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Sample records for boiling water reactors

  1. Advanced boiling water reactor

    International Nuclear Information System (INIS)

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

  2. High Pressure Boiling Water Reactor

    International Nuclear Information System (INIS)

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

  3. 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

  4. 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...

  5. 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.)

  6. Serious accidents on boiling water reactors (BWR)

    International Nuclear Information System (INIS)

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

  7. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

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

  8. Fuel recycling in boiling water reactors

    International Nuclear Information System (INIS)

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

  9. 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)

  10. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

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

  11. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

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

  12. 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)

  13. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

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

  14. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

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

  15. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

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

  16. Detonating gas in boiling water reactors

    International Nuclear Information System (INIS)

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

  17. 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.

  18. Design certification program of the simplified boiling water reactor

    International Nuclear Information System (INIS)

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

  19. 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

  20. 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.)

  1. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

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

  2. 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

  3. 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)...

  4. 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...

  5. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

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

  6. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

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

  7. 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)....

  8. 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...

  9. 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)

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

    International Nuclear Information System (INIS)

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

  11. 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)

  12. Two compartment water rod for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matzner, B.; Wolters, B.A.

    1993-07-27

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

  13. Two compartment water rod for boiling water reactors

    International Nuclear Information System (INIS)

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

  14. 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)

  15. 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.

  16. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Boiling water reactor operator training and qualification in Japan

    International Nuclear Information System (INIS)

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

  20. Digital control application for the advanced boiling water reactor

    International Nuclear Information System (INIS)

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

  1. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. 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)

  6. 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

  7. 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

  8. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

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

  9. Stability monitoring of the Dodewaard boiling-water reactor

    International Nuclear Information System (INIS)

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

  10. 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)

  11. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

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

  12. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

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

  13. 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

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

    International Nuclear Information System (INIS)

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

  15. 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)

  16. 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

  17. 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.

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

    International Nuclear Information System (INIS)

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

  19. 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

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

    Science.gov (United States)

    Lu, Shanlai

    1997-12-01

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

  1. 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

  2. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

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

  3. 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)

  4. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Science.gov (United States)

    2013-06-14

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

  5. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. 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

  8. 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.

  9. 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

  10. 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

  11. Recycling heterogeneous americium targets in a boiling water reactor

    International Nuclear Information System (INIS)

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

  12. Electrochemical sensors for application to boiling water reactors

    International Nuclear Information System (INIS)

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

  13. Electrochemical sensors for application to boiling water reactors

    International Nuclear Information System (INIS)

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

  14. 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

  15. Multi-cycle boiling water reactor fuel cycle optimization

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

  17. Industrial application of APOLLO2 to boiling water reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

  5. 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

  6. 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

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

    OpenAIRE

    Fridström, Richard

    2010-01-01

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

  8. 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.)

  9. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. 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

  16. 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 ...

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Hsoung-Wei Chou; Chin-Cheng Huang

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2012-01-01

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

  8. 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 ...

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

    International Nuclear Information System (INIS)

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

  10. 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.

  11. Introduction to the study of boiling in water reactors

    International Nuclear Information System (INIS)

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

  12. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. 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

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

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

  5. 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

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

    International Nuclear Information System (INIS)

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

  7. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Ion chromatography campaigns at several boiling water reactors

    International Nuclear Information System (INIS)

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

  11. 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.

  12. Advanced analytical techniques for boiling water reactor chemistry control

    International Nuclear Information System (INIS)

    The analytical techniques applied can be divided into 5 classes: OFF-LINE (discontinuous, central lab), AT-LINE (discontinuous, analysis near loop), ON-LINE (continuous, analysis in bypass). In all cases pressure and temperature of the water sample are reduced. In a strict sense only IN-LINE (continuous, flow disturbance) and NON-INVASIVE (continuous, no flow disturbance) techniques are suitable for direct process control; - the ultimate goal. An overview of the analytical techniques tested in the pilot loop is given. Apart from process and overall water quality control, standard for BWR operation, the main emphasis is on water impurity characterization (crud particles, hot filtration, organic carbon); on stress corrosion crackling control for materials (corrosion potential, oxygen concentration) and on the characterization of the oxide layer on austenites (impedance spectroscopy, IR-reflection). The above mentioned examples of advanced analytical techniques have the potential of in-line or non-invasive application. They are different stages of development and are described in more detail. 28 refs, 1 fig., 5 tabs

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

    International Nuclear Information System (INIS)

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

  14. 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

  15. 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

  16. Increased fuel column height for boiling water reactor fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Matzner, B.

    1993-06-15

    Rods to maintain said fuel rods upstanding and permitting the exit of water and generated steam is described; a fuel bundle channel extending from said lower tie plate to the vicinity of said upper tie plate, and surrounding said fuel rods therebetween for producing an isolated flow region through said matrix of upstanding fuel rods for the generation of steam by nuclear reaction within said fuel rods; a first plurality of said fuel rods being full length fuel rods for extending fully between said upper and lower tie plates; and, a second plurality of said fuel rods being part length fuel rods for extending part way from a supported disposition on said lower tie plate to a point of fuel rod termination below said upper tie plate whereby a vacated vertical interval is defined between the upper end of said part length fuel rod and said upper tie plate; the improvement to said first plurality of full length fuel rods comprising in combination: said full length fuel rods including a first lower region having a first and smaller diameter containing said pellets of fissionable material; and, at least some of said full length fuel rods including an upper region containing said plenum which is devoid of fuel pellets having a second and larger diameter for providing to said plenum an expanded volume whereby the flow area overlying said part length fuel rods defines additional outflow area adjacent said plenums and the active length of fissionable pellets within said full length fuel rods can be increased.

  17. Increased fuel column height for boiling water reactor fuel rods

    International Nuclear Information System (INIS)

    Rods to maintain said fuel rods upstanding and permitting the exit of water and generated steam is described; a fuel bundle channel extending from said lower tie plate to the vicinity of said upper tie plate, and surrounding said fuel rods therebetween for producing an isolated flow region through said matrix of upstanding fuel rods for the generation of steam by nuclear reaction within said fuel rods; a first plurality of said fuel rods being full length fuel rods for extending fully between said upper and lower tie plates; and, a second plurality of said fuel rods being part length fuel rods for extending part way from a supported disposition on said lower tie plate to a point of fuel rod termination below said upper tie plate whereby a vacated vertical interval is defined between the upper end of said part length fuel rod and said upper tie plate; the improvement to said first plurality of full length fuel rods comprising in combination: said full length fuel rods including a first lower region having a first and smaller diameter containing said pellets of fissionable material; and, at least some of said full length fuel rods including an upper region containing said plenum which is devoid of fuel pellets having a second and larger diameter for providing to said plenum an expanded volume whereby the flow area overlying said part length fuel rods defines additional outflow area adjacent said plenums and the active length of fissionable pellets within said full length fuel rods can be increased

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

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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.

  1. 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

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

    International Nuclear Information System (INIS)

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

  3. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. 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)

  7. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. 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

  16. 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

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2015-09-01

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

  1. 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

  2. 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

  3. Fuzzy logic control of water level in advanced boiling water reactor

    International Nuclear Information System (INIS)

    The feedwater control system in the Advanced Boiling Water Reactor (ABWR) is more challenging to design compared to other control systems in the plant, due to the possible change in level from void collapses and swells during transient events. A basic fuzzy logic controller is developed using a simplified ABWR mathematical model to demonstrate and compare the performance of this controller with a simplified conventional controller. To reduce the design effort, methods are developed to automatically tune the scaling factors and control rules. As a first step in developing the fuzzy controller, a fuzzy controller with a limited number of rules is developed to respond to normal plant transients such as setpoint changes of plant parameters and load demand changes. Various simulations for setpoint and load demand changes of plant performances were conducted to evaluate the modeled fuzzy logic design against the simplified ABWR model control system. The simulation results show that the performance of the fuzzy logic controller is comparable to that of the Proportional-Integral (PI) controller, However, the fuzzy logic controller produced shorter settling time for step setpoint changes compared to the simplified conventional controller

  4. 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

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

    International Nuclear Information System (INIS)

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

  6. 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)

  7. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Cheng, Ling

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

  12. 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

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

    International Nuclear Information System (INIS)

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

  14. 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

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

    International Nuclear Information System (INIS)

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

  16. 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.

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

    International Nuclear Information System (INIS)

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

  19. 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

  20. 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.

  1. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. 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.

  6. 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

  7. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. 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)

  11. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. 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.

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

    International Nuclear Information System (INIS)

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

  16. 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

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

    International Nuclear Information System (INIS)

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

  18. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. 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.)

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

    International Nuclear Information System (INIS)

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

  4. 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)

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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

  8. 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...

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

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

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

  10. 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

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

    International Nuclear Information System (INIS)

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

  12. 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

  13. 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.

  14. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. 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.

  18. 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

  19. 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

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

    International Nuclear Information System (INIS)

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

  1. 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)

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

    International Nuclear Information System (INIS)

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

  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. 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

  6. 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)

  7. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. 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

  19. 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

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

    International Nuclear Information System (INIS)

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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. Remarks on boiling water reactor stability analysis. Pt. 1. Theory and application of bifurcation analysis

    International Nuclear Information System (INIS)

    Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants. (orig.)

