities for cookstove intervention. Based on data collected from informal interviews with Haitians and NGOs, the team, Scott Sadlon, Robert Cheng, and Kayje Booker, identified and recommended stove testing and comparison as a high priority need that could be filled by LBNL. In response to that recommendation, five charcoal stoves were tested at the LBNL stove testing facility using a modified form of version 3 of the Shell Foundation Household Energy Project Water Boiling Test (WBT). The original protocol is available online. Stoves were tested for time to boil, thermal efficiency, specific fuel consumption, and emissions of CO, CO{sub 2}, and the ratio of CO/CO{sub 2}. In addition, Haitian user feedback and field observations over a subset of the stoves were combined with the experiences of the laboratory testing technicians to evaluate the usability of the stoves and their appropriateness for Haitian cooking. The laboratory results from emissions and efficiency testing and conclusions regarding usability of the stoves are presented in this report.

  18. Environmental Assessment for Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Rincon, Puerto Rico

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) proposes to consent to a proposal by the Puerto Rico Electric Power Authority (PREPA) to allow public access to the Boiling Nuclear Superheat (BONUS) reactor building located near Rincon, Puerto Rico for use as a museum. PREPA, the owner of the BONUS facility, has determined that the historical significance of this facility, as one of only two reactors of this design ever constructed in the world, warrants preservation in a museum, and that this museum would provide economic benefits to the local community through increased tourism. Therefore, PREPA is proposing development of the BONUS facility as a museum

  19. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  20. Efficient Water Management in Water Cooled Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property.' The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. Water scarcity is becoming one of the most pressing crises affecting the planet. A reliable supply of water and energy is an important prerequisite for sustainable development. A large number of nuclear power reactors are being planned in many developing countries to address these countries' increasing energy demands and their limited fossil resources. New construction is expected in the USA, Europe and Asia, as well. Reducing water use and consumption by nuclear power plants is likely to help developing countries in introducing nuclear power into their energy supply mix. A large

  1. Liquid nitrogen - water interaction experiments for fusion reactor accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, R.; Murphy, J.; Pfotenhauer, J.; Corradini, M. [Wisconsin Univ., Dept. of Engineering Physics, Madison, WI (United States)

    2001-07-01

    With the implementation of superconducting magnets in fusion reactors, the possibility exists for the interaction between water and cryogenic systems. The interaction between liquid nitrogen and water was investigated experimentally and numerically. The rate of pressurization and peak pressure were found to be driven thermodynamically by the expansion of the water and the boil-off of the liquid nitrogen and did not have a vapor explosion nature. Since the peak pressure was small in comparison to previous work with stratified geometries, the role of the geometry of the interacting fluids has been shown to be significant. Comparisons of the peak pressure and the rate of pressurization with respect to the ratio of the liquid nitrogen mass to water mass reveal no functional dependence as was observed in the liquid helium-water experiments. A simple thermodynamic model provides a fairly good description of the pressure rise data. From the data, the model will allow one to extract the interaction area of the water. As with previous liquid helium-water interaction experiments, more extensive investigation of the mass ratio and interaction geometry is needed to define boundaries between explosive and non-explosive conditions. (authors)

  2. Development of integrated modular water reactor

    International Nuclear Information System (INIS)

    In order to adapt for environmental problem to reduce emission of greenhouse effect gas and develop a power generation plant with economical efficiency and small output, a number of R and Ds on small scale reactors have been progressed without any practice. The largest subject on the development consists in how cost of construction, operation and maintenance on the small scale reactors can be reduced to those of large scale ones by its specific technology. Therefore, if by adding wide simplification of apparatuses based on introduction of novel technology and added values specific to the small scale reactors, a business model on installing many small scale reactors can be established, a practicable feasibility of the small scale reactors valuable to introduction of actual scale machine will enable to be found. Authors have progressed development on a novel small scale reactor capable of flexibly corresponding to social needs for nuclear power generation. As a result, by integrating the reactor system and introducing self pressurisation and natural circulation as an innovative 300,000 kW output class small scale reactor, a plant concept reducing feasibility to occur any large scale accident to its ultimate limit together with planning wide simplification of apparatus could be established. The reactor is the titled integrated modular water reactor (IMR), and is at present under investigation on its formability confirmation and its concept design. Here were reported on plant concepts and characteristics of IMR, in this report. (G.K.)

  3. Experimental and theoretical study of a particular type of transient flow of boiling sodium: flow excursion. Study carried out in the framework of the fast neutron reactor safety

    International Nuclear Information System (INIS)

    The flow excursion phenomenon was studied in the framework of safety analyses for sodium cooled fast reactors, as a consequence of an accident of pump breakdown without safety rod drop. The experimental study, performed on an out-of-pile circuit with sodium forced convection allowed the flow excursion due to boiling appearance in a single heating channel feeded in parallel with a by-pass to be investigated. The first phase is characterized by a relatively slow decay of the (mean) flow rate through the channel from its initial value (corresponding to starting boiling) up to a mean value near zero; its duration, rather important, is of the order of ten or thirty seconds. The second phase appears when the boiling zone occupies an important fraction (half or three quarters) of the heating length of the channel. The mean flow rate has a low value, near zero; the flow rate is oscillatory (chugging) and is formed of vapor plugs parted by liquid bundles. This chugging phase lasts two or three seconds. The computation model developed described simply the boiling liquid flow (homogeneous sliding model) and the heat transfer between the heating medium and the fluid. The transient terms are neglected in continuity and energy equations so as the fluid in the channel is incompressible and its thermal inertia nul

  4. Emergency action levels for light water reactors. Draft report

    International Nuclear Information System (INIS)

    United States nuclear power plants are currently preparing revised radiological emergency response plans. As part of these plans, each licensee is establishing plant-specific instrumentation readings, called emergency action levels (EALs), that indicate the presence of a problem and the need to declare one of four classes of emergency as designated by the U.S. Nuclear Regulatory Commission. In this report, the EALs prepared for a pressurized water reactor and a boiling water reactor are assessed to determine whether they meet the requirements of NUREG-0654, Rev. 1 Appendix 1. In addition, five recent nuclear incidents are studied to determine how their outcomes might have been affected by the new emergency response plans. Most of the EALs reviewed were judged adequate. In some cases, alternate EALs or sets of generic EALs are proposed to meet the intent of NUREG-0654 Rev 1 Appendix 1, and the wording of some of the initiating conditions for the four emergency classes is clarified. For three of the five incidents reviewed, the new emergency response plans probably would not have affected the outcome; in one case, the plans would probably have prevented core damage; and any effect on the outcome of the last incident is uncertain

  5. Transient following partial loss of feed water for thorium based natural circulation reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 920 MWth vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all allowed power levels with no primary coolant pumps. Apart from this passive design feature passive safety systems in AHWR include isolation condenser (IC) system for decay heat removal in case of unavailability of main steam condenser, emergency core cooling (includes both high pressure and low pressure ECC) system, Passive containment cooling system, Passive containment isolation and automatic depressurization system. Further, reactor core has negative void coefficient of reactivity at all power level which enhances the safety of the reactor. The Primary Heat Transport System of the reactor consists of reactor core, core inlet and outlet core bottom extensions, inlet feeders, tailpipes, steam drums, downcomer and inlet header. One BFP trip transient without standby pump initiation has been analysed. This partial loss of feed scenario leads to reactor trip and subsequent decay heat removal takes place through isolation condenser path. All other thermal hydraulic parameters remain within safety limits

