WorldWideScience

Sample records for boiling nuclear superheater reactor

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

    International Nuclear Information System (INIS)

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

  2. Study made of corrosion resistance of stainless steel and nickel alloys in nuclear reactor superheaters

    Science.gov (United States)

    Greenberg, S.; Hart, R. K.; Lee, R. H.; Ruther, W. E.; Schlueter, R. R.

    1967-01-01

    Experiments performed under conditions found in nuclear reactor superheaters determine the corrosion rate of stainless steel and nickel alloys used in them. Electropolishing was the primary surface treatment before the corrosion test. Corrosion is determined by weight loss of specimens after defilming.

  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. Effect of superheat and electric field on saturated film boiling

    Science.gov (United States)

    Pandey, Vinod; Biswas, Gautam; Dalal, Amaresh

    2016-05-01

    The objective of this investigation is to study the influence of superheat temperature and applied uniform electric field across the liquid-vapor interface during film boiling using a coupled level set and volume of fluid algorithm. The hydrodynamics of bubble growth, detachment, and its morphological variation with electrohydrodynamic forces are studied considering the medium to be incompressible, viscous, and perfectly dielectric at near critical pressure. The transition in interfacial instability behavior occurs with increase in superheat, the bubble release being periodic both in space and time. Discrete bubble growth occurs at a smaller superheat whereas vapor columns form at the higher superheat values. Destabilization of interfacial motion due to applied electric field results in decrease in bubble separation distance and increase in bubble release rate culminating in enhanced heat transfer rate. A comparison of maximum bubble height owing to application of different intensities of electric field is performed at a smaller superheat. The change in dynamics of bubble growth due to increasing superheat at a high intensity of electric field is studied. The effect of increasing intensity of electric field on the heat transfer rate at different superheats is determined. The boiling characteristic is found to be influenced significantly only above a minimum critical intensity of the electric field.

  5. 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

  6. 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

  7. 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

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

    International Nuclear Information System (INIS)

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

  9. Dynamic simulation of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

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

  10. Experimental research on heat transfer to liquid sodium and its incipient boiling wall superheat in an annulus

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Liquid sodium is mainly used as a cooling fluid in the liquid metal fast breeder reactor (LMFBR), whose heat transfer, whether convective heat transfer or boiling heat transfer, is different from that of water. So it is important for both normal and accidental operations of LMFBR to perform experimental research on heat transfer to liquid sodium and its boiling heat transfer. This study deals with heat transfer with high temperature (300-700℃) and low Pe number (20~70) and heat transfer with low temperature (250~270℃) and high Pe number (125~860), and its incipient boiling wall superheat in an annulus. Research on heat transfer involves theoretical research and experiments on heat transfer to liquid sodium. It also focuses on the theoretical analysis and experimental research on its incipient boiling wall superheat at positive pressure in an annulus. Semiempirical correlations were obtained and they were well coincident with the experimental data.

  11. 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...

  12. Determining wall superheat under fully developed nucleate boiling in plate-type research reactor cores with low-velocity upward flows

    International Nuclear Information System (INIS)

    This paper reports on plate-type research reactor cores that have involute or rectangular coolant channels with channel gap size in the range 2 ≤ d ≤ 5 mm. Heat transfer under fully developed nucleate boiling (FDNB) and low-velocity (sat) in thin channels under FDNB decreases with increasing probability of bubble contact, ΔTsat is a function of the bubble departure diameter Db as well as d, and ΔTsat can be significantly overestimated by the FDNB correlations that are conventionally used in plate-type research reactor analysis but that are based on higher pressure and larger d flow data and that predict ΔTsat as a function of local channel heat flux and pressure only (e.g., as in the Jens- Lottes and Thom correlations). A new FDNB correlation is proposed that represents the bubble contact mechanism through the dimensionless number (d -cDb)/d, where c is a fitting parameter that accounts for the statistical aspects of bubble formation and contact. The ΔTsat predictions of the new correlation agree with the experimental data to within 16% and approach those obtained from the Jens-Lottes correlation with decreasing Db/d

  13. 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)

  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. 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.)

  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. 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...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

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

  2. 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

  3. 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)

  4. Replacement superheaters and reheaters for the prototype fast reactor

    International Nuclear Information System (INIS)

    The Prototype Fast Reactor (PFR) at Dounreay first went into service in 1974. Heat from the reactor core is transferred to three identical liquid sodium circuits which in turn are coupled to three steam generator circuits. As designed, each of the three steam generator circuits comprises an evaporator designed to produce steam at a pressure of 17.85 MN/m2, a superheater and a reheater, each capable of raising the steam to a temperature of 5400C. The steam is used to drive a conventional turbo-alternator capable of generating 250 MW(e). During the early life of the plant, two leaks developed in tube-to-tubeplate welds in two of the superheaters and in one of the reheaters. The two superheaters were repaired and returned to service by explosive plugging of the leaking tubes, but in the case of the reheater, the tubeplate was so badly cracked by caustic products formed by the ingress of steam to the sodium that the reheater was unsuitable for further use. Because of this unexpected failure mechanism it was considered prudent to provide a set of spare units, so in 1979 six replacement tube bundles were ordered, three superheaters and three reheaters. The design, installation and testing of these are described. (author)

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

    Science.gov (United States)

    Mueller, Roland Guenther

    1987-06-01

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

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

    International Nuclear Information System (INIS)

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

  7. 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

  8. 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)

  9. 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

  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. Design and fabrication of steam generators (superheaters) for the prototype fast breeder reactor 'MONJU'

    International Nuclear Information System (INIS)

    In liquid metal-cooled fast breeder reactors, steam generators are one of the important equipments, and emphasis has been placed on their development in various countries in the world. Also in Japan, centering around the Power Reactor and Nuclear Fuel Development Corp., the research and development in the wide range from the fundamentals on heat transfer and flow, materials and strength for steam generators to the manufacture, operation and various tests of large mock-ups including a 50 MW steam generator have been carried out. Further, as for the manufacture and inspection, the improvement of the method of welding tubes and tube plates, the adoption of a fine focus X-ray inspection apparatus and others were carried out. Moreover, as the maintenance technique, the ultrasonic flaw detection probes for the heating tubes were developed. The steam generators (superheaters) for the FBR 'Monju' power station are the heat exchangers of helical coil tube-shell type using SUS 321 steel as the heating tube material. Based on the results of these research and development, the design and manufacture of these superheaters and their installation in the reactor auxiliary building of the FBR 'Monju' power station were completed. The outline of the design, the research and development and the manufacture of the steam generators (superheaters) are reported. (K.I.)

  12. 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.)

  13. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  14. 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

  15. 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)

  16. 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)

  17. 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.

  18. 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.)

  19. 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...

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

    International Nuclear Information System (INIS)

    The work includes validation/testing of existing neutron noise methods under well-controlled circumstances, investigation of boiling phenomena in narrow channels, and development of a novel boiling monitoring method. The work has been performed in the NIOBE facility at the HDR. Noise signals of thermocouples in the channel wall are used for velocity profile monitoring. Flow patterns in the boiling coolant are identified by means of analysis of probaof probability density functions and neutron noise spectra. Local noise effects are studied. (DG)

  1. 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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kozma, R.

    1992-05-04

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

  4. 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)....

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

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

  8. 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

  9. 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

  10. Nuclear reactor plant

    International Nuclear Information System (INIS)

    The plant consists mainly of a steam-raising unit and a steam turbine with high pressure and low pressure stages. There is at least one superheater or intermediate superheater between the steam-raising unit and the low pressure stage. In order to improve the plant efficiency, a high temperature reactor is provided as a source of heat for the superheater or intermediate superheater, which supplies the superheater heat with an energy efficiency of over 60%. This increases the overall net efficiency from 33% to over 36%. (orig.)

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. The 25 MW Super Near Boiling nuclear reactor (SNB25) for supplying co-generation energy to an Arctic Canadian Forces Base

    International Nuclear Information System (INIS)

    Nuclear energy represents a better alternative for the supply of heat and electricity to the Canadian Forces bases in the Arctic (CFS Alert and CFB Nanisivik). In this context, the Super Near-Boiling 25-MWth reactor (SNB25) has been designed as a small unpressurized LWR that displays inherent safety and is intended to run in automatic mode. The reactor employs TRISO fuel particles (20% enrichment) in zirconium-sheathed fuel rods, and is light water cooled and moderated with a normal output temperature is 95 o C at atmospheric pressure. Control is via 133 control rods and six adjustable radial reflector plates. The design work used the probabilistic simulation code MCNP 5 and the deterministic code WIMS-AECL Version 3.1, permitting a code-to-code comparison of the results. Inherent safety was confirmed and is mostly due to the large negative void reactivity coefficient of -5.17 mk per % void. A kinetic model that includes thermal-hydraulics calculations was developed to determine the reactor's behaviour in transient states, and the results further confirm the inherent safety. Large power excursions temperatures that could compromise structural integrity cannot be produced. If the coolant/moderator temperature exceeds the saturation temperature of 100 o C, the coolant begins to boil and the large negative void coefficient causes the reactor to become subcritical in 0.84 seconds. The SNB25 reactor's core life exceeds 12 years between refuellings. A group of 4 SNB25 reactors meets both the heating and electricity requirements of a base like CFB Nanisivik via a hot water network and through an organic Rankine cycle conversion plant. (author)

  19. 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.