  8. Stability monitor for Boiling Water Reactors based on the Multivariate Empirical Mode Decomposition

    International Nuclear Information System (INIS)

    Highlights: • Stability monitor based on a novel technique was developed. • Multivariate Empirical Mode Decomposition (MEMD) to BWR data was applied. • Decision rules permitting to raise an instability alarm based on MEMD. • MEMD can estimate the phase of the density wave from measurements of two LPRMs. • This method detects BWR instabilities (DR increase and out-of phase oscillations). - Abstract: In this work a stability monitor based on a novel technique is presented. This monitor permits to launch general alarms indicating incipient high decay ratios (DR) and out-of-phase oscillations, in a simultaneous way time along. The implemented methodology to determine the estimations of DR and out-of-phase oscillations is based on the Multivariate Empirical Mode Decomposition (MEMD) processing the information obtained from all LPRMs located across the core of Boiling Water Reactor (BWR). The extracted modes with the MEMD, called the Intrinsic Mode Functions (IMFs), permit to tracking the oscillation associated to the density wave. The Case 9 (presenting high DRs and apparently out-of-phase oscillations simultaneously) from the Ringhals stability benchmark was used to show the effectiveness of the proposed methodology

  9. Current status of steam dryer performance under power uprate in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Highlights: • Steam dryer performance after extended power uprate is considered. • Effects of Acoustic Side Branches (ASB) on steam dryers is analyzed. • The ABS represents a reduction in the acoustic loads to the steam dryer. • Spectrograms of signals were obtained for frequency analysis. - Abstract: This work is a compilation of the current status of the steam dryer performance after the implementation of power uprates in Boiling Water Reactors (BWR). Some Nuclear Power Plants (NPPs) have reported failures and cracking in the steam dryer by acoustic resonances that cause excessive vibration due to the increase of steam flow. The replacement of the steam dryer, structural reinforcement and the connection of Acoustic Side Branches (ASB) are the main solutions adopted in order to avoid mechanical failures. The signal analysis of the vibration of the main steam lines in a typical BWR5, was performed using the Short-Time Fourier Transform (STFT). Signals were collected by the strain gauges located around the main steam lines (MSL). Two scenarios are analyzed; the first, is the signal obtained before the installation of the ASB, and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer

  10. Remote mechanized equipment for the repair and replacement of boiling water reactor recirculation loop piping

    International Nuclear Information System (INIS)

    Equipment has been assembled for the remote repair or replacement of boiling water reactor nuclear plant piping in the diameter range of 4 to 28 inches (10-71 cm). The objectives of this program were to produce high-quality pipe welds, reduce plant downtime, and reduce man-rem exposure. The repair strategy was to permit repair personnel to install and check out the repair subsystems and then leave the radiation zone allowing the operations to be conducted at a distance of up to 300 feet (91 m) from the operator. The complete repair system comprises subsystems for pipe severing, dimensional gaging, joint preparation, counterboring, welding, postweld nondestructive inspection (conceptual design), and audio, electronic, and visual monitoring of all operations. Components for all subsystems, excluding those for postweld nondestructive inspection, were purchased and modified as needed for integration into the repair system. Subsystems were designed for two sizes of Type 304 stainless steelpipe. For smaller, 12-inch-diameter (30.5 cm) pipe, severing is accomplished by a power hack saw and joint preparation and counterboring by an internally mounted lathe. The 22-inch-diameter (56 cm) pipe is severed, prepared, and counterbored using an externally mounted, single-point machining device. Dimensional gaging is performed to characterize the pipe geometry relative to a fixed external reference surface, allowing the placement of the joint preparation and the counterbore to be optimized. For both pipe sizes, a track-mounted gas tungsten-arc welding head with filler wire feed is used

  11. Aging analysis on ethylene propylene rubber insulations sampled from boiling water reactor containments

    International Nuclear Information System (INIS)

    Frame-retardant ethylene propylene rubber insulations that were service-used for 16 years in boiling-water reactor containments were subjected to material analyses. Value of elongation at break is approximately 270%, indicating that some change in the functional properties has occurred in their normal operation history. Additional artificial aging based on wear-out approach methodology was then conducted in a 110degC thermostatic oven. Seventy days were needed to reduce the mechanical property till it cannot secure the durability against the loss-of-coolant accident simulation test, which is almost three times greater than the prediction based on the acceleration aging test against non-aged samples. Infrared spectroscopy shows that the formation of carbonyl products is also suppressed in service-used cables. Moreover, gel fraction analysis shows that cross-linking is prominent and is enhanced during the additional wear-out aging in service-used cables. The relation between elongation at break value and tensile strength shows similar indication, and demonstrates that chain scission due to oxidation is elevated in the case of acceleration aging. Cross-linking phenomenon in used cables is considered to prevent chemical degradation reaction, by preventing the oxygen permeation into the bulk. This intrinsic feature of service-used cable insulation has a possibility of contributing to the slow aging under the in-service atmosphere, which was already found by the statistical analysis of the aging state data obtained for the corresponding type of insulations. (author)

  12. Multivent effects in a large scale boiling water reactor pressure suppression system

    International Nuclear Information System (INIS)

    The steam-driven GKSS pressure suppression test facility, which contains 3 full scale vent pipes, has been used for 5 years to investigate the postulated loss-of-coolant accident in a Mark II and Type 69 boiling water reactor. Using the results from several of these tests, wetwell boundary load data (peak pressures and spectral power) during the chugging stage, have been evaluated for sparse pool response (one and two vents in the three vent pool) and for full pool response (one, two, or three vent operation in pools of constant wetwell pool area per vent). The sparse pool results indicate the pool-system, chug event boundary loads are strongly dependent on wetwell pool area per vent, with the load increasing with decreasing area. The full pool results show a substantial increase in the pool-system, chug event boundary loads upon a change from single cell to double cell operation; only minor change occurs in going from double to triple cell operation

  13. 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

  14. 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

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

    International Nuclear Information System (INIS)

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

  16. AXIAL: a system for boiling water reactor fuel assembly axial optimization using genetic algorithms

    International Nuclear Information System (INIS)

    A system named AXIAL is developed based on the genetic algorithms (GA) optimization method, using the 3D steady state simulator code Core-Master-PRESTO (CM-PRESTO) to evaluate the objective function. The feasibility of this methodology is investigated for a typical boiling water reactor (BWR) fuel assembly (FA). The axial location of different fuel compositions is found in order to minimize the FA mean enrichment needed to obtain the cycle length under the safety constraints. Thermal limits are evaluated at the end of cycle using the Haling calculation; the hot excess reactivity and the shutdown margin at the beginning of cycle are also evaluated. The implemented objective function is very flexible and complete, incorporating all the thermal and reactivity limits imposed during fuel design analysis; furthermore, additional constraints can be easily introduced in order to obtain an improved solution. The results show a small improvement in the FA average enrichment obtained with the system related to the reference case that has been studied. The results show that the system converge to an optimal solution, it is observed that the mean fuel enrichment decreases while all the constraints are satisfied. A comparison was also performed using one-point and two-points crossover operator and the results of a sensitivity study for different mutation percentage are also showed

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

    International Nuclear Information System (INIS)

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

  18. 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.

  19. 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.

  20. Cryogenic system for collecting noble gases from boiling water reactor off-gas

    International Nuclear Information System (INIS)

    In boiling water reactors, noncondensible gases are expelled from the main condenser. This off-gas stream is composed largely of radiolytic hydrogen and oxygen, air in-leakage, and traces of fission product krypton and xenon. In the Air Products' treatment system, the stoichiometric hydrogen and oxygen are reacted to form water in a catalytic recombiner. The design of the catalytic recombiner is an extension of industrial gas technology developed for purification of argon and helium. The off-gas after the recombiner is processed by cryogenic air-separation technology. The gas is compressed, passed into a reversing heat exchanger where water vapor and carbon dioxide are frozen out, further cooled, and expanded into a distillation column where refrigeration is provided by addition of liquid nitrogen. More than 99.99 percent of the krypton and essentially 100 percent of the xenon entering the column are accumulated in the column bottoms. Every three to six months, the noble-gas concentrate accumulated in the column bottom is removed as liquid, vaporized, diluted with steam, mixed with hydrogen in slight excess of oxygen content, and fed to a small recombiner where all the oxygen reacts to form water. The resulting gas stream, containing from 20 to 40 percent noble gases, is compressed into small storage cylinders for indefinite retention or for decay of all fission gases except krypton-85, followed by subsequent release under controlled conditions and favorable meteorology. This treatment system is based on proven technology that is practiced throughout the industrial gas industry. Only the presence of radioactive materials in the process stream and the application in a nuclear power plant environment are new. Adaptations to meet these new conditions can be made without sacrificing performance, reliability, or safety

  1. 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)

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

  5. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  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

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

    International Nuclear Information System (INIS)

    Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

  15. Water quality control in primary cooling system of crud concentration suppressed boiling water reactor, (1)

    International Nuclear Information System (INIS)

    The No.2 Unit of Fukushima-Daini Nuclear Power Plant (2F-2; 1,100 MWe) was commercially operated for 10,320 effective full power hours (EFPH) as its first fuel cycle. The basic design concept of the 2F-2 incorporated the following two features : (1) Application of procedures for reducing shutdown dose rate based on the Japanese Improvement and Standardization Program, (2) Low crud generation to minimize radioactive waste by careful material selection for the primary system. Thus, it was possible to keep the average Fe concentration in the condensate water at less than 6 ppb during the first fuel cycle. As a result of this low value, the average life of powdered resin precoated prefilters was extended to about a month, and the average chemical regeneration period of the deep bed demineralizers was extended to more than one year. The water chemistry of the 2F-2 was characterized by low 60Co and high 58Co radioactivities in the reactor water, which resulted in a low shutdown dose rate determined mainly by 58Co depositing on the primary piping. For example, average dose rate around the primary piping just after reactor shutdown was about 70 mR/h, about 75 % of which was from 58Co depositing on the pipe inner surfaces. The contribution of 60Co was about 25 %. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. 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)

  19. 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.

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

    International Nuclear Information System (INIS)

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

  1. The installation of PEANO at the Halden Boiling Water Reactor: first test and results

    International Nuclear Information System (INIS)

    After extensive testing of PEANO with data from process simulators, the next step was to set-up an installation in a real process, where the signal validation is performed online. For this purpose the Halden Boiling Water Reactor was used. One implication is that recorded process data from past operation would be used for the training of the system. This type of data is corrupted with errors, faults, process noise or even previous sensor problems, which need to be removed before the data can be used. The pre-processing was therefore a very import step during this installation. A 15 minute average of 29 process signals, spread out over the primary, secondary and tertiary loops, was used. At the end of the design process a fuzzy-neural network resulted containing 5 clusters, that has been trained with over 20.000 patterns. To establish the TCP/IP connection to the process computer and receiving the process data in real-time, some extra software was developed. With this installation it has been shown that it is possible to have the PEANO Server and PEANO Client (monitoring unit) running on one machine (e.g. in the control room), while additional monitoring units are connected from a remote location (e.g. main office building). The first results show that the installation of PEANO is capable of performing its validation task properly, even during transients. Both start-up and shutdown situations can be handled without any problems. In situations where incoming patterns represent unknown process situations that have not been encountered during the training, the 'I don't know' answer was given. To test the ability to detect a sensor failure off-line tests have been run, where sensor faults and drifts were added (author) (ml)

  2. Radiolysis of boiling water

    Science.gov (United States)

    Yang, Shuang; Katsumura, Yosuke; Yamashita, Shinichi; Matsuura, Chihiro; Hiroishi, Daisuke; Lertnaisat, Phantira; Taguchi, Mitsumasa

    2016-06-01

    γ-radiolysis of boiling water has been investigated. The G-value of H2 evolution was found to be very sensitive to the purity of water. In high-purity water, both H2 and O2 gases were formed in the stoichiometric ratio of 2:1; a negligible amount of H2O2 remained in the liquid phase. The G-values of H2 and O2 gas evolution depend on the dose rate: lower dose rates produce larger yields. To clarify the importance of the interface between liquid and gas phase for gas evolution, the gas evolution under Ar gas bubbling was measured. A large amount of H2 was detected, similar to the radiolysis of boiling water. The evolution of gas was enhanced in a 0.5 M NaCl aqueous solution. Deterministic chemical kinetics simulation elucidated the mechanism of radiolysis in boiling water.