  6. CFD modelling of subcooled flow boiling for nuclear engineering applications

    International Nuclear Information System (INIS)

    In this work a general-purpose CFD code CFX-5 was used for simulations of subcooled flow boiling. The subcooled boiling model, available in a custom version of CFX-5, uses a special treatment of the wall boiling boundary, which assures the grid invariant solution. The simulation results have been validated against the published experimental data [1] of high-pressure flow boiling in a vertical pipe covering a wide range of conditions (relevant to the pressurized water reactor). In general, a good agreement with the experimental data has been achieved. To adequately predict the lateral distribution of two-phase flow parameters, the modelling of two-phase flow turbulence and non-drag forces under wall boiling conditions have been also investigated in the paper. (author)

  7. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  8. Transient boiling and void formation during postulated reactivity-initiated accident in BWR: Experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (author)

  9. Light water cooled, high temperature and high performance nuclear power plants concept of once-through coolant cycle, supercritical-pressure, light water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Supercritical-pressure, light water cooled nuclear reactors corresponding to nuclear reactors of once-through boilers, are of theoretical development from LWR. Under supercritical pressure, a steam turbine can be driven directly with cooled water with high enthalpy, as not seen boiling and required for recycling. The reactor has no steam-water separation and recycling systems on comparison with the boiling water type LWR, and is the same once-through type as supercritical-pressure thermal power generation plants. Then, all of cooling water at reactor core are sent to turbine. The reactor has no steam generator, and pressurizer, on comparison with PWR. As it requires no steam-water separator, steam drier, and recycling system on comparison with BWR, it becomes of smaller size and has shape and size nearly equal to those of PWR. And, its control bars can be inserted from upper direction like PWR, and can use its driving system. Here was introduced some concepts on high-temperature and high-performance light water reactor, nuclear power generation using a technology on supercritical-pressure thermal power generation. (G.K.)

  10. The Effect of Different Boiling and Filtering Devices on the Concentration of Disinfection By-Products in Tap Water

    Directory of Open Access Journals (Sweden)

    Glòria Carrasco-Turigas

    2013-01-01

    Full Text Available Disinfection by-products (DBPs are ubiquitous contaminants in tap drinking water with the potential to produce adverse health effects. Filtering and boiling tap water can lead to changes in the DBP concentrations and modify the exposure through ingestion. Changes in the concentration of 4 individual trihalomethanes (THM4 (chloroform (TCM, bromodichloromethane (BDCM, dibromochloromethane (DBCM, and bromoform (TBM, MX, and bromate were tested when boiling and filtering high bromine-containing tap water from Barcelona. For filtering, we used a pitcher-type filter and a household reverse osmosis filter; for boiling, an electric kettle, a saucepan, and a microwave were used. Samples were taken before and after each treatment to determine the change in the DBP concentration. pH, conductivity, and free/total chlorine were also measured. A large decrease of THM4 (from 48% to 97% and MX concentrations was observed for all experiments. Bromine-containing trihalomethanes were mostly eliminated when filtering while chloroform when boiling. There was a large decrease in the concentration of bromate with reverse osmosis, but there was a little effect in the other experiments. These findings suggest that the exposure to THM4 and MX through ingestion is reduced when using these household appliances, while the decrease of bromate is device dependent. This needs to be considered in the exposure assessment of the epidemiological studies.

  11. Boils (Furunculosis)

    Science.gov (United States)

    ... boil starts to drain, wash the area with antibacterial soap and apply some triple antibiotic ointment and a ... avoid spreading the infection to others. Use an antibacterial soap on boil-prone areas when showering, and dry ...

  12. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  13. Pressurized water reactor flow skirt apparatus

    Science.gov (United States)

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  14. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor)

  15. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

    This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided

  16. Corrosion failures and related research subjects in light water reactor power plants

    International Nuclear Information System (INIS)

    The purpose of this paper is to review the recent corrosion incidents and the remedial actions taken by the utility industry. The following corrosion failures and related subjects are presented in this paper. (1) Intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs). (2) Interdendritic stress corrosion cracking (IDSCC) of stainless steel piping weldments in BWRs. (3) Irradiation-assisted stress corrosion cracking (IASCC) in in-core component of nonsensitized austenitic stainless steel. (4) Intergranular attack (IGA) of steam generator tube material in pressurized water reactors (PWRs). (5) Corrosion fatigue and erosion corrosion of carbon steel piping. (6) SCC of reactor component materials. (7) Microbial induced corrosion. (author)

  17. Study on a decay heat removal system of light water reactors using air coolers

    International Nuclear Information System (INIS)

    In the present work, a passive decay heat removal system for light water reactors (LWRs) based on a new concept is studied referring to an air cooling system (ACS) of the fast breeder reactor Monju. The present study will contribute to the reduction of severe accident risks of nuclear power plants. In this system, a blower for an air cooler (AC) is operated using the rotation of a small steam turbine by generated steam in order to cool heat transfer tubes by forced convection of air. The purpose of the present work is to investigate the plant transient caused by a station blackout (SBO) using the plant system code NETFLOW++ and decay heat removal characteristics. A calculation model is the Advanced Boiling Water Reactor (ABWR) in Japan. (author)

  18. Modeling of microstructure evolution in austenitic stainless steels irradiated under light water reactor condition

    Science.gov (United States)

    Gan, J.; Was, G. S.; Stoller, R. E.

    2001-10-01

    A model for microstructure development in austenitic alloys under light water reactor irradiation conditions is described. The model is derived from the model developed by Stoller and Odette to describe microstructural evolution under fast neutron or fusion reactor irradiation conditions. The model is benchmarked against microstructure measurements in 304 and 316 SS irradiated in a boiling water reactor core using one material-dependent and three irradiation-based parameters. The model is also adapted for proton irradiation at higher dose rate and higher temperature and is calibrated against microstructure measurements for proton irradiation. The model calculations show that for both neutron and proton irradiations, in-cascade interstitial clustering is the driving mechanism for loop nucleation. The loss of interstitial clusters to sinks by interstitial cluster diffusion was found to be an important factor in determining the loop density. The model also explains how proton irradiation can produce an irradiated dislocation microstructure similar to that in neutron irradiation.