  20. Irradiation of Superheater Test Fuel Elements in the Steam Loop of the R2 Reactor

    International Nuclear Information System (INIS)

    The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of ∼ 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO2 pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation (∼ 750 deg C) and one long time irradiation (∼ 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements

  1. 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.)

  2. Self-Sustaining Thorium Boiling Water Reactors

    OpenAIRE

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

    2012-01-01

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

  3. Advanced steam cycles for light water reactors. Final report

    International Nuclear Information System (INIS)

    An appraisal of the potential of adding superheat to improve the overall LWR plant cycle performance is presented. The study assesses the economic and technical problems associated with the addition of approximately 5000F of superheat to raise the steam temperature to 10000F. The practicality of adding either nuclear or fossil superheat to LWR's is reviewed. The General Electric Company Boiling Water Reactor (BWR) model 238-732 (BWR/6) is chosen as the LWR starting point for this evaluation. The steam conditions of BWR/6 are representative of LWR's. The results of the fossil superheat portion of the evaluation are considered directly applicable to all LWR's. In spite of the potential of a nuclear superheater to provide a substantial boost to the LWR cycle efficiency, nuclear superheat offers little promise of development at this time. There are difficult technical problems to resolve in the areas of superheat fuel design and emergency core cooling. The absence of a developed high integrity, high temperature fuel for operation in the steam/water environment is fundamental to this conclusion. Fossil superheat offers the potential opportunity to utilize fossil fuel supplies more efficiently than in any other mode of central station power generation presently available. Fossil superheat topping cycles evaluated included atmospheric fluidized beds (AFB), pressurized fluidized beds, pressurized furnaces, conventional furnaces, and combined gas/steam turbine cycles. The use of an AFB is proposed as the preferred superheat furnace. Fossil superheat provides a cycle efficiency improvement for the LWR of two percentage points, reduces heat rejection by 15 percent per kWe generated, increases plant electrical output by 54 percent, and burns coal with an incremental net efficiency of approximately 40 percent. This compares with a net efficiency of 36--37 percent which might be achieved with an all-fluidized bed fossil superheat plant design

  4. 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.)

  5. 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)

  6. 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

  7. 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)

  8. 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)

  9. 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)

  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. 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)

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. 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

  4. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  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. 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)

  7. 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.

  8. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

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

    International Nuclear Information System (INIS)

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

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

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

    International Nuclear Information System (INIS)

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

  14. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  15. 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)

  16. 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)

  17. 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.)

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  19. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-02-22

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

  20. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-03-09

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

  1. 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

  2. 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 ...

  3. 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

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  5. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US 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 open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes 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 open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two 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 US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  6. Further development of the water separator - intermediate superheater for nuclear power stations with LWR's

    International Nuclear Information System (INIS)

    A superheater with two heating steam stages which is optimised and favourable as regards heat consumption, and which has increased heat transmission surfaces is introduced. These large surfaces can be accommodated without problems with the described, horizontal separator - superheater design. (RW)

  7. 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...

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

    Science.gov (United States)

    Sorensen, Kirk; Juhasz, Albert

    2007-01-01

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

  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. 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...

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  16. Microwave superheaters for fusion

    International Nuclear Information System (INIS)

    The microwave superheater uses the synchrotron radiation from a thermonuclear plasma to heat gas seeded with an alkali metal to temperatures far above the temperature of material walls. It can improve the efficiency of the Compact Fusion Advanced Rankine (CFAR) cycle described elsewhere in these proceedings. For a proof-of-principle experiment using helium, calculations show that a gas superheat ΔT of 20000K is possible when the wall temperature is maintained at 10000K. The concept can be scaled to reactor grade systems. Because of the need for synchrotron radiation, the microwave superheater is best suited for use with plasmas burning an advanced fuel such as D-3He. 5 refs

  17. 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

  18. 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)

  19. Nuclear research reactors

    International Nuclear Information System (INIS)

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  1. 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)

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

    Science.gov (United States)

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

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  3. 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.

  4. 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

  5. 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)

  6. 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)

  7. 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.

  8. 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.

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

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

    International Nuclear Information System (INIS)

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

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  13. 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)

  14. 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 ...

  15. 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

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  17. 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.

  18. 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

  19. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

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

    International Nuclear Information System (INIS)

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

  1. 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. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: fundamental aspects of two-phase flow and boiling heat transfer

    International Nuclear Information System (INIS)

    Separate abstracts are included for each of the papers presented concerning critical flow of two-phase mixtures; two-phase flow instrumentation; critical heat flux and effects of local disturbances; heat transfer and rewetting during reflood; hydrodynamic mechanisms in boiling heat transfer; and entrainment and droplet deposition in two-phase flow. Five papers have been previously abstracted and input to the data base

  4. 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)

  5. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-01

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

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

    International Nuclear Information System (INIS)

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

  9. 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)

  10. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  11. 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.)

  12. 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

  13. 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

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

    Science.gov (United States)

    2013-07-31

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

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

    Science.gov (United States)

    2012-05-08

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

  17. 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

  18. 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

  19. 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

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

    International Nuclear Information System (INIS)

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

  1. 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.

  2. Nuclear reactor theory

    International Nuclear Information System (INIS)

    This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

  3. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Widener, S.K.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

  7. Boiling water reactor off-gas systems evaluation

    International Nuclear Information System (INIS)

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

  8. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  9. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  10. 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...

  11. 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.

  12. 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

  13. 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

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

    International Nuclear Information System (INIS)

    At present RDIPE is close to 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)

  15. 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.

  16. 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

  17. 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.

  18. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  19. 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

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

    OpenAIRE

    Schmitz-Feuerhake, I; Dannheim, B; Heimers, A; Oberheitmann, B.; Schröder, H; Ziggel, H

    1997-01-01

    Exceptional elevation of children's leukemia appearing 5 years after the 1983 startup of the Krümmel nuclear power plant, accompanied by a significant increase of adult leukemia cases, led to investigations of radiation exposures of the population living near the plant. The rate of dicentric chromosomes in peripheral lymphocytes of seven parents of children with leukemia and in 14 other inhabitants near the plant was significantly elevated and indicated ongoing exposures over the years of its...

  1. 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

  2. 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)

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  8. 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

  9. 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

  10. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  11. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  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. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  15. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  16. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  17. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  18. 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)

  19. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  20. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  1. 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

  2. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  3. 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)

  4. 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)

  5. 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.

  6. 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

  7. Electrochemical sensors for application to boiling water reactors

    International Nuclear Information System (INIS)

    An effective measure in combating the intergranular cracking of stainless steel in Boiling Water Reactors (BWRs) is the control of the electrochemical corrosion potential (ECP). It has been found that when the ECP of austenitic stainless steel alloys susceptible to cracking is decreased below -0.230 V(SHE) cracks will not initiate. Similarly, the decrease in potential 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

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  13. SNAP Nuclear Space Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R

    1966-01-01

    This booklet describes the principles of nuclear-reactor space power plants and shows how they will contribute to the exploration and use of space. It compares them with chemical fuels, solar cells, and systems using energy from radioisotopes. The SNAP (Systems for Nuclear Auxiliary Power) Program, begun in 1955, is described.

  14. 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.

  15. 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

  16. 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

  17. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  18. 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

  19. Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

    International Nuclear Information System (INIS)

    In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

  20. Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M., E-mail: amir@cdtn.br, E-mail: hcr@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br, E-mail: campolina@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/UFMG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel Artur P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

  1. Visualization of pool boiling on downward-facing convex surfaces

    International Nuclear Information System (INIS)

    Visualizations and quenching experiments were performed to investigate effect of material properties on pool boiling from downward-facing, convex stainless steel and copper surfaces in saturated water. Video images showed that more than one boiling regimes can co-exist on the surface. Maximum heat flux (MHF) occurred first at lowermost position, then propagated radially outward to higher inclination positions and its local value decreased with increased inclination. However, the wall superheats corresponding to MHF were independent of the local surface inclinations. MHF propagated ∼10 times slower on stainless-steel than on copper and was ∼12% and 40% lower on stainless-steel than on copper at θ = 0 degree and θ 7.91 degree, respectively. Results confirmed that transition boiling consisted of two distinct regions: high wall superheat, in which heat flux increased relatively slowly, and low wall superheat, in which heat flux increased precipitously with time. Nuclear boiling regime also consisted of two distinct regions: high heat flux nucleate boiling, in which heat flux decreased with increased inclination, and low heat flux nucleate boiling, in which heat flux increased with increased inclination

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author)

  7. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  8. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  9. 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.)