  3. Nuclear-coupled thermal-hydraulic stability analysis of boiling water reactors

    Science.gov (United States)

    Karve, Atul A.

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model we developed from: the space-time modal neutron kinetics equations based on spatial omega-modes, the equations for two-phase flow in parallel boiling channels, the fuel rod heat conduction equations, and a simple model for the recirculation loop. The model is represented as a dynamical system comprised of time-dependent nonlinear ordinary differential equations, and it is studied using stability analysis, modern bifurcation theory, and numerical simulations. We first determine the stability boundary (SB) in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value and then transform the SB to the practical power-flow map. Using this SB, we show that the normal operating point at 100% power is very stable, stability of points on the 100% rod line decreases as the flow rate is reduced, and that points are least stable in the low-flow/high-power region. We also determine the SB when the modal kinetics is replaced by simple point reactor kinetics and show that the first harmonic mode has no significant effect on the SB. Later we carry out the relevant numerical simulations where we first show that the Hopf bifurcation, that occurs as a parameter is varied across the SB is subcritical, and that, in the important low-flow/high-power region, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line. Hence, a point on the 100% rod line in the low-flow/high-power region, although stable, may nevertheless be a point at which a BWR should not be operated. Numerical simulations are then done to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is determined that the NRC requirement of DR flow/high-power region and hence these points

  4. Scaling laws and design aspects of a natural-circulation-cooled simulated boiling water reactor fuel assembly

    International Nuclear Information System (INIS)

    In order to study the thermohydraulic behavior of a natural-circulation-cooled boiling water reactor (BWR) fuel assembly, such as void drift, flow pattern distribution, and stability, a scaled loop geometry is designed. For modeling the steam/water flow in a BWR fuel assembly, scaling criteria are derived using the one-dimensional drift-flux model. Thermal equilibrium and subcooled boiling conditions are treated separately, resulting in one overall set of criteria. Scaling on all flow regimes that can be present in a normal fuel assembly leads to fixing both the assembly mass flux and the geometric dimensions. When Freon-12 is used as a modeling fluid, model assembly dimensions must be 0.46 of the prototype. Total power consumption must be reduced by a factor 50. To sustain cooling by natural circulation, a modeled chimney and downcomer are included

  5. Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

    International Nuclear Information System (INIS)

    An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have a strong influence on the Type I and Type II instabilities observed at low and high channel powers, respectively. Also, it was found that the coupled multipoint kinetics model and the modal point kinetics model predict the same threshold power for out-of-phase instability if the coupling coefficient in the former model is half the eigen value separation between the fundamental and the first harmonic mode in the latter model. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (orig.)

  6. Water chemistry improvements in an operating boiling water reactor (BWR) and associated benefits

    International Nuclear Information System (INIS)

    Kernkraft Muhleberg (KKM) nuclear power plant is a BWR/4, the older of the two BWRs in Switzerland located in the outskirts of Bern. The plant is currently in its 37th year of continuous power operation, and has implemented major water chemistry improvements, including, hydrogen water chemistry (HWC), depleted zinc oxide (DZO) addition, NobleChem™, and On-Line NobleChem™ applications. In addition, the KKM plant has also performed other improvements such as maintaining low reactor water conductivity to mitigate intergranular stress corrosion crack (IGSCC) initiation and growth, as well as taking numerous actions to control radiation source term reduction. The actions taken to control the latter include replacement of the brass condenser tubes and an active cobalt source term reduction plan by eliminating the stellite control rod pins and rollers. These water chemistry improvements at the KKM plant have resulted in lower operating dose rates, lower drywell (shut down) dose rates and mitigation of shroud cracks. It is important to note that KKM is the only plant in the BWR industry that has monitored shroud internal diameter (ID) crack growth rates on a consistent basis using ultrasonic testing (UT) since 1993, thus providing an enormously valuable contribution to the BWR industry's in-plant crack growth rate data base. KKM plant has also installed tie rods in the shroud in 1996, an industry accepted approach. In addition, KKM also implemented NobleChem™ and On-Line NobleChem™ (OLNC) along with low hydrogen injection as additional proactive measures in 2000 and 2005 respectively to mitigate the growth of shroud cracks. There is reasonably clear evidence that since the implementation of OLNC, there is a consistent reduction in shroud crack growth rates showing mitigation of existing cracks. It is also evident that the drywell dose rates are showing a continuing decrease following 60Co source term reductions, DZO and OLNC implementations. This paper

  7. Development of a fully-consistent reduced order model to study instabilities in boiling water reactors

    International Nuclear Information System (INIS)

    A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to other existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or Cmn*V,D - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good

  8. 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

  9. 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

  10. 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.)

  11. 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.

  12. 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

  13. Acoustic Analysis for a Steam Dome and Pipings of a 1,100 MWe-Class Boiling Water Reactor

    International Nuclear Information System (INIS)

    For the integrity evaluation of steam dryers in up-rated nuclear power plants, we have applied acoustic analysis to a nuclear power plant steam dome and main steam pipings. We have selected a 1,100 MWe-class boiling water reactor as a subject of the analysis. We have constructed a three-dimensional finite element model, and conducted acoustic analyses. The analysis result suggested that the origin of steam pressure pulsation in high frequency range was due to vortex shedding at standpipes. (authors)

  14. 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.)

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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

  20. 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.

  1. The research on control rod insertion of a boiling water reactor with water hydraulic drive

    International Nuclear Information System (INIS)

    This thesis reports on the hydraulic driving system, powered by an accumulator. This drive system is mainly used for the drive of control rods of nuclear reactors. In case of strong earthquakes, control rods are set in gaps between fuel assemblies to scram nuclear reactors. Characteristics of the system have not been analyzed. The analysis of this system is necessary in order to present the designs that are intended to be a variety of situations. So we developed the model of the hydraulic control rod driving system. The model that we have created is able to reproduce the actual driving. Also, there is a load on the system by an earthquake. This load is caused by the contact of the deformed fuel assembly and control rod. This load model is obtained by solving the equation of motion of the beam. (author)

  2. SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

    International Nuclear Information System (INIS)

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

  3. SWR 1000: A Next-Generation Boiling Water Reactor Ready for Deployment

    International Nuclear Information System (INIS)

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 design is particularly suitable for countries whose power systems do not include any large power plants. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (IandC) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10x10 rod array to a 12x12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free start-up, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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)

  11. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    Science.gov (United States)

    Trianti, Nuri; Su'ud, Zaki; Arif, Idam; Riyana, EkaSapta

    2014-09-01

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  12. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    International Nuclear Information System (INIS)

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained

  13. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

    International Nuclear Information System (INIS)

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix

  14. 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.

  15. 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.

  16. 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.

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2011-09-01

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

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

    International Nuclear Information System (INIS)

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

  20. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    International Nuclear Information System (INIS)

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. The paper describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motorgenerator set.

  1. In-core power sharing and fuel requirement study for a decommissioning Boiling Water Reactor using the linear reactivity model

    International Nuclear Information System (INIS)

    Highlights: • Linear reactivity model (LRM) was modified and applied to Boiling Water Reactor. • The power sharing and fuel requirement study of the last cycle and two cycles before decommissioning was implemented. • The loading pattern design concept for the cycles before decommissioning is carried out. - Abstract: A study of in-core power sharing and fuel requirement for a decommissioning BWR (Boiling Water Reactor) was carried out using the linear reactivity model (LRM). The power sharing of each fuel batch was taken as an independent variable, and the related parameters were set and modified to simulate actual cases. Optimizations of the last cycle and two cycles before decommissioning were both implemented; in the last-one-cycle optimization, a single cycle optimization was carried out with different upper limits of fuel batch power, whereas, in the two-cycle optimization, two cycles were optimized with different cycle lengths, along with two different optimization approaches which are the simultaneous optimization of two cycles (MO) and two successive single-cycle optimizations (SO). The results of the last-one-cycle optimization show that it is better to increase the fresh fuel power and decrease the thrice-burnt fuel power as much as possible. It also shows that relaxing the power limit is good to the fresh fuel requirement which will be reduced under lower power limit. On the other hand, the results of the last-two-cycle (cycle N-1 and N) optimization show that the MO is better than SO, and the power of fresh fuel batch should be decreased in cycle N-1 to save its energy for the next cycle. The results of the single-cycle optimization are found to be the same as that in cycle N of the multi-cycle optimization. Besides that, under the same total energy requirement of two cycles, a long-short distribution of cycle length design can save more fresh fuel

  2. 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.

  3. Checking technical measurements on climatic data during sand blasting and spraying work in the condensation chamber of the boiling water reactor Gundremmingen

    International Nuclear Information System (INIS)

    During sand blasting and spraying work in the condensation chambers of boiling water reactors prescribed climatic data must be adhered to. For this purpose temporary air conditioners are used. The technical measurement examination here should provide information as to whether the air conditioners used were to fulfill the parameter curve specifications. (orig.)

  4. Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

    International Nuclear Information System (INIS)

    This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value. (author)

  5. Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

    International Nuclear Information System (INIS)

    In this paper, we present the three-dimensional finite element seismic response analysis of the full-scale boiling water reactor BWR5 at the Kashiwazaki-Kariwa Nuclear Power Plant subjected to the Niigata-ken Chuetsu-Oki (NCO) earthquake that occurred on 16th July 2007. During the earthquake, the automatic shutdown of the reactors was performed successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found through in-depth investigation that there was no significant damage of the reactor cores or other important systems, structures and components (SSCs). In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of SSCs. Although the lumped mass model has worked well so far for a seismic proof design, more precise methods should be developed to understand response behaviors visually. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize the dynamic behaviors of this model. Through the comparison of the analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value. (author)

  6. 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.

  7. 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.

  8. 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)

  9. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

    2015-05-15

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

  10. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    International Nuclear Information System (INIS)

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

  11. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    Energy Technology Data Exchange (ETDEWEB)

    Rubio, Rafael, E-mail: rrubio@iberdrola.es [Iberdrola Generación Nuclear S.A., Madrid (Spain); Jimenez, Gonzalo [Universidad Politécnica de Madrid (Spain)

    2014-08-15

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

  12. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    International Nuclear Information System (INIS)

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available

  13. 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

  14. 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

  15. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    The results of two in-reactor defect tests of Zircaloy-clad U3Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270oC. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U3Si at 300oC is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lund, A.L.