  19. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mausolf, Edward J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mcnamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schwantes, Jon M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  20. Coolant technology of water cooled reactors. V. 1: Chemistry of primary coolant in water cooled reactors

    International Nuclear Information System (INIS)

    This report is a summary of the work performed within the framework of the Coordinated Research Programme on Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors organized by the IAEA and carried out from 1987 to 1991. It is the continuation of a programme entitled Reactor Water Chemistry Relevant to Coolant-Cladding Interaction (IAEA-TECDOC-429), which ran from 1981 to 1986. Subsequent meetings resulted in the title of the programme being changed to Coolant Technology of Water Cooled Reactors. The results of this Coordinated Research Programme are published in four volumes with an overview in the Technical Reports Series. The titles of the volumes are: Volume 1: Chemistry of Primary Coolant in Water Cooled Reactors; Volume 2: Corrosion in the Primary Coolant Systems of Water Cooled Reactors; Volume 3: Activity Transport Mechanisms in Water Cooled Reactors; Volume 4: Decontamination of Water Cooled Reactors. These publications should be of interest to experts in water chemistry at nuclear power plants, experts in engineering, fuel designers, research and development institutes active in the field and to consultants to these organizations. Refs, figs and tabs

  1. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  2. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Hydrogen atom has two isotopes: deuterium 1H2 and tritium 1H3. The deuterium oxide D2O is called heavy water due to its density of 1105.2 Kg/m3. Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D2O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D2O management is required to preserve it. (author)

  3. The entropy balance for boiling flow

    Energy Technology Data Exchange (ETDEWEB)

    Collado, Francisco-Javier E-mail: fjk@posta.unizar.es

    2001-10-01

    Subcooled forced convection boiling of water is recognized as one of the best means of accommodating the very high heat fluxes that plasma facing components of fusion reactors have to withstand. The boiling curve, giving the wall temperature in function of the applied flux and flow conditions, is essential for the design of such cooling configurations. In this paper, a new entropy balance for subcooled boiling flow, which allows the wall temperature to be obtained, is presented and successfully compared with experimental data from the Joint US-EURATOM R and D Program. The derivation of this entropy balance is based on a new strict application of the Reynolds theorem to multiphase flows recently proposed by the author.

  4. Pool Boiling Behavior and Critical Heat Flux on Zircaloy and SiC Claddings in Deionized Water under Atmospheric Pressure

    International Nuclear Information System (INIS)

    Recently several researches on SiC material as an alternative of the nuclear fuel cladding have been conducted. From a fundamental point of view, Snead et al. did an extensive investigation on SiC properties. Their work revealed non-irradiated and irradiated material properties. In addition to the existing literature data, they even added new data, particularly in the high-temperature irradiation regime. Moreover, Carpenter has studied performance of a SiC fuel cladding in his Ph. D. thesis. With extensive in-core tests at MITR-II, his works showed the effects of cladding design for monolith and triplex types. He concluded that manufacturing techniques of the SiC cladding affected corrosion rates and swelling behavior after irradiation. For more practical nuclear applications, oxidation rates of a SiC cladding was investigated with a comparison assessment of those of a zircaloy-4 cladding. Lee et al. adopted an oxidation process under the conditions of the Loss of Coolant Accidents (LOCA) in LWRs. They found that SiC oxidation rates were greatly lower than those of zircaloy-4. In order to demonstrate the superiority of SiC cladding in terms of thermal performance, in this study pool boiling heat transfer experiments were carried out in a pool of saturated deionized water (DI water) at atmospheric pressure. For a comparison study, zircaloy-4 claddings, which are current fuel claddings in LWRs, were used as a reference case. Not only measuring nucleate boiling heat transfer coefficient (NBHTC) and critical heat flux (CHF) but also observing boiling behavior of both the claddings were conducted. In this study, pool boiling heat transfer experiments with zircaloy and SiC heaters were carried out. Comparison of the CHF and nucleate boiling heat transfer of the zircaloy-4 and SiC cladding were compared. Specifically, sophisticated high-speed photographs of nucleate boiling, the CHF, and film boiling phenomena were captured. · Structural integrity of the SiC heaters was

  5. Enhancement of CHF water subcooled flow boiling in tubes using helically coiled wires

    Science.gov (United States)

    Celata, G. P.; Cumo, M.; Mariani, A.

    1994-01-01

    The present paper reports the results of an experimental investigation about the occurrence of the critical heat flux (CHF) in subcooled flow boiling of water, carried out to ascertain the influence of thermal hydraulic parameters on CHF under conditions typical of themronuclear fusion divertor thermal hydraulic design. Helically coiled wires were used as turbulence promoters to enhance the CHF with respect to the smooth channel. Geometric characteristics of stainless steel 304 Type test sections were: 6.0 and 8.0 mm i.d., 0.25 mm wal thickness, 0.1 and 0.15 m heated length, horizontal and vertical (upflow) position. Test sections were uniformly heated using d.c. current. A maximum CHF of about 30 MW/sq m was reached with smooth tubes under the following conditions: T(sub in) = 30 C, p = 4.6 MPa, u = 10 m/s, D = 8.0 mm, L = 0.1 m. Helically coiled wires (d = 1.0 mm, pitch = 20.0 mm) allowed an increase of the CHF up to 50%, with reference to smooth channels, coupled with a moderate increase of pressure drop (down to 25%). Pressure revealed a negative effect on the efficiency of turbulence promoters. No observable influence of the channel orientation was detected.

  6. Physical insight in the burnout region of water-subcooled flow boiling

    International Nuclear Information System (INIS)

    The present paper reports the results of a visualization study of the burnout in subcooled flow boiling of water, with square cross-section annular geometry (formed by a central heater rod contained in a duct characterised by a square cross-section). In order to obtain clear pictures of the flow phenomena, he coolant velocity is in the range 3-9 m.s-1 and the resulting heat flux is in the range 7-13 MW.m-2. From video images (single frames were taken with a light exposure of 1 μs) the following general behaviour of vapour bubbles was observed: when the rate of bubble generation is increasing, with bubbles growing in the superheated layer close to the heating wall, their coalescence produces a sort of elongated bubble called a vapour blanket. One of the main features of the vapour blanket is that it is rooted to the nucleation site on the heated surface. Bubble dimensions, as well as those of the hot spots, are given as a function of thermal-hydraulic tested conditions. (authors)

  7. Confinement by Carbon Nanotubes Drastically Alters the Boiling and Critical Behavior of Water Droplets

    OpenAIRE

    Chaban, Vitaly V.; Prezhdo, Victor V.; Prezhdo, Oleg V.

    2012-01-01

    Vapor pressure grows rapidly above the boiling temperature, and past the critical point liquid droplets disintegrate. Our atomistic simulations show that this sequence of events is reversed inside carbon nanotubes (CNT). Droplets disintegrate first and at low temperature, while pressure remains small. The droplet disintegration temperature is independent of the CNT diameter. In contrast, depending on CNT diameter, a temperature that is much higher than the bulk boiling temperature is required...

  8. Burnout in subcooled flow boiling of water. A visual experimental study

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G.P.; Mariani, A.; Zummo, G. [ENEA, Engineering Div., National Institute of Thermal Fluid-Dynamics, Rome (Italy); Cumo, M. [University of Rome la Sapienza, Rome (Italy)

    2000-12-01

    The objective of the present work is to perform a photographic study of the burnout in highly subcooled flow boiling, in order to provide a qualitative description of the flow pattern under different conditions of boiling regime: ONB (onset of nucleate boiling), subcooled flow boiling and thermal crisis. In particular, the flow visualisation is focused on the phenomena occurring on the heated wall during the thermal crisis up to the physical burnout of the heater. Vapour bubble parameters are measured from flow images recorded, while the wall temperature is measured with an indirect method, by recording the heater elongation during all flow regimes studied. The combination of bubble parameters and wall temperature measurements as well as direct observations of the flow pattern, for all flow regimes, are collected in graphs which provide a useful global point of view of boiling phenomena, especially during boiling crisis. Under these conditions, a detailed analysis of the mechanisms leading to the critical heat flux is reported, and the so called events sequence, from thermal crisis occurrence up to heater burnout, is illustrated. (authors)

  9. Burnout in subcooled flow boiling of water. A visual experimental study

    International Nuclear Information System (INIS)