  10. 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

  11. Integral nuclear reactor

    International Nuclear Information System (INIS)

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  12. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  13. 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

  14. 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)

  15. 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

  16. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  17. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  18. 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

  19. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  20. 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

  1. 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.

  2. 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

  3. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  4. 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.

  5. 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

  6. 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)

  7. 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.)

  8. Film boiling of R-11 on liquid metal surfaces

    International Nuclear Information System (INIS)

    An interesting problem is the effect of an immiscible liquid heating surface on the process of film boiling. Such surfaces raise questions concerning interface stability to disturbances, effects of gas bubbling, and vapor explosions in layered systems. The specific motivation for this study was to investigate film boiling from a liquid surface with application to cooling of molten reactor core debris by an overlying pool of reactor coolant. To investigate this phenomenon, and apparatus consisting of a nominal six-inch diameter steel vessel to hold the liquid metal and boiling fluid was constructed; coolant reservoirs, heaters, controllers, and allied instrumentation were attached. A transient energy balance was performed on the liquid metal pool by a submerged assembly of microthermocouples in the liquid metal and an array of thermocouples on the wall of the test vessel. The thermocouple data were used to determine the boiling heat flux as well as the boiling superheat. On an average basis, the deviation between the prediction of the Berenson model and the experimental data was less than one percent when Berenson was corrected for thermal radiation effects. Evidence from visualization tests of R-11 in film boiling over molten metal pools to superheats in excess of 600 K supports this conclusion. 13 refs

  9. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  10. Progress in the Development of Compressible, Multiphase Flow Modeling Capability for Nuclear Reactor Flow Applications

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Berry; R. Saurel; F. Petitpas; E. Daniel; O. Le Metayer; S. Gavrilyuk; N. Dovetta

    2008-10-01

    In nuclear reactor safety and optimization there are key issues that rely on in-depth understanding of basic two-phase flow phenomena with heat and mass transfer. Within the context of multiphase flows, two bubble-dynamic phenomena – boiling (heterogeneous) and flashing or cavitation (homogeneous boiling), with bubble collapse, are technologically very important to nuclear reactor systems. The main difference between boiling and flashing is that bubble growth (and collapse) in boiling is inhibited by limitations on the heat transfer at the interface, whereas bubble growth (and collapse) in flashing is limited primarily by inertial effects in the surrounding liquid. The flashing process tends to be far more explosive (and implosive), and is more violent and damaging (at least in the near term) than the bubble dynamics of boiling. However, other problematic phenomena, such as crud deposition, appear to be intimately connecting with the boiling process. In reality, these two processes share many details.

  11. 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

  12. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  13. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  14. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    An improvement in the construction of liquid metal cooled nuclear reactors of the kind in which the fuel assembly is submerged in a pool of coolant contained by a primary vessel housed in a concrete vault, is described. In this modification the roof of the vault carries heat exchangers immersed in the pool of coolant, the lower ends of which are hydraulically damped against oscillation caused by seismic disturbances. (U.K.)

  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. 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.)

  17. 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

  18. 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)

  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. 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

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. 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)

  4. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  5. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    In the middle of 21st century, the population on the earth is expected to double, and the energy that mankind consumes to triple. The nuclear fusion which is said the ultimate energy source for mankind is expected to solve this energy problem. As for fusion reactors, fuel materials exist inexhaustibly, distributing evenly, they have high safety in principle, the product of burning is harmless nonradioactive substance that does not require the treatment and disposal, and the attenuation of induced radioactivity due to neutrons is quick and the effect to global environment is little. The basic plan of second stage nuclear fusion research and development was decided in 1975, aiming at attaining the critical plasma condition. JT-60 has attained it in 1987. The project of international thermonuclear fusion experimental reactor (ITER) was started, and the conceptual design was carried out. Under such background, the third stage basic plan was decided in 1992, and its objective is self ignition condition, long time burning and the basis of the reactor engineering technology. The engineering design of the ITER is investigated. (K.I.)

  6. Experimental investigation of the formation of critical boiling nuclei in superheated sodium

    International Nuclear Information System (INIS)

    This work deals with the problem of the growth of boiling nuclei up to critical size in superheated sodium. Contrary to previous investigations the sodium was superheated to a certain degree without initiating the onset of boiling. The lifetime of the superheated state - defined as the waiting time - and thereby the time dependence of the growth of boiling nuclei was measured. The experiments were performed with stagnant sodium in an almost gas free working apparatus. It was not aspired to simulate real reactor conditions. Sodium temperature and superheat were up to 9600C and 3600C respectively. The purity of the sodium was varied by cold trap purification. (orig./TK)

  7. Nuclear reactor building

    Science.gov (United States)

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  8. 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

  9. 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

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

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

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

  11. 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

  12. 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)

  13. 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 ...

  14. 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)

  15. Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

    International Nuclear Information System (INIS)

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

  16. Measurement in nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor construction has a flux detector comprising a bundle of fibre optics each having a bead incorporating a substance which scintillates on being struck by neutrons or gamma radiations. The other ends of the fibre optics terminate at an image intensifier. The optical fibres may be of glass made from a mixture of silica, alkaline earth metal oxide, cerous oxide and alkali metal oxide. The beads may be incorporated in a disc forming a detector head, which is in a protective guide tube, through which an inert gas may be passed. (author)

  17. 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.

  18. 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

  19. 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.

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

    International Nuclear Information System (INIS)

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

  1. Experimental study on boiling transfer characteristics on external cooling for hemispherical downward heating surface supposing IVR-AM

    International Nuclear Information System (INIS)

    The scale-down experiments supposing IVR-AM were made on the pool boiling heat transfer from the hemispherical downward facing heating surface. The boiling phenomena were realized by flooding the preheated downward facing hemispherical vessel with sub-cooled water or saturated water under the atmospheric pressure. The hemispherical vessel supposing the scale-downed pressure vessel of the nuclear reactor was made of SUS304 stainless steel. The molten lead supposing the molten core, which was retained in the vessel and preheated up to about 500 degree centigrade, was used as the heat source and cooled down by flooding the vessel into the water to realize the boiling heart transfer during the quenching process. Experiments were carried out by changing the sub-cooling of the flooding water, the size of the vessel. The observations by using the digital video camera were performed and made clear the special characteristics taking place near the heating surface during the quenching process. The measurements for the temperature distribution, the wall superheat and surface heat flux by using the fifteen thermocouples placed inside of the vessel wall was also carried out to make clear the boiling heat transfer during the quenching process. (author)

  2. Saint Laurent B1 nuclear power plant: Thermal and mechanical behavior of the 01 RES superheater condensates cooler

    Science.gov (United States)

    Mathiot, M.; Fromal, A.

    1982-04-01

    The thermal and mechanical performance of long thin tubes used to recover condensates from a PWR superheater and to raise the temperature of water fed into the steam generator was assessed, to see if temperature gradients liable to damage the tubes occur during load variations. Results show that maximum stress occurs during stable operation at nominal load. Stress decreases in transient regimes. Stress is decreased by modifying the procedure for opening the steam inlet valve.

  3. 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)

  4. Coupled BWR calculations with the numerical nuclear reactor software system

    International Nuclear Information System (INIS)

    The Numerical Nuclear Reactor (NNR) is a software suite for integrated high-fidelity reactor core simulations including neutronic and thermal-hydraulic feedback. Using solution modules with formulations to reflect the multi-dimensional nature of the system, NNR offers a comprehensive core modeling capability with pin-by-pin representation of fuel assemblies and coolant channels. Originally developed for pressurized water reactors, the NNR analysis capabilities have recently been extended for boiling water reactor (BWR) applications as part of EPRI Fuel Reliability Program. The neutronics methodology is extended to treat non-periodic structure of BWR fuel assemblies, and a new Eulerian two-phase CFD boiling heat transfer model has been integrated with the software system. This paper summarizes the experience with, and results of, the first-of-a-kind coupled calculations as demonstration of a fully-integrated, high-fidelity simulation capability for assessment of margin to crud-induced failure from fuel-duty perspective. (authors)

  5. Water Cooled FBNR Nuclear Reactor

    International Nuclear Information System (INIS)

    A new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are developed by the INPRO program of the International Atomic Energy Agency (IAEA). The IAEA is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The Fixed Bed Nuclear Reactor (FBNR) is a small (70 MWe) nuclear reactor that meets all the established requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve as a dual purpose plant to produce simultaneously both electricity and desalinated water thus making it especially suitable to the needs of most of developing countries. FBNR is developed with the support of the IAEA under its program of Small Reactors Without On-Site Refuelling (SRWOSR). The FBNR reactor uses the pressurized water reactor (PWR) technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact. (author)

  6. 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

  7. 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.