    1997-11-01

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

  4. 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

  5. An in-line diffuse reflection spectroscopy study of the oxidation of stainless steel under boiling water reactor conditions

    International Nuclear Information System (INIS)

    A novel cell unit was constructed to measure in-line the oxide layer build-up on a stainless steel sample by Diffuse Reflection Spectroscopy (DRS; ultraviolet, visible, near infrared) under boiling water reactor (BWR) conditions. The stainless steel samples, observed in the cell through a sapphire window, are contacted with oxidising hot water (300oC, 9.0 MPa). Using a cold finger with the optical fibre probe, the spectroscopic investigations (200-1000 nm wavelength) were performed at a fixed position from the sapphire window. The DRS spectra are the result of the coupling of both absorption (chemical) and interferometric (physical) processes. Analysis of these spectral components allows the independent determination of the oxide layer thickness. The build-up of the oxide layer may be directly observed and quantified, nanometre by nanometre, from 2 to 200 nm. This powerful technique may be used to study the early corrosion rates of stainless steel under BWR conditions and should allow the development of a strategy to reduce corrosion. (Author)

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

  11. 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...

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

    OpenAIRE

    Kulesza Joel A.; Roudén Jenny; Green Eva-Lena

    2016-01-01

    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 ...

  13. 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

  14. Automatic fuel lattice design in a boiling water reactor using a particle swarm optimization algorithm and local search

    International Nuclear Information System (INIS)

    Highlights: ► The automatic procedure was developed to design the radial enrichment and gadolinia (Gd) distribution of fuel lattice. ► The method is based on a particle swarm optimization algorithm and local search. ► The design goal were to achieve the minimum local peaking factor. ► The number of fuel pins with Gd and Gd concentration are fixed to reduce search complexity. ► In this study, three axial sections are design and lattice performance is calculated using CASMO-4. - Abstract: The axial section of fuel assembly in a boiling water reactor (BWR) consists of five or six different distributions; this requires a radial lattice design. In this study, an automatic procedure based on a particle swarm optimization (PSO) algorithm and local search was developed to design the radial enrichment and gadolinia (Gd) distribution of the fuel lattice. The design goals were to achieve the minimum local peaking factor (LPF), and to come as close as possible to the specified target average enrichment and target infinite multiplication factor (k∞), in which the number of fuel pins with Gd and Gd concentration are fixed. In this study, three axial sections are designed, and lattice performance is calculated using CASMO-4. Finally, the neutron cross section library of the designed lattice is established by CMSLINK; the core status during depletion, such as thermal limits, cold shutdown margin and cycle length, are then calculated using SIMULATE-3 in order to confirm that the lattice design satisfies the design requirements.

  15. The effective convectivity model for simulation of molten metal layer heat transfer in a boiling water reactor lower head

    International Nuclear Information System (INIS)

    The paper is concerned with development of models for assessment of Control Rod Guide Tube (CRGT) cooling efficiency in Severe Accident Management (SAM) for a Boiling Water Reactor (BWR). In case of core melt relocation under a certain accident condition, there is a potential of stratified (with a metal layer atop) melt pool formation in the lower plenum. For simulations of molten metal layer heat transfer we are developing the Effective Convectivity Model (ECM) and Phase-change ECM (PECM). The models are based on the concept of effective convectivity previously developed for simulations of decay-heated melt pool heat transfer. The PECM platform takes into account mushy zone convection heat transfer and compositional convection that enables simulations of non-eutectic binary mixture solidification and melting. The ECM and PECM are validated against various heat transfer experiments for both eutectic and non-eutectic mixtures, and benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is applied to heat transfer simulation of a stratified heterogeneous debris pool in the presence of CRGT cooling. The PECM simulation results show no focusing effect in the metal layer on top of a debris pool formed in the BWR lower plenum and apparent efficacy of the CRGT cooling which can be served as an effective SAM measure to protect the vessel wall from thermal attacks and mitigate the consequences of a severe accident. (author)

  16. Failure analysis of the standby liquid control system for a boiling water reactor with fuzzy cognitive maps

    International Nuclear Information System (INIS)

    Highlights: → FCMs are proposed in order to determine failure modes in systems and equipment in BWRs. → A simplified model is compared with the fault tree analysis technique. → Five case scenarios are studied in order to test the performance of the method. → The proposed method shows consistency with the traditional fault tree technique. - Abstract: A fuzzy cognitive maps (FCM) application is proposed as a simple method to determine failure modes and effects of the standby liquid control system (SLC) during anticipated transient without scram (ATWS) in a boiling water reactor (BWR). The SLC has an important contribution to the total core damage frequency in a BWR. This is the first step in the development of an expert system that could involve many emergency systems of a BWR to simulate accident sequences, through the knowledge representation and reasoning with FCM designs in order to automate the decision making process. A simplified model of the SLC is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, in order to see that both techniques show similar results even if the approaches are different.

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

    International Nuclear Information System (INIS)

    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 233U 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 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 fuel assembly

  18. 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)

  19. Implementation of the HELIOS code for the Halden boiling water reactor (HBWR)

    International Nuclear Information System (INIS)

    Since the beginning of 1998 the 2-D transport code HELIOS has been used for neutronic physics calculations at the Institutt for energiteknikk. This note describes the system which has been built up to enable HELIOS to quickly and accurately perform the calculations required and is a concise review of the HELIOS system at the Halden Reactor Project. The Halden Reactor is a small but diverse test reactor. A brief outline of the core and its contents is given to give an indication of the complexity of some of the test rigs to be found in this reactor. HELIOS is a neutron and gamma transport code. A brief outline of the code itself is given as well as a description of its implementation at Halden. HELIOS is a very versatile code resulting in many applications. The majority of its uses so far are described in this note. The most common application is the calculation of fuel depletion functions which are used for power determination at Halden. One of the problems of accurately modelling the HBWR is what fraction of the reactor needs to be modelled. An example of a fuel depletion calculation is given to demonstrate the effects of varying the size of model used. This then leads to the minimum size required to accurately model the surroundings of a fuel assembly

  20. 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

  1. 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...

  2. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

    ... are specifically designed to ensure that the reactor can be shutdown and decay heat can be removed..., 2009 (74 FR 62829). On June 12, 2009 (74 FR 28112), the NRC amended its regulations to require... proposed rule in the Federal Register on January 20, 2011 (76 FR 3540). The public comment period for...

  3. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    International Nuclear Information System (INIS)

    organizations from 7 countries, SMART, integrated reactor, developed by Korea Atomic Energy Research Institute, Republic of Korea; CAREM, Argentina integrated reactor; MRX, integrated reactor, developed by Japan Atomic Energy Research Institute; UNITERM, NPP with integrated reactor, development by Research and development institute of power engineering (NIKIET), Russian Federation; AHEC-80, Russian NPP with integrated reactor, developed by OKB Mechanical Engineering (OKBM), Nizhny Novgorod, Russia. Moreover, following Boiling Water Reactor (BWR) projects have been subjected to the system comparative analysis. 1) Large Sized Reactors: ABWR, developed by Hitachi, Ltd, Japan, Toshiba Corporation, Japan ? G.E. USA; BWR-90, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; BWR-90+, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; SWR 1000, developed by Siemens Corporation, Germany; ESBWR, developed by General Electric Company, USA. 2) Medium Sized Reactors: SBWR, developed by General Electric Company, USA; HSBWR, developed by Hitachi Company, Ltd. 3) Small Sized Reactors: SSBWR, developed by Hitachi, Ltd, Japan; VK-300, BWR reactor, developed by Research and Development Institute of Power Engineering (NIKIET), Russia. Some data on the analysis of the condition and prospects of energy production and energy consumption, stations and networks in Kazakhstan are given. According to this analysis of nuclear power plants of average and low power are considered to be the most appropriate to construction in Kazakhstan. Recommendations on a choice of the most safe, reliable and economically competitive reactors have been made among the above-mentioned ones PWR, WWER and BWR for construction in Kazakhstan

  4. 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)

  5. 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.

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  8. Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

    International Nuclear Information System (INIS)

    concept - to calculate the control rod's working life both in the control position and the shut-down position - will automatically lead to an optimization of the control rod strategy. Control rod optimisation is demonstrated by accumulating the total amount of control rods required in a medium-sized BWR up to the total reactor holding period. At least 60% of the first core inventory - for this control rod type an existing EMPIRICAL MODEL is already available - may be used up to the total operating period without any safety loss. Looking to the present disposal situation this concept represents a practical way to reduce all high level waste. In addition benefit of utilizing this concept is that it minimizes tritium emission. Control-rods utilized within Boiling Water Reactors (BWR) are designed for the purpose to control and shape the neutron flux profile in the reactor, to adjust the range of regulation referring to the weight rate of the reactor coolant and thirdly by- shutting down the reactor at any time and under any conditions with regard to nuclear aspects, mechanical integrity and control rod history. The designation control- or shut down rod characterize the particular field of activity for a given control rod. The focal point of my work had shown to be a calculation of the nuclear working life of any control rod design as well as an optimisation method with reference to the holding period for a given control rod inventory as a result of measuring data and a theoretical analysis describing the parameters in a general validity form. (author)

  9. 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

  10. 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

  11. 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)

  12. Nuclear Co-Generation Desalination Complex with Simplified Boiling Water Reactor VK-300

    International Nuclear Information System (INIS)

    With regard for the global-scale development of desalination technologies and the stable growth demand for them, Russia also takes an active part in the development of these technologies. Nuclear Desalination Complex (NDC) with VK-300 reactor facility is a modification of a nuclear power unit with VK-300 reactor developed for application at Russian nuclear cogeneration plants with the heat supply for desalination needs, up to 400 Gcal/h of thermal energy and with the simultaneous electricity generation of about 180 MW. The report considers a VK-300 reactor based and NDC with MED based distillation desalination units with horizontal-tube film evaporators. As it provides with thermal energy a desalination complex with the capacity of 300.000 m3/day, a nuclear plant consisting of two VK-300 power units allows production of distillate with the cost of 0.58 dollars/m3. In this case, the electricity supply to the power system is 357 MW(e). The electricity cost is 0.029 dollars/kWh. (author)

  13. LWR-PROTEUS Verification of Reaction Rate Distributions in Modern 10 x 10 Boiling Water Reactor Fuel

    International Nuclear Information System (INIS)

    HELIOS, CASMO-4, and MCNP4B calculations of reaction rate distributions in a modern, fresh 10 x 10 boiling water reactor fuel element have been validated using the experimental results of the LWR-PROTEUS Phase I project corresponding to full-density water moderation conditions (core 1B). The reaction rate distributions measured with a special gamma-scanning machine employing twin germanium detectors consisted of total fission Ftot and 238U-capture C8. The average statistical errors for the gamma scans were better than 0.5% for Ftot and 0.9% for C8. The rod-by-rod measurements were performed on 60 different fuel rods selected from the central part of a test zone consisting of actual, fresh SVEA-96+ fuel elements, thus gaining in realism by departing from conventional fuel rod mockups. In the case of Ftot, the root-mean-square (rms) of the rod-by-rod distribution of differences between calculational and experimental (C-E) values has been found to be 1.1% for HELIOS and for CASMO-4, and 1.3% for MCNP4B. For C8, the rms values of the (C-E) distributions are 1.0, 1.3, and 1.4% as obtained with HELIOS, CASMO-4, and MCNP4B, respectively. The effects of using different data libraries (ENDF/B-V, ENDF/B-VI, and JEF-2.2) with MCNP4B were also studied and have been found to be small

  14. Detailed B-10 depletion in control rods operatingin a Nuclear Boiling Water Reactor

    OpenAIRE

    Johnsson, John

    2011-01-01

    In a nuclear power plant, control rods play a central role to control the reactivity ofthe core. In an inspection campaign of three control rods (CR 99) operated in theKKL reactor in Leibstadt, Switzerland, during 6 respectively 7 consecutive cycles,defects were detected in the top part of the control rods due to swelling caused bydepletion of the neutron-absorbing 10B isotope (Boron-10). In order to correlatethese defects to control rod depletion, the 10B depletion has in this study beencalc...