    The objective of the present work is to perform a photographic study of the burnout in highly subcooled flow boiling, in order to provide a qualitative description of the flow pattern under different conditions of boiling regime: ONB (onset of nucleate boiling), subcooled flow boiling and thermal crisis. In particular, the flow visualisation is focused on the phenomena occurring on the heated wall during the thermal crisis up to the physical burnout of the heater. Vapour bubble parameters are measured from flow images recorded, while the wall temperature is measured with an indirect method, by recording the heater elongation during all flow regimes studied. The combination of bubble parameters and wall temperature measurements as well as direct observations of the flow pattern, for all flow regimes, are collected in graphs which provide a useful global point of view of boiling phenomena, especially during boiling crisis. Under these conditions, a detailed analysis of the mechanisms leading to the critical heat flux is reported, and the so called events sequence, from thermal crisis occurrence up to heater burnout, is illustrated. (authors)

  10. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  11. Mixed oxide fuel for water cooled reactors

    International Nuclear Information System (INIS)

    The problems connected with introduction of plutonium extracted from spent fuels of operating NPPs into water cooled reactor fuel cycle are considered. The trends in formation of the World market of mixed fuel are illustrated taking as examples Great Britain and Japan

  12. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability

  13. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  14. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  15. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  16. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  17. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  18. Absence of genotoxic activity from milk and water boiled in microwave oven in somatic cells from Drosophila melanogaster

    International Nuclear Information System (INIS)

    This paper reports an experiment for evaluation of the possible genotoxic effects of food prepared in a microwave oven, through the mutation test and somatic recombination, in wings of Drosophila melanogaster. Two crossing have been performed: a standard cross-ST and a high bioactivation cross - HB resulting in marked trans -heterozygote descendents (MH) and balanced heterozygotes (BH). The 72 hours larvas were fed with water and milk boiled both in the microwave oven and in the traditional way. The MH individual wings were analyzed, where the spots can be induced either by mutation or mitotic recombination. The experiment presented negative results related to the genotoxic effects of the water and milk boiled using the microwave oven, in MH descendents of both crossing. Therefore, under these experimental conditions, genotoxic activity were not presented by milk and water boiled in the microwave oven. However, an extensive study using different techniques is necessary to investigate the action of the food prepared in the microwave oven on the genetic material

  19. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  20. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

  1. Behavior of water reactor fuel rod

    International Nuclear Information System (INIS)

    This paper reviewed the fuels used widely in forms of (1) Zircaloy-sheathed UO2 fuel in light water-commercial power reactor, (2) Zircaloy-sheathed PuO2-UO2 fuel in plutonium-thermal reactor and advanced reactor (ATR), (3) aluminide and silicide fuel in Material Testing Reactors. From fundamental view points, physical/chemical properties and irradiation behaviors of both fuels and zircaloy claddings are briefly reviewed in chapters 1 and 2. Change of the fuel rod physical parameters with progress of burn-up are summed up in chapter 3. Some fuel troubles and failures encountered in past usage of worldwide LWR fuels are introduced with counterplans taken. In the last session of this chapter, recent results of R and D works have been carried out by fuel vendors are reviewed. Especially, in-core behaviors of PCI-remedy fuels developed to use for high burn-up extension and for load-follow operation are highlighted. Reactor accidents occurred through past forty years are surveyed and reviewed. Fuel behaviors during the reactivity initiated accident (RIA), the power-coolant mismatch (PCM), and the loss-of-coolant accident (LOCA) are taken into this review by using disclosed literatures. Safety criteria being used in Japanese licensing authorities are introduced relating to the fuel design limit. (author)

  2. Advanced light water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Giedraityte, Zivile [Helsinki University of Technology, Otaranta 8D-84, 02150 Espoo (Finland)

    2008-07-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  3. Boiling and burnout phenomena under transient heat input, 1

    International Nuclear Information System (INIS)

    In order to simulate the thermo-hydrodynamic conditions at reactor power excursions, a test piece was placed in a forced convective channel and heated with exponential power inputs. The boiling heat transfer and the burnout heat flux under the transient heat input were measured, and pressure and water temperature changes in the test section were recorded at the same time. Following experimental results were obtained; (1) Transient boiling heat transfer characteristics at high heat flux stayed on the stationary nucleate boiling curve of each flow condition, or extrapolated line of the curves. (2) Transient burnout heat flux increased remarkably with decreasing heating-time-constant, when the flow rate was lower and the subcooling was higher. (3) Transient burnout phenomena were expressed with the relation of (q sub(max) - q sub(sBO)) tau = constant at several flow conditions. This relation was derived from the stationary burnout mechanism of pool boiling. (auth.)

  4. Numerical thermal analysis of water's boiling heat transfer based on a turbulent jet impingement on heated surface

    Science.gov (United States)

    Toghraie, D.

    2016-10-01

    In this study, a numerical method for simulation of flow boiling through subcooled jet on a hot surface with 800 °C has been presented. Volume fraction (VOF) has been used to simulate boiling heat transfer and investigation of the quench phenomena through fluid jet on a hot horizontal surface. Simulation has been done in a fixed Tsub=55 °C, Re=5000 to Re=50,000 and also in different Tsub =Tsat -Tf between 10 °C and 95 °C. The effect of fluid jet velocity and subcooled temperature on the rewetting temperature, wet zone propagation, cooling rate and maximum heat flux has been investigated. The results of this study show that by increasing the velocity of fluid jet of water, convective heat transfer coefficient at stagnation point increases. More ever, by decreasing the temperature of the fluid jet, convective heat transfer coefficient increases.

  5. A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Sang Jun; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling. 16 refs., 6 figs., 1 tab. (Author)

  6. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    International Nuclear Information System (INIS)

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins

  7. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

    Directory of Open Access Journals (Sweden)

    Diego Mandelli

    2015-01-01

    Full Text Available In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

  8. Assessment of corrosion and fatigue damage to light water reactor metal containments

    International Nuclear Information System (INIS)

    This paper presents a generic procedure for estimating aging damage, evaluating structural integrity, and identifying mitigation activities for safe operation of boiling water reactor (BWR) Mark I metal containments and ice-condenser type pressurized water reactor (PWR) cylindrical metal containments. The mechanisms of concern that can cause aging damage to these two types of containments are corrosion and fatigue. Assessment of fatigue damage to bellows is also described. Assessment of corrosion and fatigue damage described in this paper include: containment design features that are relevant to aging assessment, several corrosion and fatigue mechanisms, inspection of corrosion and fatigue damage, and mitigation of damage caused by these mechanisms. In addition, synergistic interaction between corrosion and fatigue is considered. Possible actions for mitigating aging include enhanced inspection methods, maintenance activities based on operating experience, and supplementary surveillance programs. Field experience related to aging of metal containments is reviewed. Finally, conclusions and recommendations are presented

  9. Physicochemical and Sensory Properties of Boiled Prosopis africana Seed Endosperm Macerated in Various Ethanol-water Mixtures

    Directory of Open Access Journals (Sweden)