  8. 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

  9. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

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

  10. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    International Nuclear Information System (INIS)

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

  11. 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.

  12. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

  13. 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)

  14. 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...

  15. 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 ...

  16. 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)

  17. 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.)

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. 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.

  3. 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.

  4. 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.)

  5. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  6. Nuclear reaction data and nuclear reactors

    International Nuclear Information System (INIS)

    These two volumes contain the lecture notes of the workshop 'Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety', which was held at the Abdus Salam ICTP in the Spring of 2000. The workshop consisted of five weeks of lecture courses followed by practical computer exercises on nuclear data treatment and design of nuclear power systems. The spectrum of topics is wide enough to timely cover the state-of-the-art and the perspectives of this broad field. The first two weeks were devoted to nuclear reaction models and nuclear data evaluation. Nuclear data processing for applications to reactor calculations was the subject of the third week. On the last two weeks reactor physics and on-going projects in nuclear power generation, waste disposal and safety were presented

  7. 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.)

  8. Optimal control of nuclear reactors

    International Nuclear Information System (INIS)

    The modern control theory is applied to the design of control systems for experimental nuclear reactors that do not belong to power reactors, the component forms of optimal control systems for nuclear reactors are demonstrated. The adoption of output quadratic integral criterion and incomplete state feedback technique can make these systems both efficient and economical. Moreover, approximate handling methods are given so as to simplify the calculations in design. In addition, the adoptable reference values of parameters are given in the illustration

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-06

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

  10. 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

  11. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a nuclear reactor fuel assembly comprising fuel elements arranged in a supporting frame composed of two end pieces, one at the top and the other at the bottom, on which are secured the ends of a number of vertical tubes, each end piece comprising a plane bottom on which two series of holes are made for holding the tubes and for the passage of the coolant. According to the invention, the bottom of each end piece is fixed to an internal plate fitted with the same series of holes for holding the tubes and for the fluid to pass through. These holes are of oblong section and are fitted with fixing elements cooperating with corresponding elements for securing these tubes by transversal movement of the inside plate

  12. Nuclear reactor inspection device

    International Nuclear Information System (INIS)

    A typical embodiment of the invention combines a novel cellular end fitting for a nuclear reactor fuel assembly with a new design for a fuel rod end cap and a radiation sensing device probe to provide a means for swiftly and accurately distinguishing sound fuel rods from those rods that have developed leaks. For example, a somewhat thinner than usual fuel rod end cap is accessible through the open cellular structure of the end fitting to permit a hollow metal probe to contact the fuel rod end cap. This direct contact excludes most of the water, metal and other shielding materials from the volume between the interior of the fuel rod and the radiation detector, thereby improving the quality of the fuel rod examination. A bridge and trolley structure for accurately positioning the probe also is described

  13. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

  14. Nuclear reactor measurement system

    International Nuclear Information System (INIS)

    An instrument to detect the temperature and flow-rate of the liquid metal current of a coolant fluid sample from adjacent sub-assemblies of a liquid metal-cooled nuclear reactor is described. It includes three thermocouple hot junctions mounted in series, each intended for exposure to a sample-current from a single sub-assembly, electromagnetic coils being mounted around an induction core which detects variations in the liquid metal flow-rate by deformation of the lines of flux. The instrument may also include a thermocouple to detect the mean temperature of the sample-current of coolant fluid from several sources, the result being that the temperature of the coolant fluid current in a sub-assembly may be inferred from the three temperature readings associated with this sub-assembly

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-08-02

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-07-01

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

  20. 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

  1. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

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

  2. 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

  3. 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

  4. Sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  5. 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)

  6. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-01

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

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

    International Nuclear Information System (INIS)

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

  14. 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

  15. 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)

  16. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  17. Sensitivity and Uncertainty Analysis of Boiling Water Reactor Stability Simulations

    OpenAIRE

    Gajev, Ivan

    2012-01-01

    The best estimate codes are used for licensing of Nuclear Power Plants (NPP), but with conservative assumptions. It is claimed that the uncertainties are covered by the conservatism of the calculation. Nowadays, it is possible to estimate certain parameters using non-conservative data with the complement of uncertainty evaluation, and these calculations can also be used for licensing. As NPPs are applying for power up-rates and life extension, new licensing calculations need to be performed. ...

  18. Automatic control rod programming for boiling water reactors

    International Nuclear Information System (INIS)

    The objective of long-term control rod programming is to develop a sequence of exposure-dependent control rod patterns that assure the safe and efficient depletion of the nuclear fuel for the duration of the cycle. A two step method was effected in the code OCTOPUS to perform this task automatically for the Pennsylvania and Power Light Co.' BWRs. Although the execution of OCTOPUS provides good or satisfactory results, its input and execution mode has been improved by making it more user friendly and automatic. (authors)

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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.

  4. 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

  5. Teaching About Nature's Nuclear Reactors

    CERN Document Server

    Herndon, J M

    2005-01-01

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  6. Safety device for nuclear reactor

    International Nuclear Information System (INIS)

    This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container

  7. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

    Directory of Open Access Journals (Sweden)

    Amir Zacarias Mesquita

    2013-01-01

    Full Text Available In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF. At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout. This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN. It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool.

  8. Analysis on flow characteristic of nuclear heating reactor

    International Nuclear Information System (INIS)

    The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam mass, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and stability of the system, especially at lower pressure, secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit has not reaches its saturation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0 K<ΔT<28 K) there exists three regions for system flow condition, namely, (1) stable two-phase flow, (2) bulk and subcooled boiling unstable flow, (3) subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, is showed in the above inlet subcooling range, and based on it the instability map of the system is given through experiment and calculation. (3 refs., 9 figs.)

  9. Experimental analysis on thermohydraulic characteristic of nuclear heating reactor

    International Nuclear Information System (INIS)

    The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction mass flow rate and stability of the system, especially at lower pressure; secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit does not reach its situation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0K<ΔT<28 K) there exists three regions for system flow condition, namely, stable two-phase flow, bulk and subcooled boiling unstable flow and subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, are shown in the above inlet subcooling range, and based on it the instability map of the system are given through experiment and calculation

  10. 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

  11. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  12. Film boiling on spheres in single- and two-phase flows.

    Energy Technology Data Exchange (ETDEWEB)

    Liu, C.; Theofanous, T. G.

    2000-08-29

    Film boiling on spheres in single- and two-phase flows was studied experimentally and theoretically with an emphasis on establishing the film boiling heat transfer closure law, which is useful in the analysis of nuclear reactor core melt accidents. Systematic experimentation of film boiling on spheres in single-phase water flows was carried out to investigate the effects of liquid subcooling (from 0 to 40 C), liquid velocity (from 0 to 2 m/s), sphere superheat (from 200 to 900 C), sphere diameter (from 6 to 19 mm), and sphere material (stainless steel and brass) on film boiling heat transfer. Based on the experimental data a general film boiling heat transfer correlation is developed. Utilizing a two-phase laminar boundary-layer model for the unseparated front film region and a turbulent eddy model for the separated rear region, a theoretical model was developed to predict the film boiling heat transfer in all single-phase regimes. The film boiling from a sphere in two-phase flows was investigated both in upward two-phase flows (with void fraction from 0.2 to 0.65, water velocity from 0.6 to 3.2 m/s, and steam velocity from 3.0 to 9.0 m/s) and in downward two-phase flows (with void fraction from 0.7 to 0.95, water velocity from 1.9 to 6.5 m/s, and steam velocity from 1.1 to 9.0 m/s). The saturated single-phase heat transfer correlation was found to be applicable to the two-phase film boiling data by making use of the actual water velocity (water phase velocity), and an adjustment factor of (1 - {alpha}){sup 1/4} (with a being the void fraction) for downward flow case only. Slight adjustments of the Reynolds number exponents in the correlation provided an even better interpretation of the two-phase data. Preliminary experiments were also conducted to address the influences of multi-sphere structure on the film boiling heat transfer in single- and two-phase flows.