  15. Benchmark for a 3D Monte Carlo boiling water reactor fluence computational package - MF3D

    International Nuclear Information System (INIS)

    A detailed three dimensional model of a quadrant of an operating BWR has been developed using MCNP to calculate flux spectrum and fluence levels at various locations in the reactor system. The calculational package, MF3D, was benchmarked against test data obtained over a complete fuel cycle of the host BWR. The test package included activation wires sensitive in both the fast and thermal ranges. Comparisons between the calculational results and test data are good to within ten percent, making the MF3D package an accurate tool for neutron and gamma fluence computation in BWR pressure vessel internals. (orig.)

  16. On core debris behaviour in the pressure vessel lower head of nordic boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I.; Ikonen, K. [VTT Energy, Espoo (Finland); Hedberg, K. [Vattenfall Energistystem AB, Stockholm (Sweden); Thomsen, K. [Risoe National Lab., Roskilde (Denmark)

    1997-10-01

    In-vessel melt progression in Nordic BWRs has been studied as part of the RAK-2 project within the Nordic Nuclear Safety Programme 1994-1997. A part of the study was the evaluation of the late phase melt progression phenomena and the thermal behaviour of core debris, the pressure vessel wall and the lower head penetrations during a severe accident. The investigations presented here focus on BWR cases. The MELCOR/Bottom Head Package was applied to investigate the core debris bed behaviour and thermal response of structures in the case of the Olkiluoto 1 and 2 reactor vessel lower head. Both low and high pressure scenarious were analysed with sensitivity studies addressing the effects of debris bed porosity, debris particle size and reflooding of the dry debris bed. Lower head failure mechanisms and timing were examined by allowing instrument tube failure (normal case) or by deactivating the penetration failure model with an input option. Due to modelling assumptions in MELCOR, all presented calculations examine thermal behaviour of a rubble bed in the lower head. Calculated results are evaluated against experimental data. Studies using Forsmark 3 input data were carried out with the MAAP4 code. Studied cases covered also low and high pressure sequences, and a number of sensitivity calculations varying a few key parameters were performed. Only creep rupture of the reactor pressure vessel (RPV) was considered in the MAAP4 analyses. 33 refs.

  17. 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.

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

    Science.gov (United States)

    Kulesza, Joel A.; Roudén, Jenny; Green, Eva-Lena

    2016-02-01

    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.

  19. On core debris behaviour in the pressure vessel lower head of nordic boiling water reactors

    International Nuclear Information System (INIS)

    In-vessel melt progression in Nordic BWRs has been studied as part of the RAK-2 project within the Nordic Nuclear Safety Programme 1994-1997. A part of the study was the evaluation of the late phase melt progression phenomena and the thermal behaviour of core debris, the pressure vessel wall and the lower head penetrations during a severe accident. The investigations presented here focus on BWR cases. The MELCOR/Bottom Head Package was applied to investigate the core debris bed behaviour and thermal response of structures in the case of the Olkiluoto 1 and 2 reactor vessel lower head. Both low and high pressure scenarious were analysed with sensitivity studies addressing the effects of debris bed porosity, debris particle size and reflooding of the dry debris bed. Lower head failure mechanisms and timing were examined by allowing instrument tube failure (normal case) or by deactivating the penetration failure model with an input option. Due to modelling assumptions in MELCOR, all presented calculations examine thermal behaviour of a rubble bed in the lower head. Calculated results are evaluated against experimental data. Studies using Forsmark 3 input data were carried out with the MAAP4 code. Studied cases covered also low and high pressure sequences, and a number of sensitivity calculations varying a few key parameters were performed. Only creep rupture of the reactor pressure vessel (RPV) was considered in the MAAP4 analyses

  20. 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)

  1. 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)

  2. 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

  3. 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

  4. Boiling Water Reactor Turbine Trip (TT) Benchmark - Vol. IV Volume IV: Summary Results of Exercise 3

    International Nuclear Information System (INIS)

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling of the core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for that purpose. The present volume is the last in a series of four and summarises the results of the third benchmark exercise, which analyses a turbine trip (TT) in a BWR in its entirety, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. Exercise 3 also analyses four extreme scenarios which allowed participants to test the capabilities of their code(s) in terms of coupling and feedback modelling. The data made available from experiments carried out at the plant make the present benchmark particularly valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. Nuclear physics calculation for IMF-MOX fuel irradiation test in the halden boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Noh, Jae Man; Kim, Ha Yong [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    As a part of activity for future fuel development project in KAERI, test MOX fuel rods are going to be loaded and irradiated in Halden reactor under a KAERI's joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device developed by KAERI. The test fuel assembly rig contains total six test fuel rods: three MOX rods and three inert matrix fuel rods. For comparison purposes, one of three MOX rods will be fabricated by BNFL and the other two MOX rods will be manufactured jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with (Zr-Y-Er-Pu) oxide by PSI. One of the three IMF rods will be fabricated with wet process and the other two rods will be fabricated with dry process. Neutronic calculation was preliminarily performed for the test fuel assembly to generate the rod power histories of test rods which will be use to evaluate the irradiation performance of the test fuel rods. A power scenario is proposed to meet the experiment conditions, target linear power and burnup. In accordance with this power scenario, the maximum linear power rate is about 350 W/cm, considering the axial power peaking factor of 1.05, for both IMF and MOX. After five years irradiation, the achieved burnup will be about 49 MWD/kgHM for MOX and 490 MWD/kgHM for IMF. To get the proposed power scenario, the expected irradiation position may be in the ring 6 of the HBWR core the position of test fuel to be loaded in HBWR will also depend on the environment. A surrounding of these spikes, as considered, in a symmetric configuration would give a flat power distribution in the IMF/MOX-IFA and hence the most successful results. (author). 21 figs., 4 tabs.

  8. 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

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

    International Nuclear Information System (INIS)

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

  10. Test results employed by General Electric for boiling water reactor containment and vertical vent loads

    International Nuclear Information System (INIS)

    During a safety relief valve blowdown, air contained in the relief line discharges into the suppression pool with the resulting oscillations of the air bubble causing dynamic loading on the containment. The magnitude and characteristics of such loading depend upon the geometry of the discharge device at the end of the safety relief line. Extensive small scale and large scale testing was performed to evaluate the performance of a four-arm quencher discharge device. Results of these tests, description of test facility, instrumentation and test procedures are described. During a loss-of-coolant accident, steam flows through vertical vent pipes such as employed in Mark I and II Containments and condenses in the suppression pool at the vent exit. During this condensation process, a steam bubble which forms at the vent exit will collapse irregularly leading to water impingement on the vent pipe. The water impingement phenomenon causes lateral loading on the vertical vents. The loading phenomena and series of tests performed to evaluate the load magnitudes are described. During a later part of the safety relief valve blowdown, steam discharges into the suppression pool through the safety relief line end discharge device. Extensive tests were carried out to investigate the high temperature condensation phenomenon and the temperature threshold limits for the occurrence of condensation vibrations for various configurations including the quencher configuration, of the relief line and discharge device. Results of these tests including a description of the test facility, instrumentation and test procedures have been included

  11. The Behavior of Corrosion Products in Sampling Systems under Boiling Water Reactor Conditions

    International Nuclear Information System (INIS)

    A high pressure loop has been used to simulate sampling systems employed under BWR conditions. The reliability of the sampling method was studied in a series of six test runs. A variety of parameters that are thought to influence the reliability of the sampling was investigated. These included piping geometry, water oxygen content, flow, temperature and temperature gradients. Amongst other things the results indicate that the loss by deposition of iron containing corrosion products does not exceed 50 %; this figure is only influenced to a minor extent by the above mentioned parameters. The major part of the corrosion products thus deposited is found along the first few meters of the piping and cooler coil. A moderate prolongation of a pipe which is already relatively long should thus be incapable of producing a major influence on the sampling error

  12. 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

  13. 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

  14. 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

  15. Process and device for measuring the level in a reactor pressure vessel of a boiling water reactor

    International Nuclear Information System (INIS)

    The differential pressure is measured between the lower space filled with liquid and a comparison column which is connected to the upper part space filled with steam. From this measurement and the liquid and steam densities in the pressure vessel and in the comparison column and the effects of flow at the pressure sampling positions, the level is determined in an evaluation unit. To determine the densities of liquids and steam, the reactor pressure or the change of pressure with time for transient processes is measured. The density of the comparison column is determined by temperature measurement. The effects of flow are determined by flow measurements. All the measurements are taken to an evaluation unit. (orig./HP)

  16. 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

  17. 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

  18. 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

  19. 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...

  20. 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

  1. 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.

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

    International Nuclear Information System (INIS)

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

  3. The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

    International Nuclear Information System (INIS)

    Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

  4. Cathodic polarization curves of the oxygen reduction reaction on various structural materials of boiling water reactors in high temperature-high purity water

    International Nuclear Information System (INIS)

    Cathodic polarization curves of the O2 reduction reaction were measured by using electrodes made from typical structural materials of boiling water reactors (BWRs) to evaluate the effects of kind of material on the electrochemical corrosion potential (ECP) calculation. To estimate ECPs at any region in the BWRs on the basis of the BWR environmental conditions, anodic and cathodic polarization curves should be obtained in advance under relevant conditions. The concentration of oxidants such as O2 and H2O2 in coolant changes depending on the region in which they exist. As well, reduction reaction rates might differ depending on the kind of materials. In this work, the cathodic polarization curves of type 316L stainless steel (316L SS) and Alloy 182 were measured in high purity water at 553 K with different O2 concentrations and compared with those of type 304 SS (304 SS). The results showed that the cathodic polarization curves differed depending on the kind of materials at the activation-controlled region. But, the difference in the ECP vs. O2 concentration relationship was small when the ECPs were calculated by using both anodic and cathodic polarization curves measured on the objective material. (author)

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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

  8. Investigation of space- and time-dependent processes in the core of a boiling water reactor with a modular computer model

    International Nuclear Information System (INIS)

    A modular, one-dimensional space-dependent model for determination of the transient behavior in the core of a boiling water reactor was developed for use with an available program system for reactor calculations. It consists of the neutron-kinetic module DYNMOD, the thermohydraulic modules DYNSID 1-3 and the module STATEMP for investigation of the stationary initial values. The coupling between the thermohydraulic and the neutron-kinetic processes is ensured by direct manipulation of the effective cross-sections. Two experiments obtained during the testing of the nuclear plant in Wuergassen were computed with this model to gain information on its capability. Satisfactory results were obtained and the influence of space dependence could be studied using the model. (orig.)