    James E. Obiegbuna

    2013-09-01

    Full Text Available The processing of boiled Prosopis africana endosperm for better utilization using ethanol-water mixtures was explored. Prosopis africana seeds were boiled for 5 h to softness and the endosperm fraction separated from the kernel (cotyledon and the hull. The endosperm was divided into five equal parts which were individually macerated in absolute (Abs ethanol, 80, 60 and 40% ethanol in water prior to sun-drying (32±2°C, 3 days. The fifth sample, which served as control, was left untreated with ethanol. The samples were ground using a hand milling machine and analyzed for the proximate composition, water and oil absorption capacities, foaming capacity and foam stability, bulk density, emulsion activity and stability, colour preference, texture preference and overall acceptability. The results revealed that treatment of the endosperm significantly affected the moisture, protein, fat, ash and carbohydrate contents; water and oil absorption capacities, foaming capacity and foam stability; and the sensory properties. The moisture and protein contents, oil absorption capacity, foam stability, appearance, texture and overall acceptability of endosperm treated with 40% ethanol in water differed significantly (p<0.05 from that treated with absolute ethanol. There was also significant (p<0.05 differences in moisture, protein and carbohydrate contents, oil absorption capacity and foam stability of the 40% ethanol in water treated endosperm and the control. Slightly above 40% ethanol in water (50-60% should be used to macerate Prosopis africana endosperm to reduce the cost of using absolute ethanol.

  10. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  11. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  12. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [and others

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  13. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [Argonne National Lab., IL (United States)] [and others

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  14. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289 degree C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  15. Microflora of nuclear research reactor pool water

    International Nuclear Information System (INIS)

    The circulation of pool water through the nuclear reactor core produces a bactericidal effect on the microflora due to the influence of various kinds of radiation. The microbe contents return to their initial level in 2 to 4 months after the circulation has stopped. The microflora comprises mainly cocci in large numbers, G-positive rods and fungi, and lower amounts of G-negative rods as compared with the water with which the reactor pool was initially filled. Increased amounts are present of radiation-resistant forms exhibiting intense production of catalase and nuclease. Supposedly, the presence of these enzymes is in some way beneficial to the microbes in their survival in the high-radiation zones. (author). 1 fig., 2 tabs., 12 refs

  16. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  17. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  18. Forced convective boiling heat transfer of water in vertical rectangular narrow channel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chong, E-mail: chenchong_2012@163.com; Gao, Pu-zhen, E-mail: gaopuzhen@hrbeu.edu.cn; Tan, Si-chao; Chen, Han-ying; Chen, Xian-bing

    2015-09-15

    Highlights: • Chen correlation cannot well predict the coefficient of rectangular channel. • Kim and Mudawar correlation is the best one among the Chen type correlations. • Lazarek and Black correlation predicted 7.0% of data within the ±30% error band. • The new correlation can well predict the coefficient with a small MAE of 14.4%. - Abstract: In order to research the characteristics of boiling flows in a vertical rectangular narrow channel, a series of convective boiling heat transfer experiments are performed. The test section is made of stainless steel with an inner diameter of 2 × 40 mm and heated length of 1100 mm. The 3194 experimental data points are obtained for a heat flux range of 10–700 kW/m{sup 2}, a mass flux range of 200–2400 kg/m{sup 2} s, a system pressure range of 0.1–2.5 MPa, and a quality range of 0–0.8. Eighteen prediction models are used to predict the flow boiling heat transfer coefficient of the rectangular narrow channel and the predicted value is compared against the database including 3194 data points, the results show that Chen type correlations and Lazarek and Black type correlations are not suitable for the rectangular channel very much. The Kim and Mudawar correlation is the best one among the 18 models. A new correlation is developed based on the superposition concept of nucleate boiling and convective boiling. the new correlation is shown to provide a good prediction against the database, evidenced by an overall MAE of 14.4%, with 95.2% and 98.6% of the data falling within ±30% and ±35% error bands, respectively.

  19. Forced convective boiling heat transfer of water in vertical rectangular narrow channel

    International Nuclear Information System (INIS)

    Highlights: • Chen correlation cannot well predict the coefficient of rectangular channel. • Kim and Mudawar correlation is the best one among the Chen type correlations. • Lazarek and Black correlation predicted 7.0% of data within the ±30% error band. • The new correlation can well predict the coefficient with a small MAE of 14.4%. - Abstract: In order to research the characteristics of boiling flows in a vertical rectangular narrow channel, a series of convective boiling heat transfer experiments are performed. The test section is made of stainless steel with an inner diameter of 2 × 40 mm and heated length of 1100 mm. The 3194 experimental data points are obtained for a heat flux range of 10–700 kW/m2, a mass flux range of 200–2400 kg/m2 s, a system pressure range of 0.1–2.5 MPa, and a quality range of 0–0.8. Eighteen prediction models are used to predict the flow boiling heat transfer coefficient of the rectangular narrow channel and the predicted value is compared against the database including 3194 data points, the results show that Chen type correlations and Lazarek and Black type correlations are not suitable for the rectangular channel very much. The Kim and Mudawar correlation is the best one among the 18 models. A new correlation is developed based on the superposition concept of nucleate boiling and convective boiling. the new correlation is shown to provide a good prediction against the database, evidenced by an overall MAE of 14.4%, with 95.2% and 98.6% of the data falling within ±30% and ±35% error bands, respectively

  20. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  1. Experimental study of mass boiling in a porous medium model

    International Nuclear Information System (INIS)

    This manuscript presents a pore-scale experimental study of convective boiling heat transfer in a two-dimensional porous medium. The purpose is to deepen the understanding of thermohydraulics of porous media saturated with multiple fluid phases, in order to enhance management of severe accidents in nuclear reactors. Indeed, following a long-lasting failure in the cooling system of a pressurized water reactor (PWR) or a boiling water reactor (BWR) and despite the lowering of the control rods that stops the fission reaction, residual power due to radioactive decay keeps heating up the core. This induces water evaporation, which leads to the drying and degradation of the fuel rods. The resulting hot debris bed, comparable to a porous heat-generating medium, can be cooled down by reflooding, provided a water source is available. This process involves intense boiling mechanisms that must be modelled properly. The experimental study of boiling in porous media presented in this thesis focuses on the influence of different pore-scale boiling regimes on local heat transfer. The experimental setup is a model porous medium made of a bundle of heating cylinders randomly placed between two ceramic plates, one of which is transparent. Each cylinder is a resistance temperature detector (RTD) used to give temperature measurements as well as heat generation. Thermal measurements and high-speed image acquisition allow the effective heat exchanges to be characterized according to the observed local boiling regimes. This provides precious indications precious indications for the type of correlations used in the non-equilibrium macroscopic model used to model reflooding process. (author)

  2. Burnup determination of water reactor fuel

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency in consultation with the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The meeting was hosted by the Commission of the European Communities, at the Transuranium Research Laboratory, Joint Research Centre Karlsruhe, in the Federal Republic of Germany. This subject was dealt with for the first time by the IAEA. It was found to correspond adequately to this type of Specialist Meeting and to be suitable at a moment when the extension of burnup constitutes a major technical and economical issue in fuel technology. It was stressed that analysis of highly burnt fuels, mixed oxides and burnable absorber bearing fuels required extension of the experimental data base, to comply with the increasing demand for an improved fuel management, including better qualification of reactor physics codes. Twenty-seven participants from eleven countries plus two international organizations attended the Meeting. Twelve papers were given during three technical sessions, followed by a panel discussion which allowed to formulate the conclusions of the meeting and recommendations to the Agency. In addition, participants were invited to give an outline of their national programmes, related to Burnup Determination of Water Reactor Fuel. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