  13. Film boiling on spheres in single- and two-phase flows. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Liu, C.; Theofanous, T.G.

    1994-12-01

    Film boiling on spheres in single- and two-phase flows was studied experimentally and theoretically with an emphasis on establishing the film boiling heat transfer closure law, which is useful in the analysis of nuclear reactor core melt accidents. Systematic experimentation of film boiling on spheres in single-phase water flows was carried out to investigate the effects of liquid subcooling (from 0 to 40{degrees}C), liquid velocity (from 0 to 2 m/s), sphere superheat (from 200 to 900{degrees}C), sphere diameter (from 6 to 19 mm), and sphere material (stainless steel and brass) on film boiling heat transfer. Based on the experimental data a general film boiling heat transfer correlation is developed. Utilizing a two-phase laminar boundary-layer model for the unseparated front film region and a turbulent eddy model for the separated rear region, a theoretical model was developed to predict the film boiling heat transfer in all single-phase regimes. The film boiling from a sphere in two-phase flows was investigated both in upward two-phase flows (with void fraction from 0.2 to 0.65, water velocity from 0.6 to 3.2 m/s, and steam velocity from 3.0 to 9.0 m/s) and in downward two-phase flows (with void fraction from 0.7 to 0.95, water velocity from 1.9 to 6.5 m/s, and steam velocity from 1.1 to 9.0 m/s). The saturated single-phase heat transfer correlation was found to be applicable to the two-phase film boiling data by making use of the actual water velocity (water phase velocity), and an adjustment factor of (1-{alpha}){sup 1/4} (with {alpha} being the void fraction) for downward flow case only. Slight adjustments of the Reynolds number exponents in the correlation provided an even better interpretation of the two-phase data. Preliminary experiments were also conducted to address the influences of multisphere structure on the film boiling heat transfer in single- and two-phase flows.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

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

    International Nuclear Information System (INIS)

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  17. 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

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

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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

  1. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  2. 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

  3. Taking a fresh Look at boiling heat transfer on the road to improved nuclear economics and efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Baglietto, E.; Pointer, W. D.

    2016-08-01

    In the effort to reinvigorate innovation in the way we design, build, and operate the nuclear power generating stations of today and tomorrow, nothing can be taken for granted. Not even the seemingly familiar physics of boiling water. The Consortium for the Advanced Simulation of Light Water Reactors, or CASL, is focused on the deployment of advanced modeling and simulation capabilities to enable the nuclear industry to reduce uncertainties in the prediction of multi-physics phenomena and continue to improve the performance of todays light water reactors and their fuel. An important part of the CASL mission is the development of a next generation thermal hydraulics simulation capability, integrating the history of engineering models based on based on experimental experience with the computing technology of the future. (Author)

  4. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  5. Steam separator-superheater with drawing of a fraction of the dried steam

    International Nuclear Information System (INIS)

    This invention concerns a vertical separator-superheater of the steam from a high pressure expansion turbine before it is admitted to an expansion turbine at a lower pressure, by heat exchange with steam under a greater pressure, and drawing of a fraction of the dried steam before it is superheated. Such drawing off is necessary in the heat exchange systems of light water nuclear reactors. Its purpose is to provide a separator-superheater that provides an even flow of non superheated steam and a regular distribution of the steam to be superheated to the various superheating bundles, with a significantly uniform temperature of the casing, thereby preventing thermal stresses and ensuring a minimal pressure drop. The vertical separator-superheater of the invention is divided into several vertical sections comprising as from the central area, a separation area of the steam entrained water and a superheater area and at least one other vertical section with only a separation area of the steam entrained water

  6. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  7. A process for superheating steam in a nuclear power station circuit and device for putting in operation this process

    International Nuclear Information System (INIS)

    A process is described for superheating steam in a nuclear power station circuit, comprising a turbine with a high pressure chamber and a low pressure chamber. It consists in superheating the steam between the high and low pressure chambers of the turbine by using as heating fluid water under pressure at vaporisation temperature, directly taken from the recirculation or circulation flow water of the reactor or of the steam generators. The process is adapted to a pressurised water reactor using a once-through steam generator comprising in succession an economiser, a vaporiser and a superheater, the superheating water being taken at the vaporiser intake. It is also adapted for a boiling water reactor, in that the water is taken directly from the reactor vessel and at a suitable level in the recirculation water

  8. The future of nuclear reactors

    International Nuclear Information System (INIS)

    The Atomic Energy Commission Advisory Committee on Reactor Safeguards began work in early 1948 with the firm and unanimous conviction that nuclear power could not survive a significant damaging accident. They as a committee felt that their job was to make reactors so safe that no such event would ever occur. However, ambitious reactor planners did not like all the buts and cautions that the committee was raising. They seemed to delay unduly their setting sail into the brave new world of clean, cheap, safe nuclear energy. The committee was soon nicknamed the Committee on Reactor Prevention. Reactors, of course, represented a tremendous step into the future. To an unprecedented extent, they were based on theory. But the committee did not have the luxury of putting a preliminary model into operation and waiting for difficulties to show up. In assessing new designs and developments, they had to anticipate future difficulties. Their proposals in good part were accepted, but their deep emphasis on safety did not become a part of the program. Today, forty years later, the author still believes both in the need for nuclear reactors and in the need of a thorough-going, pervasive emphasis on their safety. Real, understandable safety can be achieved, and that achievement is the key to our nuclear future. The details he gives are only examples. The need for reactors that are not only safe but obviously safe can be ignored only at our peril

  9. Studies on nuclear reactor design

    International Nuclear Information System (INIS)

    this thesis presents two studies for safety aspects in nuclear reactor design. the fission process that occurs in the reactor core is the most important process for the harmful effect of produced radiation especially neutrons with different energies and gamma radiations for their strong penetrability . so studying the criticality of the fissile materials in the reactor is one of the most important safety aspects for the reactor design, the attenuation of the neutrons and gammas using suitable shielding materials with suitable thicknesses is the second study that is discussed in this thesis

  10. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines

  11. Radionuclides in United States commercial nuclear power reactors

    International Nuclear Information System (INIS)

    In the next ten to twenty years, many of the commercial nuclear power reactors in the United States will be reaching their projected lifetime of forty years. As these power plants are decommissioned, it seems prudent to consider the recycling of structural materials such as stainless steel. Some of these materials and components have become radioactive through either nuclear activation of the elements within the components or surface contamination with radioactivity form the operational activities. In order to understand the problems associated with recycling stainless steel from decommissioned nuclear power reactors, it is necessary to have information on the radionuclides expected on or in the contaminated materials. A study has been conducted of radionuclide contamination information that is available for commercial nuclear power reactors in the United States. There are two types of nuclear power reactors in commercial use in the United States, pressurized water reactors (PWRs) and boiling water reactors (BWRs). Before presenting radionuclide activities information, a brief discussion is given on the major components and operational differences for the PWRs and BWRs. Radionuclide contamination information is presented from 11 PWRs and over 8 BWRs. These data include both the radionuclides within the circulating reactor coolant water as well as radionuclide contamination on and within component parts

  12. Technique of nuclear reactors controls

    International Nuclear Information System (INIS)

    This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author)

  13. 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

  14. 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

  15. Low-power nuclear reactors

    International Nuclear Information System (INIS)

    A brief development history of low-power nuclear reactors is presented in this paper. Nowadays, some countries have plans to build a series of small nuclear power plants (also floating ones) for use in remote regions. Present constructions of such NPP are presented in this paper. (author)

  16. 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.

  17. 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.