  9. Neutronic evaluation of a non-fertile fuel for the disposition of weapons-grade plutonium in a boiling water reactor

    International Nuclear Information System (INIS)

    A new non-fertile, weapons-grade plutonium oxide fuel concept is developed and evaluated for deep burn applications in a boiling water reactor environment using the General Electric 8x8 Advanced Boiling Water Reactor (ABWR) fuel assembly dimensions and pitch. Detailed infinite lattice fuel burnup results and neutronic performance characteristics are given and although preliminary in nature, clearly demonstrate the fuel's potential as an effective means to expedite the disposition of plutonium in existing light water reactors. The new non-fertile fuel concept is an all oxide composition containing plutonia, zirconia, calcia, and erbia having the following design weight percentages: 8.3; 80.4; 9.7; and 1.6. This fuel composition in an infinite fuel lattice operating at linear heat generation rates of 6.0 or 12.0 kW/ft per rod can remain critical for up to 1,200 and 600 Effective Full Power Days (EFPD), respectively, and achieve a burnup of 7.45 x 1020 f/cc. These burnups correspond to a 71--73% total plutonium isotope destruction and a 91--94% destruction of the 239Pu isotope for the 0--40% moderator steam void condition. Total plutonium destruction greater than 73% is possible with a fuel management scheme that allows subcritical fuel assemblies to be driven by adjacent high reactivity assemblies. The fuel exhibits very favorable neutron characteristics from beginning-of-life (BOL) to end-of-life (EOL). Prompt fuel Doppler coefficient of reactivity are negative, with values ranging between -0.4 to -2.0 pcm/K over the temperature range of 900 to 2,200 K. The ABWR fuel lattice remains in an undermoderated condition for both hot operational and cold startup conditions over the entire fuel burnup lifetime

  10. 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

  11. 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)

  12. Contamination of the steam of boiling water reactors because of impurities in the water dissolving in it

    International Nuclear Information System (INIS)

    Raising the parameters at single-circuit nuclear power plants leads to the more intensive passage of various impurities from the water into the steam because of their solubility. Solubility with regard to the contamination of steam by salts must be taken into account at steam pressures above 140 atm [absolute], and with regard to contamination by products of corrosion of structural materials (oxides of iron, aluminum, cobalt, etc.) beginning with 40--60 atm. This document discusses this problem area. Experimental data on the distribution coefficients of the basic contaminants of water are given in this article. 6 refs., 3 figs

  13. 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.

  14. 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

  15. The application of the internal friction damping nondestructive evaluation technique for detecting incipient cracking in critical components, by-pass lines and piping systems in boiling water reactors

    International Nuclear Information System (INIS)

    A technical feasibility program of utilizing an internal friction damping (IFD) nondestructive evaluation (NDE) technique under laboratory and field conditions for piping systems has been initiated. The applicability of the IFD-NDE technique for four inch (10.1 cm) stainless steel by-pass lines and ten inch (25.4 cm) feedwater lines has been established. The overall objectives of the program include: 1. application of the IFD-NDE technique for four inch (10.1 cm) stainless steel lines that simulate by-pass lines, 2. application of the IFD-NDE technique to full size feedwater lines in an on-line boiling water reactor, 3. evaluate the feasibility of utilizing the IFD-NDE technique as an incipient crack detection method for laboratory stress corrosion cracking tests. (orig.)

  16. A Robust Multivariable Feedforward/Feedback Controller Design for Integrated Power Control of Boiling Water Reactor Power Plants

    International Nuclear Information System (INIS)

    In this paper, a methodology for synthesizing a robust multivariable feedforward/feedback control (FF/FBC) strategy is proposed for an integrated control of turbine power, throttle pressure, and reactor water level in a nuclear power plant. In the proposed method, the FBC is synthesized by the robust control approach. The feedforward control, which is generated via nonlinear programming, is added to the robust FBC system to further improve the control performance. The plant uncertainties, including unmodeled dynamics, linearization, and model reduction, are characterized and estimated. The comparisons of simulation responses based on a nonlinear reactor model demonstrate the achievement of the proposed controller with specified performance and endurance under uncertainty. It is also important to note that all input variables are manipulated in an orchestrated manner in response to a single output's setpoint change

  17. Economic simplified boiling water reactor (ESBWR) response to an extended station blackout/ loss of all AC power

    International Nuclear Information System (INIS)

    U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackout for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event

  18. The benefit of hydrogen addition to the boiling water reactor environment on stress corrosion crack initiation and growth in Type 304 stainless steel

    International Nuclear Information System (INIS)

    Integranular stress corrosion cracking (IGSCC) in type 304 stainless steel in high temperature high purity water requires the simultaneous presence of sensitized material, high tensile stress and oxygen. Laboratory and in-reactor stress corrosion test have shown the benefits of adding hydrogen to the boiling water reactor feedwater to reduce the dissolved oxygen concentration and thereby reduce the chemical driving force for IGSCC. The purpose of this program was to verify the benefit of hydrogen additions on the stress corrosion crack behavior. The program investigated the fatigue and constant load crack growth behavior using fracture mechanics specimens in hydrogen water chemistry (HWC). Isothermal heat treatments were used to sensitize type 304 stainless steel. Additionally, full size pipe tests containing actual welds were used to evaluate crack initiation, as well as propagation of cracks, to verify the results of the fracture mechanics tests. These pipe tests were performed under a trapezoidal loading cycle. The results of the small specimen tests show that hwc inhibits IGSCC in type 304 stainless steel. The effects of cyclic loading at slow frequencies which promote IGSCC were also evaluated. Computer aided methods were used in the collection and interpretation of the high temperature crack growth data

  19. 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)

  20. X-ray radiographic experimental investigation of the reference level in hydrostatic level measurement systems for boiling water reactors

    International Nuclear Information System (INIS)

    Hydrostatic level control is of high priority for normal operation safety of BWR-type reactors. For the prevention of undershooting the limiting values precise and secure measuring techniques are indispensable. Actually the filling level is determined form the hydrostatic pressure difference to a reference column. For the secure non-invasive detection of the phase boundary water/steam in inclined tubes the X-ray radiography has been chosen. The experiments were aimed to study possible geometric influences on the water/steam phase boundary. It was shown that the reference filling level is not significantly changed in spite of permanent phase transitions, provided an ideal mechanical construction of the system is given. Future experiments shall be focused on the analysis of interface behavior in case of non-ideal geometries (welds).

  1. 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

  2. The Swr 1000: a nuclear power plant concept with boiling water reactor for maximum safety and economy of operation

    International Nuclear Information System (INIS)

    The SWR 1000 is a design concept for a light water reactor nuclear power plant that meets all requirements regarding plant safety, economic efficiency and environ-mental friendliness. As a result of the plant's safety concept, the occurrence of core damage can, for all practical intents and purposes, be ruled out. If a core melt accident should nevertheless occur, the molten core can be retained inside the RPV, thus ensuring that all consequences of such an accident remain restricted to the plant itself. The power generating costs of the SWR 1000 are lower than with those of coal-fired and combined-cycle power plants. Power generation using nuclear energy does not release carbon dioxide to the environment, thus meeting the need for sustainable protection of our global climate. (author)

  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. 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

  5. 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

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

    International Nuclear Information System (INIS)

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. 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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. 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)

  11. 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

  12. 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.

  13. Development of fast algorithms for estimating stress corrosion crack growth rate in sensitized stainless steels in boiling water reactor coolant environments

    International Nuclear Information System (INIS)

    A simplified method is proposed for modeling the chemistry and potential distribution in a stress corrosion crack in sensitized stainless steel in boiling water reactor (BWR) coolant environments. The model is based on an assumption that only those species that are present at the largest concentrations in the crevice determine the potential distribution down the crack. The advantage of this method is that it permits simplification of the mathematics and allows predictions to be made of the potential and concentration distributions without knowing various parameters, such as the equilibrium constants for homogeneous chemical reactions in the cavity and diffusion coefficients of species that are present at relatively low concentrations near the crack tip. In some important cases, analytical expressions can be obtained for the pH, potential near the crack tip, and crack propagation rate. The conditions for which the potential on the crack flanks and in the vicinity of the crack tip coincides with the free corrosion potential in the local environment, and hence for which there exists a balance between rates of the local anodic and cathodic partial processes, are determined. The impact of the potential drop in the external environment on the potential and concentration distributions down the crack and on the crack propagation rate is also investigated. Excellent agreement is obtained between calculated and measured crack growth rates

  14. Calculation of the radiation dose from an upset condition in the off-gas system for a boiling water power reactor

    International Nuclear Information System (INIS)

    A study has been made which considered the upset conditions to result in a rupture of the delay line or charcoal adsorber portions of the radioactive off-gas treatment system for a boiling water power reactor. Radiation dose calculations were made for an individual at a 300-meter boundary fence. The doses calculated were the whole body immersion dose and the thyroid, bone and lung doses due to inhalation. The relationship between the various operating and upset parameters of the off-gas system and the radiation doses were investigated. A semi-infinite cloud model with a ground level release was assumed. For a delay line rupture, the calculated gamma immersion dose varies from a high of 9 rad for a break at the condenser to a low of 0.2 rad for a break at the maximum end of a 300-minute delay line. The thyroid dose from inhalation of radioiodine was calculated to vary from 3 to 6 millirem for a delay line rupture and to be 0.6 rem for a charcoal bed rupture. The highest gamma immersion dose from a charcoal adsorber bed rupture was calculated to be 1.5 rad for the low flow rate condition with either an ambient or chilled bed system. Curves have been constructed which show the variation of the calculated doses with the various input parameters. (U.S.)

  15. 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)

  16. 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.

  17. 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)

  18. 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.