  3. Good practices in heavy water reactor operation

    International Nuclear Information System (INIS)

    The value and importance of organizations in the nuclear industry engaged in the collection and analysis of operating experience and best practices has been clearly identified in various IAEA publications and exercises. Both facility safety and operational efficiency can benefit from such information sharing. Such sharing also benefits organizations engaged in the development of new nuclear power plants, as it provides information to assist in optimizing designs to deliver improved safety and power generation performance. In cooperation with Atomic Energy of Canada, Ltd, the IAEA organized the workshop on best practices in Heavy Water Reactor Operation in Toronto, Canada from 16 to 19 September 2008, to assist interested Member States in sharing best practices and to provide a forum for the exchange of information among participating nuclear professionals. This workshop was organized under Technical Cooperation Project INT/4/141, on Status and Prospects of Development for and Applications of Innovative Reactor Concepts for Developing Countries. The workshop participants were experts actively engaged in various aspects of heavy water reactor operation. Participants presented information on activities and practices deemed by them to be best practices in a particular area for consideration by the workshop participants. Presentations by the participants covered a broad range of operational practices, including regulatory aspects, the reduction of occupational dose, performance improvements, and reducing operating and maintenance costs. This publication summarizes the material presented at the workshop, and includes session summaries prepared by the chair of each session and papers submitted by the presenters

  4. The decommissioning of the water boiler reactor

    International Nuclear Information System (INIS)

    Following completion of service, the Water Boiler Reactor (WBR) has been decommissioned by the Institute of Nuclear Energy Research (INER) under the Atomic Energy Council's (AEC) regulation. The WBR is a light water moderated and graphite reflected research reactor with peak thermal power of 100 kW. The unique feature of the WBR is that it is fueled with uranyl sulfate (UO2SO4) which is in liquid form. Since there is another research reactor owned by I7NER of megawatt scale in the planning stages for decommissioning, the WBR project was conducted with great care to accumulate experience. Extensive planning by INER and step-by-step regulative activities by AEC were followed regardless of the structural simplicity of the WBR. Valuable information was gathered in the task and will be useful for preparing future decommissioning needs. The major work in the WBR decommissioning project was finished within six months and the accumulated dose received during the work was 1 9.63mSv. (author)

  5. Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

    International Nuclear Information System (INIS)

    Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

  6. Controlling hydrogen behavior in light water reactors

    International Nuclear Information System (INIS)

    In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

  7. Mathematical modelling of water radiolysis kinetics under reactor conditions

    International Nuclear Information System (INIS)

    Experimental data on coolant radiolysis (RBMK-1000 reactor) were used to construct mathematical model of water radiolysis kinetics under reactor conditions. Good agreement of calculation results with the experiment is noted

  8. A simultaneous observation of bubble growth and microlayer behavior for an isolated boiling regime of saturated water

    International Nuclear Information System (INIS)

    The bubble growth rate and microlayer behavior were simultaneously visualized for an isolated boiling regime of saturated water. The increase rate of the bubble volume dropped sharply when the microlayer was totally depleted. However, the contribution of the superheated liquid layer evaporation to the bubble volume increase was comparable to or even higher than that of the microlayer evaporation during the time when the microlayer evaporation was active. The microlayer under the coalesced bubble was much thicker than that under single isolated bubble. (author)

  9. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S

    2009-03-15

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  10. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    International Nuclear Information System (INIS)

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  11. A modern control room for Indian Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a next generation nuclear power plant being developed by Bhabha Atomic Research Centre, India. AHWR is a vertical, pressure tube type, heavy-water-moderated, boiling light-water-cooled, innovative reactor, relying on natural circulation for core cooling in all operating and accident conditions. In addition, it incorporates various passive systems for decay heat removal, containment cooling and isolation. In addition to the many passive safety features, AHWR has state of the art I and C architecture based on extensive use of computers and networking. In tune with the many advanced features of the reactor, a centralized modern control room has been conceived for operation and monitoring of the plant. The I and C architecture enables the implementation of a fully computerised operator friendly control room with soft Human Machine Interfaces (HMI). While doing so, safety has been given due consideration. The control and monitoring of AHWR systems are carried out from the fully computer-based operator interfaces, except safety systems, for which only monitoring is provided from soft HMI. The control of the safety systems is performed from dedicated hardwired safety system panels. Soft HMI reduces the number of individual control devices and improves their reliability. The paper briefly describes the I and C architecture adopted for the AHWR plant along with the interfaces to the main and backup control rooms. There are many issues involved while introducing soft HMI based operator interfaces for Nuclear Power Plants (NPP) compared to the conventional plants. Besides discussing the implementation issues, the paper elaborates the design considerations that have undergone in the design of various components in the main control room especially operator workstations, shift supervisor console, safety system panels and large display panels. Mainly task based displays have been adopted for the routine operator interactions of the plant

  12. Nonlinear stability analysis of a reduced order model of nuclear reactors: A parametric study relevant to the advanced heavy water reactor

    International Nuclear Information System (INIS)

    Research highlights: → We model power oscillations in boiling water reactors using a lumped parameter model. → The nature and amplitudes of oscillations is obtained using a nonlinear analysis. → The method of multiple scales has been used for the analytical treatment. → Fuel temperature coefficient of reactivity determines the nature of oscillations. → The presented systematic method of analysis useful for reduced order reactor models. - Abstract: In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy water reactor (AHWR). The neutronics of the AHWR is modeled using point reactor kinetic equations while a one-node lumped parameter model is assumed both for the fuel and the coolant for modeling the thermal-hydraulics. Nonlinearities in the heat transfer process are ignored and attention is focused on the nonlinearity introduced by the reactivity feedback. It is found that the steady-state operation of the AHWR mathematical model looses stability via. a Hopf bifurcation resulting in power oscillations as some typical bifurcation parameter like the void coefficient of reactivity is varied. The bifurcation is found to be subcritical for the parameter values corresponding to the AHWR. However, with a decrease in the fuel temperature coefficient of reactivity the bifurcation turns to supercritical implying global stability of the steady state operation in the linear stability regime. Moreover slight intrusion into the instability regime results in small-amplitude limit cycles leaving the possibility of retracting back to stable operation.

  13. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  14. Is light water reactor technology sustainable?

    International Nuclear Information System (INIS)

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  15. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

    2001-07-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  16. Environmentally assisted cracking in light water reactors. Semiannual report, April--September 1991: Volume 13

    Energy Technology Data Exchange (ETDEWEB)

    Kassner, T F; Ruther, W E; Chung, H M; Hicks, P D; Hins, A G; Park, J Y; Soppet, W K; Shack, W J [Argonne National Lab., IL (United States)

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with {approx} 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  17. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

  18. Development of a thermal–hydraulic analysis code for the Pebble Bed Water-cooled Reactor

    International Nuclear Information System (INIS)

    Highlights: ► Main design features of the PBWR were put forward. ► Thermal–hydraullics analysis code for the PBWR was developed and verified. ► Key thermal–hydraullics parameters were calculated in normal operation. ► The PBWR has a great pressure loss but an excellent heat transfer characteristic. ► Maximum fuel temperature and MDNBR are in conformity with safety criterion. - Abstract: The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly. Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper.