  18. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

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

  19. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-08-08

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

  20. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is to

  1. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  2. Design for reactor core safety in nuclear power plants

    International Nuclear Information System (INIS)

    This Guide covers the neutronic, thermal, hydraulic, mechanical, chemical and irradiation considerations important to the safe design of a nuclear reactor core. The Guide applies to the types of thermal neutron reactor power plants that are now in common use and fuelled with oxide fuels: advanced gas cooled reactor (AGR), boiling water reactor (BWR), pressurized heavy water reactor (PHWR) (pressure tube and pressure vessel type) and pressurized water reactor (PWR). It deals with the individual components and systems that make up the core and associated equipment and with design provisions for the safe operation of the core and safe handling of the fuel and other core components. The Guide discusses the reactor vessel internals and the reactivity control and shutdown devices mounted on the vessel. Possible effects on requirements for the reactor coolant, the reactor coolant system and its pressure boundary (including the pressure vessel) are considered only as far as necessary to clarify the interface with the Safety Guide on Reactor Coolant and Associated Systems in Nuclear Power Plants (IAEA Safety Series No. 50-SG-D13) and other Guides. In relation to instrumentation and control systems the guidance is mainly limited to functional requirements

  3. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-06-01

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage

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

    International Nuclear Information System (INIS)

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

  5. nuclear reactor design calculations

    International Nuclear Information System (INIS)

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  6. Nuclear reactor internal structures

    International Nuclear Information System (INIS)

    The upper internal structures of the reactor are connected to the closing head so as to be readily removed with the latter and a skirt connected to the lower portion of said upper structures so as to surround the latter, extends under the control rods when they are removed from the reactor core. Through such an arrangement the skirt protects the control rods and supports the vessel closing-head and the core upper structures, whenever the head is severed from the vessel and put beside the latter in order to discharge the reactor

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  11. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author)

  12. Boiling heat transfer in porous media composed of particles

    International Nuclear Information System (INIS)

    The boiling heat transfer in the porous media composed of spherical fuel elements exerts significant influences on the reactor's efficiency and safety. In the present study an experimental setup was designed and the boiling heat transfer in the porous media composed of spheres of regular distribution was investigated. Four spheres with diameters of 5mm, 6mm, 7mm and 8mm were used in the test sections. The greater particle diameter led to lower heat transfer coefficient, and resulted in higher wall superheat of original nucleation boiling. The variation of heat transfer coefficient was divided into three groups according to two-phase flow patterns and void fraction. A correlation of heat transfer coefficient was proposed with a mean relative deviation of ± 16%. (author)

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  18. 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

  19. 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)

  20. 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)

  1. 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.

  2. 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.

  3. Fixed bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Full text: The fixed bed nuclear reactor (FBNR) is essentially a pressurized light water reactor (PWR) having spherical fuel elements constituting a suspended reactor core at its lowest bed porosity. The core is movable thus under any adverse condition, the fuel elements can leave the reactor core naturally through the force of gravity and fall into the passively cooled fuel chamber or leave the reactor all together entering the spent fuel pool. It is a small and modular reactor being simple in design. Its spent fuel is in such a convenient form and size that may be utilized directly as the source for irradiation and applications in agriculture and industry. This feature results in a positive impact on waste management and environmental protection. The principle features of the proposed reactor are that the concept is polyvalent, simple in design, may operate either as fixed or fluidized bed, have the core suspended contributing to inherent safety, passive cooling features of the reactor. The reactor is modular and has integrated primary system utilizing either water, supercritical steam or helium gas as its coolant. Some of the advantages of the proposed reactor are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. The characteristics of the Fluidized Bed Nuclear Reactor (FBNR) concept may be analyzed under the light of the requirements set for the IV generation nuclear reactors. It is shown that FBNR meet the goals of (1) Providing sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production, (2) Minimize and manage their nuclear waste and notably reduce the long term stewardship burden in the future, thereby improving protection for the public health and the environment, (3) Excel in safety and reliability

  4. Innovative designs of nuclear reactors

    International Nuclear Information System (INIS)

    The world development scenarios predict at least a 2.5 time increase in the global consumption of primary energy in the first half of the twenty-first century. Much of this growth can be provided by the nuclear power which possesses important advantages over other energy technologies. However, the large deployment of nuclear sources may take place only when the new generation of reactors appears on the market and will be free of the shortcomings found in the existing nuclear power installations. The public will be more inclined to accept nuclear plants that have better economics; higher safety; more efficient management of the radioactive waste; lower risk of nuclear weapons proliferation, and provided that the focus is made on the energy option free of ∇e2 generation. Currently, the future of nuclear power is trusted to the technology based on fast reactors and closed fuel cycle. The latter implies reprocessing of the spent nuclear fuel of the nuclear plants and re-use of plutonium produced in power reactors

  5. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-02

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

  8. Neutrino physics with nuclear reactors

    International Nuclear Information System (INIS)

    This is a lecture given at the Gif Summer School held in 1992 in Montpellier. It contains three chapters. These are devoted to neutrino oscillations, to the nuclear reactors used as neutrino sources, and to the experiments performed with neutrinos from nuclear reactors, respectively. The first chapter offers a theoretical frame, the second discusses the investigation capabilities of nuclear reactors as neutrino sources while the last one describes the experimental aspects. These aspects are related to the neutrino flux measurement and the flavor oscillation, the search for neutrino oscillation, the neutrino scattering on electrons, the neutrino decay, the coherent neutrino scattering on nuclei and the electron neutrino-electron antineutrino oscillations implied by the Majorana nature of neutrinos. In concluding the author points to the possible ways of refining these extremely subtle experiments, which will be approached in the near future. 117 refs., 9 figs., 11 tabs

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

    CERN Document Server

    Kessler, Günter; Schlüter, Franz-Hermann

    2014-01-01

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

  10. Nuclear reactor downcomer flow deflector

    Science.gov (United States)

    Gilmore, Charles B.; Altman, David A.; Singleton, Norman R.

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  11. 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.

  12. 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

  13. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. 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.

  18. Nuclear reactor container

    International Nuclear Information System (INIS)

    In a container of a BWR type reactor, spray water is stored in a pedestal cavity. A perforated hole is formed on the side wall of the pedestal, and a stirrer is disposed in the pedestal cavity to stir the stored spray water. During reactor operation, the door on the side wall of the pedestal is closed to prevent discharge of fission products to the dry well when a severe accident should occur. During periodical inspection for the plant, the door is opened to enable an operator to access to the inside of the pedestal. When a molten reactor core should drop to the pedestal cavity, fission products generated from the failed reactor core left in a pressure vessel pass through the spray water in the pedestal cavity. Then, most of the fission products are held in the spray water by a scrubbing effect when they pass through the spray water. In addition, the stored spray water is stirred by the stirrer to enhance the scrubbing effect thereby enabling to further decrease the amount of the fission products discharged to the dry well. (N.H.)

  19. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A method of constructing a radiation shielding plug for use in the roof of the coolant containment vault of liquid metal cooled fast breeder reactors is described. The construction allows relative movement of that part of service cables and pipes which are carried by the fixed roof and that part which is carried by the rotatable plug. (U.K.)

  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. 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)

  2. 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

  3. 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

  4. 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.

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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

  8. Nuclear reactors and disarmament

    International Nuclear Information System (INIS)

    From a brief analysis of the perspectives of nuclear weapons arsenals reduction, a rational use of the energetic potential of the ogives and the authentic destruction of its warlike power is proposed. The fissionable material conversion contained in the nuclear fuel ogives for peaceful uses should be part of the disarmament agreements. This paper pretends to give an approximate idea on the resources re assignation implicancies. (Author)

  9. Nuclear reactor container

    International Nuclear Information System (INIS)

    Upon reactor accident, hydrogen and oxygen are generated by water-zirconium reaction and radiolysis of water, which are accumulated in the reactor. If the concentration of hydrogen and oxygen exceeds a burning limit, there is a possibility of hydrogen burning to cause a danger of deteriorating the integrity of the reactor container and the equipments therein. The limit for the occurrence of the detonation is determined by a relationship between the scale of a detonation cell and the size of the container, and if the scale is greater than the container, the detonation does not occur. The scale of the cell is determined by a gas combustion rate and, if the combustion reaction is suppressed, detonation does not occur even in a large container. Then, an appropriate diluent is added to increase heat capacity of a gas mixture to thereby suppress the temperature elevation of the gas. Incombustible gases having a great heat capacity are preferred for the diluent, and CO2 is used. As the concentration of the CO2 gas to be added is increased, the detonation cell is made greater. Thus, occurrence of detonation due to combustion of the accumulated hydrogen can be prevented. (N.H.)

  10. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work

  11. Nuclear reactor PBMR and cogeneration

    International Nuclear Information System (INIS)

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  12. 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

  13. Health requirements for nuclear reactor operators

    International Nuclear Information System (INIS)

    The health prerequisites established for the qualification of nuclear reactor operators according to CNEN-NE-1.01 Guidelines Licensing of nuclear reactor operators, CNEN-12/79 Resolution, are described. (M.A.)

  14. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  15. Instrumentation for nuclear reactor control

    International Nuclear Information System (INIS)

    This lecture is concerned with engineers and technicians not specialized in nuclear reactor control. The different methods of measurement used are briefly reviewed: current or pulse measurement, and Campbell system; the electronic networks are described and a part is devoted to the cables connecting detectors and electronic assemblies

  16. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  17. Nuclear reactor effluent monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Minns, J.L.; Essig, T.H. [Nuclear Regulatory Commission, Washington, DC (United States)

    1993-12-31

    Radiological environmental monitoring and effluent monitoring at nuclear power plants is important both for normal operations, as well as in the event of an accident. During normal operations, environmental monitoring verifies the effectiveness of in-plant measures for controlling the release of radioactive materials in the plant. Following an accident, it would be an additional mechanism for estimating doses to members of the general public. This paper identifies the U.S. Nuclear Regulatory Commission (NRC) regulatory basis for requiring radiological environmental and effluent monitoring, licensee conditions for effluent and environmental monitoring, NRC independent oversight activities, and NRC`s program results.