  19. 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

  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. Post-irradiation examination of the vipac fuel assemblies IFA-104 and IFA-203, irradiated in the Halden boiling water reactor

    International Nuclear Information System (INIS)

    Two seven pin assemblies, IFA-104 and IFA-203, have been irradiated in the Halding Boiling Water Reactor. The Zircaloy clad pins contained vibrationally compacted, sharp edged UO2 particles with smear densities of 84-88% theoretical density and a fuel stack length of about 1520 mm. The IFA-104 pins operated satisfactorily during approximately 240 effective full power days at an average linear heat rating of about 340 W/cm, and a peak rating of about 480 W/cm. Ingress of water via a pressure transducer into one of the IFA-203 pins necessitated the premature termination of the irradiation after approximately 30 effective full power days at an average rating of about 460 W/cm, and a peak rating of about 550 W/cm. One IFA-104 pin and two IFA-203 pins failed early. The primary cause of these failures has been sought in the oxidation reaction of the tantalum tube which protected the W/Re thermocouples with the surrounding UO2 or with moisture in the fuel at temperatures in excess of approximately 1700degC. The fuel centre temperature and the internal gas pressure measurements during the beginning of lifetime and the results of the post-irradiation chemical and microscopic analysis on fuel pin cross-sections have been correlated with corresponding data calculated on a Gapcon-Thermal computer programme written for pellet fuel pins. Fairly good agreement could be achieved between the measured data on the vipac pins and the calculated data of a pellet pin. During the corrosion of the Zircaloy-2 cladding of the IFA-104 pins which were autoclaved in 400degC steam prior to the irradiation, relatively large amounts of H2 entered the cladding from the D2O coolant side. With 0.66 mole% H2O in the D2O the calculated hydrogen uptake preference ratio was 33

  2. RENO-CC: A new system to fuel lattice design in boiling water reactors using neural networks

    International Nuclear Information System (INIS)

    We show a new system to optimize fuel lattices in BWRs named RENO-CC. The system employs a multi state recurrent neural network (MSRNN) for optimizing a fuel lattice pin-by-pin 235U enrichment distribution. Local Power Peaking Factor (LPPF) and k-infinite (k∞) are involved in the MSRNN energy function. Both parameters are calculated by the 2D HELIOS transport code for lattice burn-up. Through the iterative process the MSRNN decreases PPF value while k∞, is kept in a rank of values, at the beginning of lattice life (BOL). The iterative process ends after 20 iterations. If PPF is not lower than limit, RENO-CC applies a fuzzy logic rule in order to recommend if the fuel lattice has an acceptable LPPF value and it might eventually be used in a fuel load. When a fuel lattice is obtained it can be used into a fuel assembly. And eventually, this fuel assembly would be used in the process of fuel load and control rod patterns optimization. So, a 3D core reactor calculation must decide if such a lattice design can fulfill the operation conditions into the reactor core. Preliminary results are shown in this paper. (authors)

  3. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

    International Nuclear Information System (INIS)

    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)

  4. Activity build-up on the circulation loops of boiling water reactors: Basics for modelling of transport and deposition processes

    International Nuclear Information System (INIS)

    In the past 20 years the radiation field of nuclear power plant loops outside the core zone was the object of investigations in many countries. In this context test loops were built and basic research done. At our Institute PSI the installation of a LWR-contamination loop is planned for this year. This experimental loop has the purpose to investigate the complex phenomena of activity deposition from the primary fluid of reactor plants and to formulate analytical models. From the literature the following conclusions can be drawn: The principal correlations of the activity build-up outside the core are known. The plant specific single phenomena as corrosion, crud-transport, activation and deposit of cobalt in the oxide layer are complex and only partially understood. The operational experience of particular plants with low contaminated loops (BWR-recirculation loops) show that in principle the problem is manageable. The reduction of the activity build-up in older plants necessitates a combination of measures to modify the crud balance in the primary circuit. In parallel to the experimental work several simulation models in the form of computer programs were developed. These models have the common feature that they are based on mass balances, in which the exchange of materials and the sedimentation processes are described by global empirical transport coefficients. These models yield satisfactory results and allow parameter studies; the application however is restricted to the particular installation. All programs lack models that describe the thermodynamic and hydrodynamic mechanisms on the surface of deposition layers. Analytical investigations on fouling of process equipment led to models that are also applicable to the activity build-up in reactor loops. Therefore it seems appropriate to combine the nuclear simulation models with the fundamental equations for deposition. 10 refs., 18 figs., 3 tabs

  5. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ...) System, which support functions for alternate water injection during station blackout. ] II. Additional..., was published in the Federal Register on June 15, 2012 (77 FR 36014), for a 60-day public...

  6. 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

  7. 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

    International Nuclear Information System (INIS)

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). 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

  8. Analysis of the Boiling Water Reactor Turbine Trip Benchmark with the Codes DYN3D and ATHLET/DYN3D

    International Nuclear Information System (INIS)

    The OECD/NRC Boiling Water Reactor (BWR) Turbine Trip Benchmark was analyzed by the code DYN3D and the coupled code system ATHLET/DYN3D. For the exercise 2 benchmark calculations with given thermal-hydraulic boundary conditions of the core, the analyses were performed with the core model DYN3D. Concerning the modeling of the BWR core in the DYN3D code, several simplifications and their influence on the results were investigated. The standard calculations with DYN3D were performed with 764 coolant channels (one channel per fuel assembly), the assembly discontinuity factors (ADF), and the phase slip model of Molochnikov. Comparisons were performed with the results obtained by calculations with 33 thermal-hydraulic channels, without the ADF and with the slip model of Zuber and Findlay. It is shown that the influence on core-averaged values of the steady state and the transient is small. Considering local parameters, the influence of the ADF or the reduced number of coolant channels is not negligible. For the calculations of exercise 3, the DYN3D model validated during the exercise 2 calculations in combination with the ATHLET system model, developed at Gesellschaft fuer Anlagen- und Reaktorsicherheit for exercise 1, has been used. Calculations were performed for the basic scenario as well as for all specified extreme versions. They were carried out using a modified version of the external coupling of the codes, the 'parallel' coupling. This coupling shows a stable performance at the low time step sizes necessary for an appropriate description of the feedback during the transient. The influence of assumed failures of different relevant safety systems on the plant and the core behavior was investigated in the calculations of the extreme scenarios. The calculations of exercises 2 and 3 contribute to the validation of DYN3D and ATHLET/DYN3D for BWR systems

  9. Customer-operator partnership. A boiling water reactor developed jointly by AREVA NP and E.ON Kernkraft

    International Nuclear Information System (INIS)

    Many countries spread all over the world have publicly expressed their intention to pursue the construction of new nuclear power plants with improved safety, economy and more straight forwarded operation and maintenance. Reasons for the intention are: The world wide increasing demand for energy and hence the general necessity to build new power plants. The concerns for increased emissions of green house gases leading to a change in the climate have brought into question the primary reliance on plants utilizing fossil fuels. A new reactor type matching the previously stated issues is AREVA NP's further development of proven BWR design. Combining AREVA's and E.ON's expertise, a project was launched to customize the final basic design for this advanced nuclear power plant having a net power output of about 1,250 MW, a net efficiency of about 37% and a design service life of 60 years. Within this joint venture the overall plant design was simplified and additionally all active safety systems have passive safety related backup systems utilizing basic laws of physics, such as gravity, enabling them to function without electrical power supplies or activation by powered instrumentation and control systems. The development takes into account the technical and accumulated operating experience of the project partners. Based on the operating experience of the project partners a simplification of the overall system engineering was performed, flexible fuel cycle length (12 to 24 months) are possible as well as a reduction of process waste was achieved. These improvements regarding the operation and economics result on the one hand in lower investment cost and on the other hand in a high availability of the plant, hence in low maintenance costs. Generally, the electrical generation costs are accomplished, which are competitive to larger-capacity nuclear power plants and fossil-fired plants. (author)

  10. Boiling of subcooled water in forced convection

    International Nuclear Information System (INIS)

    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/cm2), 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)

  11. Wear-out artificial additional aging of ethylene propylene rubber-insulated safety-related cables sampled from boiling water reactor-type containment

    International Nuclear Information System (INIS)

    Additional artificial aging based on Wear-out Approach is applied to ethylene-propylene-diene rubber insulation service used in boiling reactor-type containment for 16 years. The observed aging trend is slower than the one expected from acceleration aging results. Carbonyl formation estimated by infrared spectroscopy supports this result. According to the trajectory of relation between elongation at break and tensile strength, cross-linking seems to be proceeded in service-used samples during the additional aging to mitigate their degradation

  12. 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

  13. 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

  14. Mechanistic modeling for ring-type boiling water reactor fuel spacer design (3) run-off effect and model formulation

    International Nuclear Information System (INIS)

    The fuel spacer is one of the components of a fuel rod bundle and its role is to maintain an appropriate rod-to-rod clearance. The fuel spacer influences the liquid film flow distribution in the fuel rod bundle, so that the spacer geometry has a strong effect on thermal hydraulic characteristics of BWR such as critical power and pressure drop in the fuel bundle. In this paper, liquid film flow characteristics were experimentally investigated in a circular channel with ring-type spacers using air and water as test fluids in order to be compared with the analytical results introducing the mechanistic spacer model. The spacer model was composed of three effects; the drift flow effect, the narrow channel effect and the run-off effect. The drift flow effect and the narrow channel effect were discussed in the previous works and the run-off effect is done in this paper. This paper shows the formulation around the spacer with use of the three effects. The proposed model explains well the experimental results of liquid film flow rates and thickness carried out in reference to the spacer thickness, the gap between the channel wall and the spacer

  15. Investigations on coolant boiling in research reactors. 2

    International Nuclear Information System (INIS)

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

  16. Study of deposited crud composition on fuel surfaces in the environment of hydrogen water chemistry (HWC) of a Boiling Water Reactor at Chinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    This paper aimed at the characterization of metallic composition and surface analysis on the crud of fuel rods for unit-1 of BWR-4 at Nuclear Power Plant. The inductively coupled plasma- atomic emission spectroscopy (ICPAES) and the gamma spectrometry were carried out to analyze the corrosion product distributions and to determine the elemental compositions along the fuel rod under conditions of hydrogen water chemistry (HWC) switched from normal water chemistry (NWC) of reactor coolant in this study. Most of the crud consisted of the flakes and irregular shapes via SEM morphology. The loosely adherent oxide layer was mostly composed of hematite (α- Fe2O3) with amorphous iron oxides by XRD results. The average deposited amounts of crud was the order of 0.5 mg/cm2, indicating that the fuel surface of this plant under HWC environment appeared to be one with the lower crud deposition in terms of low iron level of feedwater. It also showed no significant difference in comparison with NWC condition. (authors)

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

    International Nuclear Information System (INIS)

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

  18. 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)

  19. 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)

  20. 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.)