  19. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289 degrees C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials

  20. NAIADQ, a computer program for calculating reactivity transients in low power experimental water reactors

    International Nuclear Information System (INIS)

    The computer code NAIADQ is designed to simulate the course and consequences of non-destructive reactivity accidents in low power, experimental, water-cooled reactor cores fuelled with metal plate elements. It is a coupled neutron kinetics-hydrodynamics-heat transfer code which uses point kinetics and one-dimensional thermohydraulic equations. Nucleate boiling, which occurs at the fuel surface during transients, is modelled by the growth of a superheated layer of water in which vapour is generated at a non-equilibrium rate. It is assumed that this vapour is formed at its saturation temperature and that it mixes homogeneously with the water in this layer. The code is written in FORTRAN IV and has been programmed to run as a catalogued procedure on an IBM operating system such as MVT or MVS, with facility for the inclusion of user routines

  1. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  2. Behavior of tritium in heavy water reactors

    International Nuclear Information System (INIS)

    In the ATR Fugen power station, the radiation control regarding the tritium in heavy water has been carried out since the heavy water was filled in the system of the reactor in November, 1977. At first, the concentration of tritium in heavy water was about 60 μCi/cc, but in November, 1981, it increased to about 1.3 mCi/cc, and the saturation concentration after 30 years is estimated to become about 17 mCi/cc. In this report, on the transfer of tritium to the work environment and general environment, its barrier, recovery, measurement and the protection against it, the experience in the Fugen power station is described. The heavy water system was constructed as the perfectly closed circuit by welding stainless steel, and a canned heavy water circulating pump has been used. The leak of heavy water in the steady operation is negligible, but attention must be paid to the transfer of tritium to the environment when the system is disassembled for the regular inspection. The measurement of tritium for individual exposure control, environment and released radioactivity, the tritium-removing equipment and protective suits, and the release of tritium to general environment are reported. (Kako, I.)

  3. Burnout experiment in subcooled forced-convection boiling of water for beam dumps of a high power neutral beam injector

    International Nuclear Information System (INIS)

    Experimental studies were made on burnout heat flux in highly subcooled forced-convection boiling of water for the design of beam dumps of a high power neutral beam injector for Japan Atomic Energy Research Institute Tokamak-60. These dumps are composed of many circular tubes with two longitudinal fins. The tube was irradiated with nonuniformly distributed hydrogen ion beams of 120 to 200 kW for as long as 10 s. The coolant water was circulated at flow velocities of 3 to 7.5 m/s at exit pressures of 0.4 to 0.9 MPa. The burnout and film-boiling data were obtained at local heat fluxes of 8 to 15 MW/m2. These values were as high as 2.5 times larger than those for the circumferentially uniform heat flux case with the same parameters. These data showed insensitivity to local subcooling as well as to pressure, and simple burnout correlations were derived. From these results, the beam dumps have been designed to receive energetic beam fluxes of as high as 5 MW/m2 with a margin of a factor of 2 for burnout

  4. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  5. Evaluation of silica behavior for reducing the precoating frequency of the reactor water cleanup system of the Hamaoka NPS

    International Nuclear Information System (INIS)

    When the silica concentration in reactor water exceeds the reference value, the ion-exchange resin powder used in the reactor water cleanup system (CUW) is replaced with new resin powder. This is referred to as the 'precoating of CUW'. Precoating of CUW generates radioactive waste; therefore, a higher frequency of CUW precoating increases the waste disposal cost. In the Advanced Boiling-Water Reactor (ABWR) of Hamaoka Unit 5, the frequency of CUW precoating has been higher than that of any other plants as a result of the high silica concentration in the reactor water; therefore, the behavior of silica in the reactor primary water circuit was examined in order to reduce the frequency of CUW precoating. A calculation model was developed for the silica behavior in the primary water (Silica Behavior Code) and the mass balances of silica in Hamaoka Units 3 and 4 (BWR5), and Unit 5 was then analyzed applying this code. A comparison of these mass balances shows two results for the difference between BWR5 and the ABWR. First, the amount of silica removed from the condensate demineralizer (CD) in the ABWR is less than that in BWR5, because silica, which is transported from the reactor water into the main steam, bypasses the CD and returns to the reactor water directly due to the high pressure heater drain line specific to the ABWR. Second, the amount of silica generated during plant operation in the ABWR is greater than that in BWR5 due to the generation of silica in the high pressure heater drain line. From the above results, it is concluded that the high pressure heater drain line, which is specific to the ABWR, is the cause of the high silica concentration in reactor water in Hamaoka Unit 5. (author)

  6. Environmentally assisted cracking in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  7. In-situ Observation of Boiling Dynamics on Fuel Cladding Surface in Non-pressurized Water Using Acoustic Emission Method

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Kaige; Baek, Seung Heon; Shim, Hee-Sang; Hur, Do Haeng; Lee, Deok Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In the PWR primary coolant system, a phenomenon of axial offset anomaly (AOA) can be caused due to accumulated boron hide out in porous CRUD deposition on the fuel cladding surface. Up to now, the CRUD deposition has been well known to be driven by subcooled nucleate boiling (SNB) on the cladding surface based on large scale experimental work. Therefore, monitoring and evaluation of the SNB-phenomenon is an important approach to study the CRUD deposition. Many attempts have been made to study the SNB and CRUD deposition using thermal hydraulic or model calculation. However, a comprehensive understanding of the SNB during CRUD deposition is still far from being realized. Acoustic emission (AE) technique, as an in-situ nondestructive evaluation (NDE) method, has been widely used to monitor the boiling activity in containers and pipes. Accordingly, this work aimed to investigate the exact AE characteristics of SNB-phenomenon on the fuel cladding surface at atmospheric pressure, with the purpose of providing an experimental groundwork for the AE investigation on SNB in high-temperature pressurized coolant system. In this study, we conducted an in-situ experimental observation of the bubble dynamic of SNB in non-pressurized water at atmospheric pressure using AE method. The AE of heater noise was confirmed to cluster between 8 and 26 khz. Three AE groups were detected during the boiling process in the Snob zones. AE group 1 and 3 seemed to be the results of bubble growth and collapse, while bubble departure from the cladding surface was reasonably associated with an isolated AE group 2.

  8. An assessment of in-tube flow boiling correlations for ammonia-water mixtures and their influence on heat exchanger size

    DEFF Research Database (Denmark)

    Kærn, Martin Ryhl; Modi, Anish; Jensen, Jonas Kjær;

    2016-01-01

    on the required heat exchanger size (surface area)is investigated during numerical design. For this purpose, two case studies related to the use of the Kalina cycle are considered: a flue gas based heat recovery boiler for acombined cycle power plant and a hot oil based boiler for a solar thermal power plant......Heat transfer correlations for pool and flow boiling are indispensable for boiler design. The correlations for predicting in-tube flow boiling heat transfer ofammonia-water mixtures are not well established in the open literature and there is a lack of experimental measurements for the full range...... of composition, vapor qualities, fluid conditions, etc. This paper presents a comparison of several flow boiling heat transfer prediction methods (correlations) for ammonia-water mixtures. Firstly, these methods are reviewed and compared at various fluid conditions. The methods include: (1) the ammonia-water...

  9. Screening reactor steam/water piping systems for water hammer

    International Nuclear Information System (INIS)

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

  10. Boiling of subcooled water in forced convection; Ebullition locale de l'eau en convection forcee

    Energy Technology Data Exchange (ETDEWEB)

    Ricque, R.; Siboul, R. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1970-07-01

    As a part of a research about water cooled high magnetic field coils, an experimental study of heat transfer and pressure drop is made with the following conditions: local boiling in tubes of small diameters (2 and 4 mm), high heat fluxes (about 1000 W/cm{sup 2}), high coolant velocities (up to 25 meters/s), low outlet absolute pressures (below a few atmospheres). Wall temperatures are determined with a good accuracy, because very thin tubes are used and heat losses are prevented. Two regimes of boiling are observed: the establishment regime and the established boiling regime and the inception of each regime is correlated. Important delays on boiling inception are also observed. The pressure drop is measured; provided the axial temperature distribution of the fluid and the axial distributions of the wall temperatures, in other words the axial distribution of the heat transfer coefficients under boiling and non boiling conditions, at the same heat flux or the same wall temperatures, are taken in account, then total pressure drop can be correlated, but probably under certain limits of void fraction only. Using the same parameters, it seems possible to correlate the experimental values on critical heat flux obtained previously, which show very important effect of length and hydraulic diameter of the test sections. (authors) [French] Dans le cadre d'une etude sur le refroidissement par l'eau des bobines electromagnetiques a champ intense, on etudie experimentalement l'echange thermique et la perte de pression avec ebullition locale a la paroi dans des tubes de petit diametre (2 et 4 mm), a flux thermique eleve (environ 1000 W/cm{sup 2}), pour des vitesses de circulation elevees (jusqu'a 25 m/s) et des pressions basses (quelques atmospheres). La paroi des tubes etant tres mince et les fuites thermiques etant annulees, les temperatures de paroi sont determinees de facon assez precise. On distingue deux phases dans l'ebullition locale; la phase d

  11. Optimum structural properties for an anode current collector used in a polymer electrolyte membrane water electrolyzer operated at the boiling point of water

    Science.gov (United States)

    Li, Hua; Fujigaya, Tsuyohiko; Nakajima, Hironori; Inada, Akiko; Ito, Kohei

    2016-11-01

    This study attempts to optimize the properties of the anode current collector of a polymer electrolyte membrane water electrolyzer at high temperatures, particularly at the boiling point of water. Different titanium meshes (4 commercial ones and 4 modified ones) with various properties are experimentally examined by operating a cell with each mesh under different conditions. The average pore diameter, thickness, and contact angle of the anode current collector are controlled in the ranges of 10-35 μm, 0.2-0.3 mm, and 0-120°, respectively. These results showed that increasing the temperature from the conventional temperature of 80 °C to the boiling point could reduce both the open circuit voltage and the overvoltages to a large extent without notable dehydration of the membrane. These results also showed that decreasing the contact angle and the thickness suppresses the electrolysis overvoltage largely by decreasing the concentration overvoltage. The effect of the average pore diameter was not evident until the temperature reached the boiling point. Using operating conditions of 100 °C and 2 A/cm2, the electrolysis voltage is minimized to 1.69 V with a hydrophilic titanium mesh with an average pore diameter of 21 μm and a thickness of 0.2 mm.

  12. Development of neutron radiography facility for boiling two-phase flow experiment in Kyoto University Research Reactor

    Science.gov (United States)

    Saito, Y.; Sekimoto, S.; Hino, M.; Kawabata, Y.

    2011-09-01

    To visualize boiling two-phase flow at high heat flux by using neutron radiography, a new neutron radiography facility was developed in the B-4 beam hole of KUR. The B-4 beam hole is equipped with a supermirror neutron guide tube with a characteristic wavelength of 1.2 Å, whose geometrical parameters of the guide tube are: 11.7 m total length and 10 mm wide ×74 mm high beam cross-section. The total neutron flux obtained from the KUR supermirror guide tube is about 5×10 7 n/cm 2 s with a nominal thermal output of 5 MW of KUR, which is about 100 times what is obtainable with the conventional KUR neutron radiography facility (E-2 beam hole). In this study a new imaging device, an electric power supply (1200 A, 20 V), and a thermal hydraulic loop were installed. The neutron source, the beam tube, and the radiography rooms are described in detail and the preliminary images obtained at the developed facility are shown.

  13. Relations between boiling water test, standard germination test and field emergence of leek (Allium porrum L.) and onion (Allium cepa L.) seeds

    OpenAIRE

    Ismail Guvenc; Haluk C. Kaymak; Sibel Duman

    2012-01-01

    The aim of this study was to determine relations occurring between boiling water test, standard germination test and field emergence of leek (Allium porrum L.) and onion (Allium cepa L.) seeds. In this study, seeds of six lots ('Kalem', 'Ala', 'Ínegöl-A, B, C and D') from three cultivars of leek and seven onion cultivars ('Early Texas Grano' (ETG), 'Panku', 'Storm', 'Banko', 'Aki', 'Kisagün' and 'Banka') seeds were used as plant material and their viability was evaluated in boiling water test...

  14. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  15. Analysis of transients in advanced heavy water reactor using lumped parameter models

    Energy Technology Data Exchange (ETDEWEB)

    Manmohan Pandey; Venkata Ramana Eaga; Sankar Sastry, P. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India); Gupta, S.K.; Lele, H.G.; Chatterjee, B. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    2005-07-01

    Full text of publication follows: Analysis of transients occurring in nuclear power plants, arising from the complex interplay between core neutronics and thermal-hydraulics, is important for their operation and safety. Numerical simulations of such transients can be carried out extensively at very low computational cost by using lumped parameter mathematical models. The Advanced Heavy Water Reactor (AHWR), being developed in India, is a vertical pressure tube type reactor cooled by boiling light water under natural circulation, using thorium as fuel and heavy water as moderator. In the present work, nonlinear and linear lumped parameter dynamic models for AHWR have been developed and validated with a distributed parameter model. The nonlinear lumped model is based on point reactor kinetics equations and one-dimensional homogeneous equilibrium model of two-phase flow. The distributed model is built with RELAP5/MOD3.2 code. Various types of transients have been simulated numerically, using the lumped model as well as RELAP5. The results have been compared and parameters tuned to make the lumped model match the distributed model (RELAP5) in terms of steady state as well as dynamic behaviour. The linear model has been derived by linearizing the nonlinear model for small perturbations about the steady state. Numerical simulations of transients using the linear model have been compared with results obtained from the nonlinear model. Thus, the range of validity of the linear model has been determined. Stability characteristics of AHWR have been investigated using the lumped parameter models. (authors)

  16. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  17. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  18. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Main report, draft report for comment. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

    1994-09-01

    On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB alternatives, which now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste. Costs for labor, transport, and disposal activities are given in 1993 dollars. Sensitivities of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances are also examined.

  19. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Appendices, draft report for comment. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

    1994-09-01

    On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s WNP-2, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives, which now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste. Costs for labor, materials, transport, and disposal activities are given in 1993 dollars. Sensitivities of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances are also examined.

  20. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Appendices, draft report for comment. Volume 2

    International Nuclear Information System (INIS)

    On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System's WNP-2, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives, which now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste. Costs for labor, materials, transport, and disposal activities are given in 1993 dollars. Sensitivities of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances are also examined