  18. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    In the reactor operating with supercritical pressure and temperature part of the water flowing through the moderator tubes is deflected at the outlet and mixed with a residual partial flow of the coolant fed into the core as well as passed along the fuel rods in opposite direction. By special guiding of the flow downward through the guide tubes of the control rods insertion of the control rods is simplified because of reduced frictional forces. By this means it is also achieved to design less critical the control rod cooling with respect to flow rate control and operating behavior in case of a scram. (orig.)

  19. 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

  20. 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.)

  1. Decommissioning of Salaspils nuclear reactor

    International Nuclear Information System (INIS)

    In May 1995, the Latvian Government decided to shut down the Research Reactor Salaspils (SRR) and to dispense with nuclear energy in future. The reactor has been out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH from 1998-1999. he Latvian Government decided on 26 October 1999 to start the direct dismantling to 'green field' in 2001. The results of decommissioning and dismantling performed in 1999-2001 are presented and discussed. The main efforts were devoted to collecting and conditioning 'historical' radioactive waste from different storages outside and inside the reactor hall. All radioactive material more than 20 tons were conditioned in concrete containers for disposal in the radioactive waste depository 'Radons' in the Baldone site. Personal protective and radiation measurement equipment was upgraded significantly. All non-radioactive equipment and material outside the reactor buildings were free-released and dismantled for reuse or conventional disposal. Weakly contaminated material from the reactor hall was collected and removed for free-release measurements. The technology of dismantling of the reactor's systems, i.e. second cooling circuit, zero power reactors and equipment, is discussed in the paper. (author)

  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. 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.

  4. New advanced small and medium nuclear power reactors: possible nuclear power plants for Australia

    International Nuclear Information System (INIS)

    In recent years interest has increased in small and medium sized nuclear power reactors for generating electricity and process heat. This interest has been driven by a desire to reduce capital costs, construction times and interest during construction, service remote sites and ease integration into small grids. The IAEA has recommended that the term 'small' be applied to reactors with a net electrical output less than 300 MWe and the term 'medium' to 300-700 MWe. A large amount of experience has been gained over 50 years in the design, construction and operation of small and medium nuclear power reactors. Historically, 100% of commercial reactors were in these categories in 1951-1960, reducing to 21% in 1991-2000. The technologies involved include pressurised water reactors, boiling water reactors, high temperature gas-cooled reactors, liquid metal reactors and molten salt reactors. Details will be provided of two of the most promising new designs, the South African Pebble Bed Modular Reactor (PBMR) of about 110 MWe, and the IRIS (International Reactor Innovative and Secure) reactor of about 335 MWe. Their construction costs are estimated to be about US$l,000/kWe with a generating cost for the PBMR of about US1.6c/kWh. These costs are lower than estimated for the latest designs of large reactors such as the European Pressurised Reactor (EPR) designed for 1,600 MWe for use in Europe in the next decade. It is concluded that a small or medium nuclear power reactor system built in modules to follow an increasing demand could be attractive for generating low cost electricity in many Australian states and reduce problems arising from air pollution and greenhouse gas emissions from burning fossil fuels

  5. 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

  6. NUCLEAR REACTOR FUEL ELEMENT

    Science.gov (United States)

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  7. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and

  8. Economic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U3O8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented

  9. Nuclear reactor fuelling machine

    International Nuclear Information System (INIS)

    The refuelling machine described comprises a rotatable support structure having a guide tube attached to it by a parellel linkage mechanism, whereby the guide tube can be displaced sideways from the support structure. A gripper unit is housed within the guide tube for gripping the end of a fuel assembly or other reactor component and has means for maintenance in the engaging condition during travel of the unit along the guide tube, except for a small portion of the travel at one end of the guide tube, where the inner surface of the guide tube is shaped so as to maintain the gripper unit in a disengaging condition. The gripper unit has a rotatable head, means for moving it linearly within the guide tube so that a component carried by the unit can be housed in the guide tube, and means for rotating the head of the unit through 1800 relative to its body, to effect rotation of a component carried by the unit. The means for rotating the head of the gripper unit comprises ring and pinion gearing, operable through a series of rotatable shafts interconnected by universal couplings. The reason for provision for 1800 rotation is that due to the variation in the neutron flux across the reactor core the side of a fuel assembly towards the outside of the core will be subjected to a lower neutron flux and therefore will grow less than the side of the fuel assembly towards the inside of the core. This can lead to bowing and possible jamming of the fuel assemblies. Full constructional details are given. See also BP 1112384. (U.K.)

  10. 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

  11. 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

  12. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    International Nuclear Information System (INIS)

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  13. Analyses of single-phase heat transfer and onset of nucleate boiling in a rod bundle with mixing vane grids

    International Nuclear Information System (INIS)

    In the framework of axial offset anomaly risk assessment in Pressurized Water Reactor (PWR) cores, an experimental program involving hydraulic and thermal-hydraulic tests on identical 5x5 bundle geometry was completed. It aimed at developing a consistent set of single-phase heat transfer model and associated onset of nucleate boiling (ONB) wall superheat criterion to further predict the existence and location of boiling zones in a PWR core, using a sub-channel Thermal-Hydraulic (T/H) code. This paper is devoted to the code-based analysis of the experimental data obtained on a bundle equipped with alternating simple support grids and mixing vane grids. Dedicated heat transfer models including a grid enhancement function are developed and the use of Frost & Dzakowic ONB wall superheat criterion is recommended along with these models. (author)

  14. Real time simulation research in 200 MW low temperature nuclear heating reactor core

    International Nuclear Information System (INIS)

    200 MW low temperature nuclear heating reactor is an important new-type reactor. Natural circulation is adopted in the flowage of reactor core. High precise models are built and selected, which are low temperature reactor power model, residual power releasing model, heat conductivity model in reactor core, thermo-hydraulic model, subcooling boiling model, CHF calculation model and so on. These models are solved using Gear arithmetic and Adams arithmetic, which are testified each other. Using appropriate arithmetic, the real time simulation of thermo-hydraulic process in the core is truly fulfilled. (authors)

  15. Liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hydrogen can be added to nuclear reactors with a liquid hydrogen-containing coolant on the suction side of a high pressure pump in the purification system. According to the invention this is performed by means of a liquid jet condenser which uses the coolant as liquid and which is preferably charged from the pressure side of the high pressure pump and conveys the liquid to a mixer connected in series with the high pressure pump. The invention is to be used especially in pressurized water reactors. (orig./PW)

  16. Liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hydrogen can be added to nuclear reactors with a liquid hydrogen-containing coolant on the suction side of a high pressure pump in the purification system. According to the invention this is performed by means of a liquid jet condenser which uses the coolant as liquid and which is preferably charged from the pressure side of the high pressure pump and conveys the liquid to a mixer connected in series with the high pressure pump. The invention is to be used especially in pressurized water reactors. (orig.)

  17. Integrated plant safety assessment: Systematic Evaluation Program. LaCrosse Boiling Water Reactor, Dairyland Power Cooperative, Docket No. 50-409

    International Nuclear Information System (INIS)

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addresed. Equipment and procedural changes have been identified as a result of the review

  18. Integrated plant safety assessment. Systematic Evaluation Program. La Crosse Boiling Water Reactor. Dairyland Power Cooperative, Docket No. 50-409. Final report

    International Nuclear Information System (INIS)

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review

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

    International Nuclear Information System (INIS)

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

  20. 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)

  1. 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)

  2. 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)

  3. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

    OpenAIRE

    Amir Zacarias Mesquita; Rogério Rivail Rodrigues

    2013-01-01

    In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF). At this time,...

  4. Recycling device of nuclear reactor

    International Nuclear Information System (INIS)

    In a recycling device of a nuclear reactor, a coolant recycling system is disposed by using an outer loop, while a branched connection pipe is connected to a feed water jet pump driving system. Further, the coolant recycling system is constituted with a remaining-heat removing system having a heat exchanger. The connection pipe branched from the downstream of the heat exchanger is connected to the suction side of the jet pump driving pump. Even when feedwater is not returned or returned only insufficiently from a condensate/feedwater system, such as in a case of reactor start up, since sufficient jet pump driving water can be ensured, reactor power can be controlled by controlling the reactor core flow rate by the driving water, to improve the operationability. Further, the burden on control rods can be decreased to improve reliability compared with the case of controlling the power only by the operation of the control rods. Further, since the recycling flow rate of coolants in the reactor core can be ensured sufficiently, occurrence of temperature difference between the upper and the lower portions of a pressure vessel can be prevented effectively, to improve reactor integrity. (N.H.)

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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

  8. Nuclear reactor safety

    International Nuclear Information System (INIS)

    Dr. Buhl feels that nuclear-energy issues are too complex to be understood as single topics, and can only be understood in relationship to broader issues. In fact, goals and risks associated with all energy options must be seen as interrelated with other broad issues, and it should be understood that there are presently no clearcut criteria to ensure that the best decisions are made. The technical community is responsible for helping the public to understand the basic incompatibility of hard and soft technologies and that there is no risk-free energy source. Four principles are outlined for assessing the risks of various energy technologies: (1) take a holistic view; (2) compare the risk with the unit energy output; (3) compare the risk with those of everyday activities; and (4) identify unusual risks associated with a particular option. Dr. Buhl refers to the study conducted by Dr. Inhaber of Canada who used this approach and concluded that nuclear power and natural gas have the lowest overall risk

  9. Liquid cooled nuclear reactors

    International Nuclear Information System (INIS)

    A construction is described for a liquid metal cooled fast reactor, in which the core is supported in a pool of liquid coolant, wherein a catchment tray is provided for any debris falling from the core. The tray comprises a complex of open top collecting vessels with central support struts, the vessels being spaced apart and arranged in layers in a lattice pitch. The lattice pitches of the vessels in each layer are off-set to the lattice pitches of the vessels in the other layers, so that upper vessels partially overlap lower vessels, and the support struts extend through interspaces defined by the vessels in off-set pitch to a common supporting sub-structure. The complex of vessels offers a complete catchment area for falling debris, whilst being pervious to liquid coolant circulating upwardly by convection. The collecting vessels preferably comprise conical dishes and are arranged in triangular lattice pitch in each layer, and the complex of vessels comprises three layers. Alternatively the collecting vessels may be rectilinear and arranged on a square lattice. The catchment tray may comprise two or more such complexes in stacked array. (U.K.)

  10. Nuclear reactor container

    International Nuclear Information System (INIS)

    A gas containing vessel has a water pool which is in communication with a dry well containing a reactor pressure vessel by way of a communication pipe is disposed. A capacity of a gas phase portion of the gas containing chamber, a capacity of the dry well, a water depth of a bent tube communicating the dry well with a pressure suppression pool of a pressure suppression chamber and a water depth of the communication pipe are determined so as to satisfy specific conditions. Since the water depth of the communication pipe is less than the water depth of the bent tube, incondensible gases and steams in the dry well flow into the water pool of the gas containing chamber at the initial stage of loss of coolant accident. Subsequently, steams in the dry well flow into the pressure suppression pool of the pressure suppression chamber by way of the bent tube. Accordingly, since the incondensible gases in the dry well do not flow into the pressure suppression chamber, pool swelling phenomenon in the pressure suppression chamber is not caused even if the water depth of the bent tube which leads to the pressure suppression chamber is great. Further, pressure increase due to transfer of the incondensible gases is decreased. (I.N.)

  11. 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

  12. 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

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

    International Nuclear Information System (INIS)

    The demonstration of inservice inspection methods with ultrasound for the 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

  14. 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

  15. 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...

  16. 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)

  17. 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)

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

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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.)

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

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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.)

  11. 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)

  12. 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.

  13. Water cooled FBNR nuclear reactor

    International Nuclear Information System (INIS)

    Full text: The world with its increasing population and the desire for a more equitable and higher standard of living, is in the search for energy that is abundant and does not contribute to the problem of global warming. The answer to this is a new paradigm in nuclear energy; i.e., through the innovative nuclear reactors that meet the IAEA's INPRO philosophies and criteria that will guarantee the generation of safe and clean energy. The emerging countries to nuclear energy that are not in hurry for energy and look into the future are looking into the participation in the development of such innovative nuclear reactors. They can start developing the non-nuclear components of such reactors in parallel with creating the nuclear infra-structures according to the guidelines of the IAEA suggested in its milestones document. In this way, they can benefit from numerous advantages that the development of a high technology can bring to their countries be it scientific, technological, economic or political. A solution to the present world economic crisis is investing in such projects that contribute to the real economy rather than speculative economy. This will help both local and world economy. One such innovative nuclear reactor is the FBNR that is being developed with the support of the IAEA in its program of Small Reactors Without On-site Refuelling. It is a small (70 MWe) reactor with simple design based on the proven PWR technology (www.sefidvash.net/fbnr). The simplicity in design and the world wide existence of water reactor technology, makes it a near term project compared to other future reactors. Small reactors are most adequate for both the developing and developed countries. They require low capital investment, and can be deployed gradually as energy demand calls for. The generation of energy at the local of consumption avoids high cost of energy transmission. The paradigm of economy of scale does not apply to the FBNR as it is a small reactor by its nature. The

  14. Three dimensional diffusion calculations of nuclear reactors

    International Nuclear Information System (INIS)

    This work deals with the three dimensional calculation of nuclear reactors using the code TRITON. The purposes of the work were to perform three-dimensional computations of the core of the Soreq nuclear reactor and of the power reactor ZION and to validate the TRITON code. Possible applications of the TRITON code in Soreq reactor calculations and in power reactor research are suggested. (H.K.)

  15. Production of radionuclides in nuclear reactor

    International Nuclear Information System (INIS)

    Given is a short review on the production of radionuclides which was performed in the Vinca Institute of Nuclear Sciences by using the nuclear reactor RA. Regarding the considerations of the possible re-starting of this reactor its use for the production of medical radionuclides should be taken into account. Listed are some of the important medical radionuclides routinely produced in nuclear reactors in the world and discussed the conditions for their obtaining in the reactor RA. (author)

  16. Computer code for nuclear reactor core thermal reliability calculation

    International Nuclear Information System (INIS)

    RASTENAR program was described for computing heat-engineering reliability of cores in nuclear reactors operating under stationary conditions. The following factors of heat-engineering reliability were found to be computable: rated critical margin; limiting critical margin; probability of initiation of critical heat removal in channel (inferior conditions of heat transfer); probability that no channel would be subject to critical heat removal; and reactor power reserve coefficient. The probability that no channel in the core would experience critical heat removal when boiling during operation of the reactor at fixed power level was taken for the principal quantitative criterion. The structure and limitations of the program were described together with the computation algorithm. The program was written for an M-220 computer

  17. Actinide transmutation in nuclear reactors

    International Nuclear Information System (INIS)

    This report has also been published as a PhD thesis. It discusses the reduction of the transuranics part of nuclear waste. Requirements and criteria for efficient burning of transuranics are developed. It is found that a large reduction of transuranics produced per unit of energy is possible when the losses in reprocessing are small and when special transuranics burner reactors are used at the end of the nuclear era to reduce the transuranics inventory. Two special burner reactors have been studied in this thesis. In chapter 3, the Advanced Liquid Metal Reactor is discussed. A method has been developed to optimize the burning capability while complying to constraints imposed on the design for safety, reliability, and economics. An oxide fueled and metallic fueled ALMR have been compared for safety and transuranics burning. Concluded is that the burning capability is the same, but that the higher thermal conductivity of the metallic fuel has a positive effect on safety. In search for a more effective waste transmuter, a modified Molten Salt Reactor was designed for this study. The continuous refueling capability and the molten salt fuel make a safe design possible without uranium as fuel. A four times faster reduction of the transuranics is possible with this reactor type. The amount of transuranics can be halved every 10 years. The most important conclusion of this work is that it is of utmost importance in the study of waste transmutation that a high burning is obtained with a safe design. In future work, safety should be the highest priority in the design process of burner reactors. (orig.)

  18. Reactors for nuclear electric propulsion

    International Nuclear Information System (INIS)

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades

  19. Nuclear reactors in remote earth

    International Nuclear Information System (INIS)

    Same basic geological principles along with other facts, have allowed us to establish the existence in the remote past (Between 2.5 and 4 x 10''9 years ago) of the uranium deposits and/or uranium mineralized volumes, which be-have as nuclear reactors. A simplified neutronic diffusion model have allowed us to describe the main characteristics of such systems. The obtained results indicate that this phenomenon was a rather frequent fact. (Author) 7 refs

  20. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    The gas temperature of a hot gas loop in gas-cooled nuclear reactor plants shall be able to be modified without influencing the gas temperature of the other loops. If necessary, it should be possible to stop the loop. This is possible by means of a mixer which is places below the heat absorbing component in the hot channel and which is connected to a cold gas line. (orig.)