  1. 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)

  2. 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

  3. Analysis of water cooled reactors stability

    International Nuclear Information System (INIS)

    A model for stability analysis of non-boiling water cooled nuclear system is developed. The model is based on linear reactor kinetics and space averaged heat transfer in reactor and heat-exchanger. The transfer functions are defined and the analysis was applied to nuclear reactor RA at 'Boris Kidric' Institute - Vinca. (author)

  4. 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

  5. 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 ...

  6. Boiled Water Temperature Measurement System Using PIC Microcontroller

    OpenAIRE

    A.T.KARUPPIAH, AZHA. PERIASAMY, P.RAJKUMAR

    2013-01-01

    The measurement system for temperature of boiled water is a critical task in industry. In this paper we designed and implemented a PIC micro controller based boiled water temperature measurement system using PIC 18F452 and national semiconductors LM35 temperature sensor. The designing system is used to measure the tank I boiled water temperature value. If the temperature value reaches the set value high temperature relay board becomes ON to control the solenoid valve. The high temperature of ...

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

    International Nuclear Information System (INIS)

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

  8. 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...

  9. 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

  10. 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].

  11. Experimental investigation on critical heat flux and transition boiling of water flow under increased pressure

    International Nuclear Information System (INIS)

    In connection with reactor safety problems (LOCA) a measuring technique has been developed which enables, within the parameter range of medium pressure (0.11 MPa - 1.20 MPa) and low mass flow densities (10 kg/m2s - 500 kg/m2s), exact experimental investigations of critical heat flux and transition boiling of water under quasi-stationary conditions. The system consists of a vertical, temperature-controlled short test section with water flowing upwards inside; an experimental loop controllable to a large extent; a quick automatic data acquisition, and numeric evaluation procedures. Quasi-stationary measured boiling curves, from nucleate boiling to film boiling (circa 450deg C), demonstrate the importance of pressure, mass flow density, and inlet subcooling, the boiling pressure being the most important parameter. The linear course of the boiling curves during transition boiling is remarkable. A frequently suspected hysteresis of the boiling curve could not be detected. The influence of surface effects (contact angle) clearly decreases with increasing pressure. For the empirical correlation of the measured data by means of indices, a statement was chosen that normalizes the heat flux density of transition boiling to the maximum heat flux density at the beginning of the post-CHF range. As a result, the experimental data obtained, and the correlation developed from them, show a better heat transfer in transition boiling than conservatively assumed in general in literature. The temperature-controlled measurements of complete boiling curves supply data for critical heat flow density and the corresponding wall overheating. A comparison with the uncontrolled operation of the test section shows differences of 5-6% only within the range of measurement accuracy of such experiments. (orig.) With 36 figs., 5 tabs

  12. A stochastic study of noise in boiling reactors

    International Nuclear Information System (INIS)

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

  13. Mass exchange calculation in a wall layer when water boiling

    International Nuclear Information System (INIS)

    Physical sense and mass exchange characteristics of liquid near-the-wall layer under boiling conditions were attempted to be stated. Equations of material and thermal balance were used to describe the mass exchange characteristics. Technique to calculate circulation ratio in the near-the-wall layer under boiling of under-heated and saturated water was suggested on the basis of the derived expressions. Comparison results of calculated and experimental data were analyzed for full-scale boiling

  14. Boils

    Science.gov (United States)

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

  15. An interface redesign for the feed-water system of the advanced boiling water reactor in a nuclear power plant in Taiwan

    International Nuclear Information System (INIS)

    A well-designed human-computer interface for the visual display unit in the control room of a complex environment can enhance operator efficiency and, thus, environmental safety. In fact, a cognitive gap often exists between an interface designer and an interface user. Therefore, the issue of the cognitive gap of interface design needs more improvement and investigation. This is an empirical study that presents the application of an ecological interface design (EID) using three cases and demonstrates that an EID framework can support operators in various complex situations. Specifically, it analyzes different levels of automation and emergency condition response at the Lungmen Nuclear Power Plant in Taiwan. A simulated feed-water system was developed involving two interface styles. This study uses the NASA Task Load Index to objectively evaluate the mental workload of the human operators and the Situation Awareness Rating Technique to subjectively assess operator understanding and response, and is a pilot study investigating EID display format use at nuclear power plants in Taiwan. Results suggest the EID-based interface has a remarkable advantage over the original interface in supporting operator performance in the areas of response time and accuracy rate under both normal and emergency situations and provide supporting evidence that an EID-based interface can effectively enhance monitoring tasks in a complex environment. (author)

  16. 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...

  17. Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

    International Nuclear Information System (INIS)

    Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H2 produced, though the trends in H2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

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

    International Nuclear Information System (INIS)

    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. Numerical Analysis of Lead-Bismuth-Water Direct Contact Boiling Heat Transfer

    Science.gov (United States)

    Yamada, Yumi; Takahashi, Minoru

    Direct contact boiling heat transfer of sub-cooled water with lead-bismuth eutectic (Pb-Bi) was investigated for the evaluation of the performance of steam generation in direct contact of feed water with primary Pb-Bi coolant in upper plenum above the core in Pb-Bi-cooled direct contact boiling water fast reactor. An analytical two-fluid model was developed to estimate the heat transfer numerically. Numerical results were compared with experimental ones for verification of the model. The overall volumetric heat transfer coefficient was calculated from heat exchange rate in the chimney. It was confirmed that the calculated results agreed well with the experimental result.

  20. 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)

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Coherent Calculation for Air-Water Flow and Boiling Flow by Using CUPID Code

    International Nuclear Information System (INIS)

    The Korea Atomic Energy Research Institute has been developing a three-dimensional thermal-hydraulic code, called CUPID, which was motivated from practical needs for the realistic simulation of two-phase flows in nuclear reactor components. This paper presents coherent simulation of an air-water flow test and a sub-cooled boiling flow test, and the model implementation of related to them. The closure relations for the air-water flow and sub-cooled boiling flow are turbulence model, interfacial non-drag force, interfacial condensation, wall evaporation model, interfacial area transport equation, and so on

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

    International Nuclear Information System (INIS)

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

  7. Arrangement for operation of a boiling water reactor where after an operating period some control rods are replaced with control rods of a higher control value

    International Nuclear Information System (INIS)

    The invention is concerned with a method for obtaining longer reactor operating cycles and/or higher power outputs from a BWR reactor. At a partial refueling of the reactor, only some of the regulating rods - in the centre of the core - are exchanged for rods of a higher regulating worth than the original regulating rods (10-20% higher). 50-80% of the regulating rods in the centre of the core are replaced, while the other positions are occupied by regulating rods that have already been used during the proceeding operating period. (L.E.)

  8. 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. Final report

    International Nuclear Information System (INIS)

    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), which is a boiling water reactor (BWR), 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 part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean 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

  9. 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. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    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), which is a boiling water reactor (BWR), 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 part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a {open_quotes}green field{close_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.

  10. Effects of lateral separation of oxidic and metallic core debris on the BWR [Boiling Water Reactor] MK I containment drywell floor

    International Nuclear Information System (INIS)

    In evaluating core debris/concrete interactions for a BWR MK I containment design, it has been common practice to assume that at reactor vessel breach, the core debris is homogeneous and of low viscosity so that it is uniformly distributed radially on the drywell floor. In a recent study performed by the NRC-sponsored Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, calculations indicate that at reactor vessel bottom head failure, the debris temperature is such that the metallic components (Zr, Fe, Ni, Cr) are completely molten while the oxidic components (UO2, ZrO2, FeO) are completely frozen. Thus, the frozen oxides are expected to remain within the reactor pedestal while the molten metallic species radially separate from the frozen oxidic species, flow through the opening in the reactor pedestal, and spread over the annular region of the drywell floor between the pedestal and the containment shell. This report assesses the impact on calculated debris gas release and the production and release of fission product-laden aerosols for two different cases of debris distribution: uniform distribution, and the laterally separated case of 95% oxides-5% metals inside the pedestal and 5% oxides-95% metals outside the pedestal. The computer codes used in this assessment are CORCON-MOD 2, MARCON 2.1B and VANESA

  11. 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

  12. Fundamental study on thermo-hydraulics during start-up in natural circulation boiling water reactors. 3. Effects of system pressure on geysering and natural circulation oscillation

    International Nuclear Information System (INIS)

    The authors have been investigating the fundamentals of thermo-hydraulic instabilities which may appear during the start-up in natural circulation BWRs in order to understand their driving mechanisms and to examine the methods preventing their occurrence with an aim of establishing a rational start-up procedure and reactor configuration. In this paper, based on our proposed driving mechanisms of geysering and natural circulation oscillation, the effects of system pressure on their occurrences and features are investigated experimentally. It is made clear that an increase in the system pressure suppresses their occurrences. From the results, a consideration is drawn to understand the differences in the start-ups between thermal natural circulation boilers using fossil fuel and the Dodewaard reactor. Lastly, a rational start-up procedure to prevent their instabilities from occurring is recommended and a new idea of separator is proposed. (author)

  13. 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

  14. 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

  15. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

  16. 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.

  17. 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.

  18. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen

    International Nuclear Information System (INIS)

    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)

  19. Studies to demonstrate the adequacy of testing results of the qualification tests for the actuator of main steam safety relive valves (MSSRV) in an advanced boiling water reactor (ABWR)

    International Nuclear Information System (INIS)

    This paper presents several studies performed to demonstrate that the testing results from the qualification tests for the actuator of the Main Steam Safety Relief Valves (MSSRV; also called SRV in this paper) in GE's Advanced Boiling Water Reactor (ABWR) are in compliance with the qualification guidelines stipulated in the applicable IEEE standards. The safety-related function of the MSSRV is to relieve pressure in order to protect the reactor pressure vessel from over-pressurization condition during normal operation and design basis events. In order to perform this function, the SRV must actuate at a given set pressure while maintaining the pressure and structural integrity of the SRV. The valves are provided with an electro-pneumatic actuator assembly that opens the valve upon receipt of an automatic or manually initiated electric signal to allow depressurization of the reactor pressure vessel (RPV). To assure the SRV can perform its intended safety related functions properly, qualification tests are needed in addition to analysis, to demonstrate that the SRV can withstand the specified environmental, dynamic and seismic design basis conditions without impairing its safety related function throughout their installed life under the design conditions including postulated design basis events such as OBE loads and Faulted (SSE) events. The guidelines used for the test methods, procedures and acceptance criteria for the qualification tests are established in IEEE std 344-1987 and IEEE std 382-1985. In the qualification tests, the specimen consists of the actuator, control valve assembly, limit switches, and limit switch support structure. During the functional, dynamic and seismic tests, the test specimen was mounted on a SRV. Qualification of safety related equipment to meet the guidelines of the IEEE standards is typically a two-step process: 1) environmental aging and 2) design basis events qualification. The purpose of the first step is to put the equipment in an

  20. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator