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Sample records for boiling heavy water cooled and moderated reactor

  1. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors)

  2. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    International Nuclear Information System (INIS)

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    International Nuclear Information System (INIS)

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

  5. Computer code for the analyses of reactivity initiated accident of heavy water moderated and cooled research reactor 'EUREKA-2D'

    International Nuclear Information System (INIS)

    Codes, such as EUREKA and EUREKA-2 have been developed to analyze the reactivity initiated accident for light water reactor. These codes could not be applied directly for the analyses of heavy water moderated and cooled research reactor which are different from light water reactor not only on operation condition but also on reactor kinetic constants. EUREKA-2D which is modified EUREKA-2 is a code for the analyses of reactivity initiated accident of heavy water research reactors. Following items are modified: 1) reactor kinetic constants. 2) thermodynamic properties of coolant. 3) heat transfer equations. The feature of EUREKA-2D and an example of analysis are described in this report. (author)

  6. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  7. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  8. Graphite-moderated and heavy water-moderated spectral shift controlled reactors

    International Nuclear Information System (INIS)

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

  9. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  10. Thorium-Fuelled Heavy-Water-Moderated Organic-Cooled Reactors

    International Nuclear Information System (INIS)

    The HWOCR is a heavy-water-moderated, organic-cooled, pressure-tube reactor similar to the ORGEL reactor concept being developed by Euratom. The performance of thorium fuel cycles in the HWOCR was evaluated and several 1000 MW(e) conceptual designs were developed for a thorium-fuelled HWOCR during an 18-month period from March 1965 to September 1966. The work was conducted in parallel with a much larger design and development effort on the uranium-fuelled concept and was restricted primarily to thorium recycle and core design aspects. The thorium fuel-cycle studies were performed under the USAEC's programme to develop economical, heavy-water reactors with optimum fuel utilization characteristics. During the initial stages, several potentially desirable combinations of fuel and clad materials were considered concurrently with investigations on compatible fuel assemblies and core geometries. Three different fuel assemblies were selected for further study: Zircaloy clad, coextruded thorium metal nested cylinders; vibratory compacted, SAP clad, thorium oxide pin bundles; and pelletized thorium monocarbide pin bundles. In the evolution of near-optimum conceptual designs, several cores featuring each of the three fuel assemblies were developed. As the work progressed, the initial goal to achieve optimum fuel utilization gradually shifted more toward consideration of economic performance. Except for relatively minor variations due to changes in the core arrangement or operating conditions, the uranium-fuelled HWOCR plant design was used in all the thorium-fuelled concepts. Three reference conceptual designs utilizing the three different fuel elements were completed and are based on extensive parameter studies and recycle considerations; the economics and fuel utilization of each design were evaluated, and technical feasibility, development costs, and compatibility with a uranium-optimized HWOCR were considered. A development programme was evolved that would lead to

  11. Progress in the Development of a Heavy-Water Moderated and Cooled Thorium-Uranium-233 Converter

    International Nuclear Information System (INIS)

    On account of its excellent neutron economy the heavy-water reactor is suitable for development as a high-gain converter or as a breeder in the thorium-uranium-233 cycle. In the Federal Republic of Germany work on these lines is being carried out at the Jülich Nuclear Research Centre in co-operation with the Siemens Company. Prominence is being given to a pressure-vessel reactor moderated and cooled by heavy water. Contrary to natural uranium types the thorium reactor will have a quasi-homogeneous core structure and a much lower heavy-water content and therefore will be in appearance more similar to a light-water reactor. With a quasi-homogeneous lattice and the fact that flux peaks are more easily avoided in a D2O core than in an H2O one, the mean specific fuel power (which is decisive for the economic efficiency of a thorium reactor) attains a higher value than in reactors with clustered fuel elements or in light-water reactors. The D2O-thorium reactor will have core power densities in the range of those attained in boiling light-water reactors. This power density is greater than in any other advanced converter. Development work can be based on the well-known technology of water-cooled reactors and on operating experience with the multipurpose reactor (MZFR) at Karlsruhe. The total time and cost required for the development of the heavy-water reactor towards a thorium-uranium-converter or breeder is therefore relatively small. This type will profit in particular from the great progress made in recent years with Zircaloy-canned ceramic fuel rods and with fuel element development. At minimum fuel cycle costs in the range of 1 mill/kWh, including costs for D2O inventory, the specific consumption of 235U for maintaining a reprocessed equilibrium cycle is calculated to be 0.20 g 235U/MWd, including diffusion tails. Calculation methods are shown to be in good agreement with lattice experiments carried out in the Siemens Argonaut Reactor at Garching. On account of the

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

    International Nuclear Information System (INIS)

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

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

  14. 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.

  15. 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)

  16. 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

  17. 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)....

  18. Heavy-water-moderated pressure-tube reactor safety

    International Nuclear Information System (INIS)

    Several countries have heavy-water-moderated, pressure-tube reactors either in commercial operation or in late prototype stages. The supporting safety research and development includes such areas as the thermohydraulics of circuit depressurization, heat transfer from the fuel, heat rejection to the moderator from dry fuel, fuel and sheath behaviour, and fuel channel integrity. We review the work done in Canada, Great Britain, Italy and Japan, and describe some of the experimental tests underlaying the methods of accident analysis. The reactors have safety systems which, in the event of an accident, are able to shut down the reactor, keep the fuel cooled, and contain any released radioactivity. We summarize the characteristics of these safety systems (shutdown, emergency core cooling, and containment) in the various reactors, and discuss other reactor characteristics which either prevent accidents or reduce their potential demand on the safety systems. (author)

  19. 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)

  20. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  4. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    International Nuclear Information System (INIS)

    The irradiation of Th232 breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U238. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

  5. Characteristics of gas-cooled reactor with water moderator and rankine cycle

    International Nuclear Information System (INIS)

    Full text: Nuclear energy with both thermal and fast neutrons, despite on a number of potential benefits, will economically lose energy on organic fuels, if innovative solutions won't be found. It is presented a gas-cooled channel reactor with water-moderator, working on a piston Brayton cycle engine. Efficiency up to 45 percent is achieved. Thermophysical calculations of fuel assemblies show that the proposed reactor fuel assemblies can be constructed in a simplified scheme without heat shield that reduces the creation costs, the costs of coolant pumping, loss of neutrons and dimensions of the core

  6. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. 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...

  10. 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

  11. 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)

  12. 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)

  13. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

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

  14. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  15. Header feedwater supply and power distribution stability in channel boiling water cooled reactors

    International Nuclear Information System (INIS)

    Boundaries of radial-azimuthal instability of the reactor neutron field during the supply of all feedwater and a part of it (25%) to downtake pipes of the separating drum (75% of feedwater come to distributive group headers) are found out for NPP with a RBMK type reactor. Results of computer calculation of the transient process at NPP caused by 2% step increase of nominal pressure in a head collector of a feedwater electric pump are also presented for comparison of the above methods of feed-water supply. Calculation is carried out according to the OKA program with provision for the control system of the reactor total power. It is shown that the boundary of ''mean period'' instability does not change but the reserve in respect to the ''fast'' space instability slightly increases when header feedwater supply at NPP from RBMK is used. It is noted that requirements to the pressure regulator system quick action in a separating drum are increased when the header feedwater supply is used. This fact is explained by the fact that considerable pressure drop in a separating drum occurs during some accidents (for example, at false operation of the emergensy protective system)

  16. Accident analysis of heavy water cooled thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  17. Accident analysis of heavy water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  18. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value

  19. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors)

  20. 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

  1. 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)

  2. Cooling of concrete structure in advanced heavy water reactor

    International Nuclear Information System (INIS)

    Innovative nuclear power plants are being designed by incorporation of passive systems to the extent possible for enhancing the safety by elimination of active components. BARC has designed Advanced Heavy Water Reactor (AHWR) incorporating several passive systems to facilitate the fulfillment of safety functions of the reactor during normal operation, residual heat removal, emergency core cooling, confinement of radioactivity etc. In addition to these passive systems, an innovative passive technology is being developed to protect, the concrete structure in high temperature zone (V1-volume). Passive Concrete Cooling System (PConCS) uses the principle of natural circulation to provide cooling outside the insulation cabinet encompassing high temperature piping. Cooling water is circulated from overhead GDWP in cooling pipes fixed over corrugated plate on outer surface of insulation cabinet and maintains low temperature of concrete structure. Modular construction of insulation cabinet and cooling pipes external to the concrete surface simplifies the design, construction and refurbishment if required. The paper describes the details of passive technology for concrete cooling. (author)

  3. 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

  4. 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)

  5. 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

  6. 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

  7. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  8. Observations on the removal of gadolinium from the moderator system of pressurised heavy water reactor (PHWR) and advanced heavy water reactor (AHWR)

    International Nuclear Information System (INIS)

    Investigation on ion exchange removal of gadolinium taken as gadolinium nitrate, which is used as neutron poison in the moderator system of Pressurised Heavy Water Reactor (PHWR) and proposed to be used in Advanced Heavy Water Reactor (AHWR) was carried out. Mixed bed operation consisting of (a) strong acid cation resin (SAC) and strong base anion resin (SBA) and (b) strong acid action resin and acrylic acid based nitrate loaded weak base anion resin were employed for the removal gadolinium from its aqueous solution at pH 5. In the former case, the outlet of the mixed bed was highly alkaline, which resulted in precipitation of gadolinium hydroxide. In the latter case, the pH of the system never crossed 6 and gadolinium was effectively picked up on the resin without getting precipitated. Series operation consisting for strong acid cation resin followed by mixed bed column consisting of strong acid cation resin and strong base anion resin/acrylic acid based weak base anion resin was also investigated. In the first case where strong base anion resin was used, there was precipitation in the system owing to the increase in pH while in the case where weak base anion resin was used there was no problem of precipitation and gadolinium removed effectively and the pH was around 6. (author)

  9. 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

  10. Loss of feed water analyses of advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. Passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. The case analysed in this paper is the loss of feedwater to steam drum which results in decrease in heat removal from core. This also causes increase in reactor pressure. Further consequences depend upon various protective and engineered safeguard systems like relief system, reactor trip, isolation condenser and advanced accumulator. Analysis has been done using code RELAP5/MOD3.2. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, qualities and flow in different part of Primary Heat Transport (PHT) system. (author)

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

    International Nuclear Information System (INIS)

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

  12. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    International Nuclear Information System (INIS)

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  13. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  14. Serious accidents on boiling water reactors (BWR)

    International Nuclear Information System (INIS)

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

  15. Comparison Of The Worth Of Critical And Exponential Measurements For Heavy-Water-Moderated Reactors

    International Nuclear Information System (INIS)

    Critical and exponential experiments in general produce overlapping information on reactor lattices. Over the past ten years the Savannah River Laboratory has been operating a heavy-water critical, the PDP, and an exponential, the SE, in parallel. This paper summarizes SRL experience to give results and recommendations as to the applicability of the two kinds of facilities in different experiments. Six types of experiments are considered below: (1) Buckling measurements in uniform isotropic lattices Here Savannah River has made extensive comparisons between single-region criticals, exponentials, substitution criticals, and PCTR type measurements. The only difficulties in the exponentials seem to lie in the radial-buckling determinations. If these can be made successfully, the exponentials can offer good competition to the criticals. Material requirements are greatest for the single-region criticals, roughly comparable for the substitution criticals and exponentials, and least for the PCTR measurements. (2) Anisotropic and void effects SRL experiments with the criticals and with critical-exponential comparisons are reviewed briefly here and at greater length in a companion paper. (3) Evaluation of control systems Adequately analysed exponential experiments appear to give good results for total-worth measurements. However, for adequate study of overall flux shaping, flux tilts, etc. a full-sized critical such as the PDP is required. (4) Temperature coefficients Exponential experiments provide an excellent method for determining the temperature coefficient of buckling for uniform lattice heating. A special facility, the PSE, at Savannah River permits such measurements up to temperatures of 215°C. For non-uniform lattice heating criticals are generally preferred. (5) Mixed lattices Actual reactors rarely use the simple uniform lattices to which the exponentials basically apply. Critical experiments with mixed loadings are used at SRL both in measuring direct effects

  16. Design guide for category II reactors light and heavy water cooled reactors

    International Nuclear Information System (INIS)

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems

  17. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  18. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U233 )O2 and (Th-Pu)O2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m3 capacity directly into the core. Gravity driven water pool of capacity 6000 m3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  19. 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)

  20. Containment for Heavy-Water Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    The safety principles applicable to heavy-water, gas-cooled reactors are outlined, with a view to establishing containment specifications adapted to the sites available in Switzerland for the construction of nuclear plants. These specifications are derived from dose rates considered acceptable, in the event of a serious reactor accident, for persons living near the plant, and are based on-meteorological and demographic conditions representative of the majority of the country's sites. The authors consider various designs for the containment shell, taking into account the conditions which would exist in the shell after the maximum credible accident. The following types of shell are studied: pre-stressed concrete; pre-stressed concrete with steel dome; pre-stressed concrete with inner, leakproof steel lining; steel with concrete side shield to protect against radiation; double shell. The degree of leak proofing of the shells studied is regarded as a feature of the particular design and not as a fixed constructional specification. The authors assess the leak proofing properties of each type of shell and establish building costs for each of them on the basis of precise plans, with the collaboration of various specialized firms. They estimate the effectiveness of the various shells from a safety standpoint, in relation to different emergency procedures, in particular release into the atmosphere through appropriate filters and decontamination of the air within the shell by recycling through batteries of filters. The paper contains a very detailed comparison of about 10 cases corresponding to various combinations of design and emergency procedure; the comparison was made using a computer programme specially established for the purpose. The results are compared with those for a reactor of the same type and power, but assembled together with the heat exchangers in a pre-stressed concrete shell. (author)

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

    International Nuclear Information System (INIS)

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

  2. Effect of flooding of annulus space between CT and PT with light water coolant and heavy water moderator on AHWR reactor physics parameters

    International Nuclear Information System (INIS)

    In AHWR lattice, the pressure tube (PT) contains light water coolant which carries away heat generated in the fuel pins. The pressure tube (PT) and calandria tube (CT) are separated by air (density=0.0014 g/cc) of wall thickness 1.79 cm. Air between pressure tube and calandria tube acts as insulator and minimize the heat transfer from coolant to moderator which is outside the calandria tube. In case of flooding or under any unforeseeable circumstances, the air gap between the coolant tube and calandria tube may be filled with the light water coolant or heavy water moderator. This paper gives the details of effect of filling the annulus space between CT and PT with light water or heavy water moderator on reactor physics parameters. (author)

  3. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  4. Reconcentration technology of heavy water used as moderator and coolant at Cernavoda Plant

    International Nuclear Information System (INIS)

    One of the basic conditions to ensure the efficient operation of the CANDU reactor plant is maintaining a high isotope purity of heavy water in the moderator and cooling system. It is thus necessary that concentration be 99,80% mol. D2 O for the the moderator, and 99,75% mol. D2 O for the coolant, with water making up the rest. When the nuclear power plant is in service, heavy water in the moderator and cooling circuits may via various ways become mixed with H2 O impurities. Therefore, an isotope purification and, reconcentration of these fluids is necessary. A heavy-water reconcentration technology involving isotope distillation in vacuum in columns filled with phosphorous bronze, was developed by the Institute of Cryogenics and Isotope Separation Rm. Valcea, based on research and experience regarding heavy water concentration in the distillation pilot

  5. An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

    International Nuclear Information System (INIS)

    An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

  6. Core design of heavy water cooled thorium breeder reactor with negative void reactivity and improved breeding performance

    International Nuclear Information System (INIS)

    A core of heavy water cooled thorium breeder reactor that produces 3.5 GWt energy using Th-233U oxide fuel has been studied to depict a concrete design specification. In order to improve the breeding performance compared to that of our previous study, one of key parameters in core design: moderator to fuel volume ratio (MFR) is re-surveyed. By reducing MFR from 1.0 to 0.6, the swing of keff during a cycle is considerably flattened, keeping negative void coefficient. The batch number is 3 and the refueling scheme employs out-in method to limit the radial power peaking factor less than 1.3. Due to efficient internal conversion, the reactivity of the core slightly increases with burnup, so that the cycle length is extended up to 1,300 days. Consequently, high averaged burnup of 80 GWd/t and breeding ratio of 1.07 at middle of cycle is achieved without any blankets. The number of control rods made of B4C is 19 and the total reactivity worth is -6.5% dk/k. The present core uses Zircaloy-4 as cladding material, the fast neutron fluence at EOC (End Of Cycle), however, exceeds its limit due to hard spectrum and long cycle length. As a part of future study, design will be further explored considering cladding integrity. (authors)

  7. 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)

  8. 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.

  9. 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

  10. Educational laboratory based on a multifunctional analyzer of a reactor of a nuclear power plant with a water-moderated water-cooled reactor

    International Nuclear Information System (INIS)

    Authors presents an educational laboratory Safety and Control of a Nuclear Power Facility established by the Department of Automation for students and specialists of the nuclear power industry in the field of control, protection, and safe exploitation of reactor facilities at operating, constructing, and designing nuclear power plants with water-moderated water-cooled reactors

  11. Numerical Analysis on the Calandria Tubes in the Moderator of a Heavy Water Reactor Using OpenFOAM and Other Codes

    Science.gov (United States)

    Chang, Se-Myong; Kim, Hyoung Tae

    2014-06-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors.

  12. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    International Nuclear Information System (INIS)

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  13. 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

  14. 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.

  15. Experimental and analytical study on thermal hydraulics in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Araya, Fumimasa; Ohnuki, Akira; Yoshida, Hiroyuki; Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Study and development of reduced-moderation spectrum water reactor proceeds as a option of the future type reactor in Japan Atomic Energy Research Institute (JAERI). The reduced-moderation spectrum in which a neutron has higher energy than the conventional water reactors is achieved by decreasing moderator-to-fuel ratio in the lattice core of the reactor. Conversion ratio in the reduced-moderation water reactor can be more than 1.0. High burnup and long term cycle operation of the reactor are expected. A type of heavy water cooled PWR and three types of BWR are discussed as follows; For the PWR, (1) critical heat flux experiments in hexagonal tight lattice core, (2) evaluation of cooling limit at a nominal power operation, and (3) analysis of rewetting cooling behavior at loss of coolant accident following with large scale pipe rupture. For the BWR, analyses of cooling limit at a nominal power operation of, (1) no blanket BWR, (2) long term cycle operation BWR, and (3) high conversion ratio BWR. The experiments and the analyses proved that the basic thermal hydraulic characteristics of these reduced-moderation water reactors satisfy the essential points of the safety requirements. (Suetake, M.)

  16. A Management Strategy for the Heavy Water Reflector Cooling System of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H. S.; Park, Y. C.; Lim, S. P. (and others)

    2007-11-15

    Heavy water is used as the reflector and the moderator of the HANARO research reactor. After over 10 years operation since first criticality in 1995 there arose some operational issues related with the tritium. A task force team(TFT) has been operated for 1 year since September 2006 to study and deduce resolutions of the issues concerning the tritium and the degradation of heavy water in the HANARO reflector system. The TFT drew many recommendations on the hardware upgrade, tritium containing air control, heavy water quality management, waste management, and tritium measurement system upgrade.

  17. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed. (author)

  18. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed

  19. 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)

  20. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  1. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

  2. Analysis of water cooled reactors stability

    International Nuclear Information System (INIS)

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

  3. 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.)

  4. 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

  5. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 oC average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

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

    International Nuclear Information System (INIS)

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

  7. 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.

  8. 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

  9. 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

  10. Effect of moderator density distribution of annular flow on fuel assembly neutronic characteristics in boiling water reactor cores

    International Nuclear Information System (INIS)

    The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. (author)

  11. Photoneutron compensating method for boric acid concentration measuring instrument in heavy water moderated reactor

    International Nuclear Information System (INIS)

    In a boric acid concentration measuring instrument in a heavy water moderated reactor, a portion of γ-ray from Na-24 and Mn-56 is reacted with heavy water to form photoneutrons. The photoneutrons cause errors in the measurement for B-10 concentration. Then, in the present invention, a sample liquid containing photoneutron sources is supplied during normal measurement and a sample liquid removed with the photoneutron sources by passing through an ion exchange resin tower is supplied upon calibration of the measuring instrument. Then, the extent for the of effect of neutron sources and γ-nuclides is obtained by calculation from the measuring value to calibration the extent of the photoneutrons. Further, a method of using a counter tube having a Cd filter is used in combination during normal measurement to enable continuous measurement without exchanging the sample liquid. Accordingly, the influence of photoneutrons can be compensated and boric acid concentration can be measured at high accuracy. (N.H.)

  12. 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

  13. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    International Nuclear Information System (INIS)

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates

  14. 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...

  15. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

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

  16. 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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    2013-06-14

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

  19. Heavy water reactors physics

    International Nuclear Information System (INIS)

    An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors)

  20. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

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

  1. Optimization of U–Th fuel in heavy water moderated thermal breeder reactors using multivariate regression analysis and genetic algorithms

    International Nuclear Information System (INIS)

    Highlights: • A new method useful for the parametric analysis and optimization of reactor core designs. • This uses the strengths of genetic algorithms (GA), and regression splines. • The method is applied to the core fuel pin cell of a PHWR design. • Tools like java, R, and codes like Serpent, Matlab are used in this research. - Abstract: An analysis and optimization of a set of neutronics parameters of a thorium-fueled pressurized heavy water reactor core fuel has been performed. The analysis covers a detailed pin-cell analysis of a seed-blanket configuration, where the seed is composed of natural uranium, and the blanket is composed of thorium. Genetic algorithms (GA) is used to optimize the input parameters to meet a specific set of objectives related to: infinite multiplication factor, initial breeding ratio, and specific nuclide’s effective microscopic cross-section. The core input parameters are the pitch-to-diameter ratio, and blanket material composition. Recursive partitioning of decision trees (rpart) multivariate regression model is used to perform a predictive analysis of the samples generated from the GA module. Reactor designs are usually complex and a simulation needs a significantly large amount time to execute, hence implementation of GA or any other global optimization techniques is not feasible, therefore we present a new method of using rpart in conjunction with GA. Due to using rpart, we do not necessarily need to run the neutronics simulation for all the inputs generated from the GA module rather, run the simulations for a predefined set of inputs, build a regression fit to the input and the output parameters, and then use this fit to predict the output parameters for the inputs generated by GA. The rpart model is implemented as a library using R programming language. The results suggest that the initial breeding ratio tends to increase due to a harder neutron spectrum, however a softer neutron spectrum is desired to limit the

  2. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

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

    International Nuclear Information System (INIS)

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

  4. 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)

  5. 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

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

    International Nuclear Information System (INIS)

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

  7. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels. Guidebook addendum: Heavy water moderated reactors

    International Nuclear Information System (INIS)

    A Guidebook on Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels (IAEA-TECDOC--233) was issued by the International Atomic Energy Agency in August 1980. This document contains a wide variety of information of the physics, thermal-hydraulics, fuels, and fuel cycle economics for light water moderated research and test reactors. In consideration of the special features of heavy water moderated research and test reactors (hereafter referred to as heavy water research reactors), this Addendum to IAEA-TECDOC--233 has been prepared to assist operators and physicists from these reactors in determining whether conversion from HEU to LEU fuel designs is technically feasible for their specific reactor, and to assist in making a smooth transition to the use of LEU fuel designs where appropriate. The organization of this Addendum follows that of IAEA-TECDOC--233 as closely as possible in order to provide a consistent presentation of the information and to minimize the repetition of information that is common to both heavy water and light water research reactors. Distinctive features of the heavy water reactors are addressed where applicable

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  10. 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

  11. 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.)

  12. 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

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

    International Nuclear Information System (INIS)

    The paper describes the performance studies of a new core cooling monitor (electrical cylindrical heater) for BWRs. Such a detector has been successfully tested at various elevations, including the lower plenum, in the Barsebaeck nuclear power plant under normal operating conditions, and also in various environments in a 160 bar loop (with sudden uncoveries) and in the laboratory (up to 1265 C). It can be operated in two modes: the core cooling mode and the temperature mode, where it actually acts as a thermometer. It currently appears ready for implementation in BWR installations. (orig.)

  14. 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

  15. 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

  16. 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

  17. 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)

  18. Review of the use and state of development of the various reactor types

    International Nuclear Information System (INIS)

    The report gives a review of the reactor types being of importance from today's point of view for use as stationary power reactors. These are heavy water reactors, light water reactors (pressurized water reactor, Soviet pressurized water reactor, Soviet light-water-graphite reactors, boiling water reactors), gas-cooled reactors (gas-graphite reactors, high temperature reactors), and fast breeder reactors. (HJ)

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

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

  4. CFD analysis of flow and temperature distribution inside the calandria of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Passive systems are being examined for the future AHWR reactor designs. One of these systems is the passive moderator cooling system, which removes heat from the moderator in case of a Station Black Out (SBO). The heavy-water moderator gets heated due to the residual heat from the core structures and rises upward due to buoyancy. This is cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator (Calandria) pressure beyond safe limits. In this paper CFD investigations are carried out to study the temperature distributions and flow distribution inside the Calandria using a three-dimensional CFD code, OpenFoam 2.2.0. The results provide a band of operable mass flow rates which are safe for operation by virtue of prediction of hot spots in the Calandria. (author)

  5. Advanced water-cooled reactor technologies. Rationale, state of progress and outlook

    International Nuclear Information System (INIS)

    Eighty per cent of the world's power reactors are water cooled and moderated. Many improvements in their design and operation have been implemented since the first such reactor started commercial operation in 1957. This report addresses the safety, environmental and economic rationales for further improvements, as well as their relevance to currently operating water reactors

  6. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

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

  7. Technologies for tritium control in fission reactors moderated with heavy water

    International Nuclear Information System (INIS)

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors' belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the 'on-line' detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs

  8. 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

  9. 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

  10. 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.

  11. 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

  12. On the physics design of Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    Full text: The AHWR is a 920 MWth, vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The prime objective is to produce power utilizing thorium available abundantly in India from a relatively simple system with enhanced safety level. It is endowed with several innovative safety features such as negative coolant void reactivity, heat removal through natural circulation and passive containment cooling. The development of reactor design has drawn heavily on the experience generated through design and operation of Pressurised Heavy Water Reactors (PHWR) and Boiling Water Reactor (BWR) in India. It was an opportunity to develop a reactor system using thorium-based fuel and gain some valuable experience. A non-proliferative thorium/U-233 based closed fuel cycle is chosen for AHWR. Plutonium discharged from PHWRs is used as the fissile seed fuel with thorium for the generation of U-233 and then as a top-up fuel in the equilibrium core along with self-sustaining U-233 in the thorium matrix. The physics design has several challenges in achieving negative void reactivity, spatial core control, on-line fuelling and minimization of inventory of plutonium fuel. It is difficult to achieve negative coolant void coefficient in a heavy water moderated pressure tube type reactor. For this a multi-pronged approach involving pitch reduction, heterogeneous cluster design and use of mild absorbers is chosen. Plutonium bearing fuel is located separately in the outer region of the cluster with self-sustaining U-233 bearing fuel in the inner region of the cluster. A small amount of mild absorber is located in the centre of the cluster. The void coefficient varies with burnup and it is a challenge to have it negative throughout the core. The state of nuclear data for the elements of interest and type of neutron spectrum in the reactor puts heavy demand on the calculation models and validation of reactivity coefficients to ensure

  13. Thorium utilization in heavy water moderated Accelerator Driven Systems

    International Nuclear Information System (INIS)

    much more realistic option, since it gives us a good gain with heavy water as coolant and even with light water cooled heavy water moderated reactor, the gain is somewhat lower, but still acceptable. We show that the self sustaining cycle for a critical system gives rather low discharge burnup. To obtain higher burnup, the critical reactor needs an external feed of fissile material. We show that the accelerator driven sub-critical mode of operation gives acceptable burnup in a self sustaining cycle. (author)

  14. The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

    International Nuclear Information System (INIS)

    The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

  15. Application of the dose limitation system to the control of carbon-14 releases from heavy-water-moderated reactors

    International Nuclear Information System (INIS)

    Heavy-water-moderated reactors produce substantially more carbon-14 than light-water reactors. Applying the principles of the systems of dose limitation, the paper presents the rationale used for establishing the release limit for effluents containing this nuclide and for the decisions made regarding the effluent treatment in the third nuclear power station in Argentina. Production of carbon-14 in PHWR and the release routes are analysed in the light of the different effluent treatment possibilities. An optimization assessment is presented, taking into account effluent treatment and waste management costs, and the collective effective dose commitment due to the releases. The contribution of present carbon-14 releases to future individual doses is also analysed in the light of an upper bound for the contribution, representing a fraction of the individual dose limits. The paper presents the resulting requirements for the effluent treatment regarding carbon-14 and the corresponding regulatory aspects used in Argentina. (author)

  16. Nuclear Reactor RA Safety Report, Vol. 5, Reactor cooling systems

    International Nuclear Information System (INIS)

    RA reactor cooling system enable cooling during normal operation and under possible accidental conditions and include: technical water system, heavy water system, helium gas system, system for heavy water purification and emergency cooling system. Primary cooling system is a closed heavy water circulation system. Heavy water system is designed to enable permanent circulation and twofold function of heavy water. In the upward direction of cooling it has a coolant role and in the downward direction it is the moderator. Separate part of the primary coolant loop is the system for heavy water purification. This system uses distillation and ion exchange processes

  17. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    International Nuclear Information System (INIS)

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the

  18. Corrosion Products Identification at Normal Water and Hydrogen Water Chemistry in Boiling Water Reactors

    International Nuclear Information System (INIS)

    The corrosion products sampled from condensate and feedwater systems of boiling water reactors (BWRs) at normal water chemistry (NWC) and hydrogen water chemistry (HWC) operating condition were analyzed with dissolution and instrumental simulation methods. The crystallite and amorphous of iron oxides were separated by means of dissolving method with appropriate chemical solution. The iron oxide composition and content were analyzed by X-ray diffraction (XRD) and inductively coupled plasma atomic emission spectrometer (ICP-AES) in this study. The insoluble iron oxides were obtained in influent and effluent of condensate demineralizer comprised mostly crystalline structure of hematite, magnetite and non-crystallite form of amorphous at NWC and HWC environments. Both goethite and lepidocrocite compositions are of minor importance in feed water system. Crystallite and amorphous compositions in the samples will be calculated from the new developing dissolution method. The crystalline phase of corrosion products are varied with water chemistry conditions in BWRs. The oxide characterization of system corrosion products includes compositions, morphology and particle size can effectively provide the ways of solving crud removal problem in different condition for the performance of condensate demineralizer. The feasibility of identifying other iron oxides and hydroxides in corrosion products is briefly discussed and the mechanisms of iron oxide formation formed around BWR piping will also be shown in detail in this report. Moreover, it will be figured out the properties of radioactive corrosion products growing in different operation periods. The results can also assist in plant units to improve the crud reduction countermeasures and to optimize the system water chemistry. (authors)

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

    International Nuclear Information System (INIS)

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

  20. Current developments and future challenges in physics analyses of the NRU heavy water research reactor

    International Nuclear Information System (INIS)

    The National Research Universal (NRU) reactor is heavy water cooled and moderated, with on-power fueling capability. TRIAD, a 3D two-group diffusion code, is currently used for support of day-to-day NRU operations. Recently, an MCNP full reactor model of NRU has been developed for benchmarking TRIAD. While reactivity changes and flux and power distributions from both methods are in reasonably good agreement, MCNP appears to eliminate a k-eff bias in TRIAD. Beyond TRIAD's capability, MCNP enables the assessment of radiation in the NRU outer structure. Challenges include improving TRIAD accuracy and MCNP performance, as well as performing NRU core-following using MCNP. (author)

  1. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

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

  2. The concept and application of miniaturisation boiling in cooling system

    International Nuclear Information System (INIS)

    The purpose of this research is to study and examine the phenomena of miniaturisation-boiling, which intensely scattered into a large number of minute liquid particles from a water droplet surface into the atmosphere, when the droplet collided with a heating surface. For the material of the heating surface, the following were used: stainless steel (SUS 303 A Cr= 17 %, Ni= 8 %), sapphire (Al2O3), brass, copper and carbon plane. The material was heated in order to study the miniaturisation-boiling and droplet bounding phenomena at a very high temperature (160 degree Celsius- 420 degree Celsius). The phenomenon was photographed by a high-speed camera (10000 fps) from the horizontal direction. The nuclear fusion reactor which needs a severe cooling, and heat removal cooling method through special boiling leads to this research. (author)

  3. The concept and application of miniaturization boiling in cooling system

    International Nuclear Information System (INIS)

    The purpose of this research is to study and examine the phenomena of miniaturization-boiling, which intensely scatters with a large number of minute liquid particles from a water droplet surface to the atmosphere, when the droplet collided with a heating surface. As the material of the heating surface, the following were used: stainless steel (SUS 303 A Cr=17%,Ni=8%), sapphire (Al3O2), brass, copper and carbon plane. The material was heated in order to study the miniaturization-boiling and droplet bounding phenomena at a very high temperature (160 degree C- 420 degree C). The phenomenon was photographed by a high-speed camera (10,000 fps) from the horizontal direction. The nuclear fusion reactor needs a very severe cooling, heat removal cooling method by special boiling is lead to this research. (Author)

  4. 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.)

  5. 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

  6. 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)

  7. Cost analysis and economic comparison for alternative fuel cycles in the heavy water cooled canadian reactor (CANDU)

    International Nuclear Information System (INIS)

    Three main options in a CANDU fuel cycle involve use of: (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option, including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. For the 3 cycles selected (natural uranium, slightly enriched uranium, recovered uranium), levelized fuel cycle cost calculations are performed over the reactor lifetime of 40 years, using unit process costs obtained from literature. Components of the fuel cycle costs are U purchase, conversion, enrichment, fabrication, SF storage, SF disposal, and reprocessing where applicable. Cost parameters whose effects on the fuel cycle cost are to be investigated are escalation ratio, discount rate and SF storage time. Cost estimations were carried out using specially developed computer programs. Share of each cost component on the total cost was determined and sensitivity analysis was performed in order to show how a change in a main cost component affects the fuel cycle cost. The main objective of this study has been to find out the most economical option for CANDU fuel cycle by changing unit prices and cost parameters

  8. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  9. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  10. Advances in heavy water reactors

    International Nuclear Information System (INIS)

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

  11. 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)

  12. Physics design of advanced heavy water reactor utilising thorium

    International Nuclear Information System (INIS)

    An Advanced Heavy Water Reactor (AHWR) is being developed in India with the aim of utilising thorium for power generation. AHWR is a vertical pressure tube type reactor cooled by boiling light water and moderated by heavy water. It has been optimised for the thorium cycle. The main design objective is to be self-sustaining in 233U with most of the power from the thorium fuel using plutonium as the external fissile feed. It incorporates several advanced safety features namely, heat removal through natural circulation and a negative void coefficient of reactivity. The reactor has been designed to produce 750 MW(th) at a discharge burnup of 20,000 MWd/H(e). The physics design of AHWR has followed an evolutionary path ranging from a seed and blanket concept to a simplified composite cluster to achieve a good thermal hydraulic coupling. We have designed a composite cluster using both kinds of fuel namely, (Th-UO2 and (Th-Pu)O2. With plutonium seed, negative void coefficient can be achieved by making the spectrum harder. This was done by using a pyrocarbon scatterer in the moderator. The void coefficient strongly depends on plutonium. As plutonium burns very rapidly, it is not possible to achieve uniformly negative void coefficient with burnup in this cluster. Alternatively, burnable poison can be used within the cluster to achieve negative void coefficient taking advantage of the flux redistribution and change in spectrum upon voiding. Here, it is possible to achieve almost constant void reactivity with burnup resulting in a good thermal hydraulic coupling. The cluster design presently incorporates a central burnable absorber region. Boiling light water coolant requires that the core power distribution be optimised with thermal hydraulic parameters. The peaking factors inside the cluster should be low so as to have significant margin in operational conditions and to avoid burnout in accident conditions. The variation of reactivity from cold clean to hot operating has

  13. Comparative study of plutonium burning in heavy and light water reactors

    International Nuclear Information System (INIS)

    There is interest in the U.S. and world-wide in reducing the burden on geological nuclear fuel disposal sites. In some disposal scenarios, the decay heat loading of the surrounding rock limits the commercial spent fuel capacity of the sites. In the long term (100 to 1,500 years), this decay heat is generated primarily by actinides, particularly 241Am and 241Pu. One possible approach to reducing this decay-heat burden would be to reprocess commercial spent nuclear fuel and use intermediate-tier thermal reactors to 'burn' these actinides and other transuranics (plutonium and higher actinides). The viability of this approach is dependent on the detailed changes in chemical and isotopic compositions of actinide-bearing fuels after irradiation in thermal reactor spectra. The intermediate-tier thermal burners could bridge the commercial water-cooled reactors and fast reactors required for ultimate consumption of the transuranics generated in the commercial reactors. This would reduce the number of such fast reactors required to complete the mission of burning transuranics. If thermal systems are to be used for the transmutation mission, it is likely that they would be similar to or are advanced versions of the systems currently used for power generation. In both the U.S. and Canada, light- and heavy-water-cooled thermal reactors are used for power generation in the commercial nuclear sector. About 103 pressurized- and boiling- light water reactors (PWRs and BRWs) are deployed in the U.S. nuclear industry while about 18 CANDU (heavy-water-cooled) reactors are used in the Canadian industry. There are substantial differences between light and heavy water-cooled reactors that might affect transmutation potential. These arise from differences in neutron balance of the reactors, in neutron energy spectra, in operational approaches (e.g., continuous refueling enhancing fuel burnup), and so on. A systematic study has been conducted to compare the transmutation potentials of

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

    International Nuclear Information System (INIS)

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

  15. New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors

    International Nuclear Information System (INIS)

    Fast reactors cooled by lead or lead-bismuth alloy offer new interesting fuel cycle and fuel management options by virtue of the superb neutronics and safety features of these heavy liquid metal (HLM) coolants. One option is once-for-life cores having relatively low power density. These cores are fueled in the factory; there is no refueling or fuel shuffling on site. A second option is very long-life cores being made of a fissioning zone and a natural uranium blanket zone. The fissioning zone very slowly drifts toward the blanket. A third option is multirecycling of light water reactor (LWR) discharged fuel without partitioning of transuranics (TRUs) in fuel-self-sustaining reactors. LWR spent fuel could provide the initial fuel loading after extracting fission products and ∼90% of its uranium. The makeup fuel is natural or depleted uranium. A fourth option is the high-burnup once-through fuel cycle using natural or depleted uranium feed. The initial fuel loading of this reactor is a mixture of enriched and natural uranium. The natural uranium utilization is 10 to 20 times higher than that of a once-through LWR. A fifth option is transmutation of TRUs from LWRs using critical HLM-cooled reactors; such reactors could be designed to have the same high actinide burning capability of accelerator-driven systems and have comparable safety, but at a substantially lower cost. These novel reactor designs and fuel management options are hereby reviewed

  16. The key design features of the Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    The 235 MWe Indian Advanced Heavy Water Reactor (AHWR) is a vertical, pressure tube type, boiling light water cooled reactor. The three key specific features of design of the AHWR, having a large impact on its viability, safety and economics, relate to its reactor physics, coolant channel, and passive safety features. The reactor physics design is tuned for maximising use of thorium based fuel, and achieving a slightly negative void coefficient of reactivity. The fulfilment of these requirements has been possible through use of PuO2-ThO2 MOX, and ThO2 -U233O2 MOX in different pins of the same fuel cluster, and use of a heterogeneous moderator consisting of pyrolytic carbon and heavy water in 80%-20% volume ratio. The coolant channels of AHWR are designed for easy replaceability of pressure tubes, during normal maintenance shutdowns. The removal of pressure tube along with bottom end-fitting, using rolled joint detachment technology, can be done in AHWR coolant channels without disturbing the top end-fitting, tail pipe and feeder connections, and all other appendages of the coolant channel. The AHWR incorporates several passive safety features. These include core heat removal through natural circulation, direct injection of Emergency Core Coolant System (ECCS) water in fuel, passive systems for containment cooling and isolation, and availability of a large inventory of borated water in overhead Gravity Driven Water Pool (GDWP) to facilitate sustenance of core decay heat removal, ECCS injection, and containment cooling for three days without invoking any active systems or operator action. Incorporation of these features has been done together with considerable design simplifications, and elimination of several reactor grade equipment. A rigorous evaluation of feasibility of AHWR design concept has been completed. The economy enhancing aspects of its key design features are expected to compensate for relative complexity of the thorium fuel cycle activities required to

  17. Development of an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated reactors

    International Nuclear Information System (INIS)

    The results of work carried out under IAEA Contract No. 2336/RB are described (subject: an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated (VVER) reactors). The basic principles of an accounting system for this type of nuclear power plant are outlined. The general structure and individual units of the information computer program used to achieve automated accounting are described and instructions are given on the use of the program. A detailed example of its application (on a simulated nuclear power plant) is examined

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  19. Experimental Boiling Water Reactor decontamination and decommissioning project

    International Nuclear Information System (INIS)

    The author begins by discussing the problems encountered during decontamination and decommissioning. Next, he discusses waste packaging and recycling. His last topic of lessons learned is subdivided into prevention and early detection, recovery issues, management issues, and noteworthy practices

  20. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  1. Methods of Containment Adopted for the EL4 Reactor and Projected Heavy-Water, Gas-Cooled Plants

    International Nuclear Information System (INIS)

    After a brief description of the plant, the paper explains the principles adopted for preventing the release of waste gas, from the EL4 reactor and refers to some of the difficulties associated with this type of containment. From the economic standpoint, the authors present the results of a comparative civil engineering study of pre-stressed concrete and steel shells for a projected 60 MW(e) power station, giving various values for accidental pressures. They demonstrate the influence of the stress values adopted. (author)

  2. 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. ...

  3. 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)

  4. 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

  5. High-conversion and high-burnup core concepts for pressure-tube-type heavy water reactors

    International Nuclear Information System (INIS)

    A high-conversion and a high-burnup core concept for a pressure-tube-type heavy water reactor are presented and analyzed from the standpoint of neutronics. These core concepts are based on the fact that neutron spectrum can be shifted by adjusting the amount of heavy water moderator outside the pressure tubes without affecting core-cooling capability. For the high-conversion core, where the heavy water moderator is replaced by a gas such as CO2 [carbon dioxide], a conversion ratio of more than 0.8 and an average discharge fuel burnup of 50GWd/t have been estimated to be attained with standard design fuel assemblies having 7.5% fissile Pu enrichment. For the high-burnup core, where fuel assemblies burned in the high-conversion (gas) region are relocated into the burner (heavy water) region, an average discharge fuel burnup of 110GWd/t has been estimated

  6. Analytical and experimental studies of fretting-corrosion and vibrations of fuel assemblies of a VVER-1000 water cooled and water moderated power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Drozdov, Y.N. [IMASH Machine Study Institute named after A.A.Blagonravov of the Russian Academy of Sciences, Moscow (Russian Federation); Tutnov, A.A.; Tutnov, A.A.; Alexeyev, E.E. [Kurchatov Institute Russian Research Centre, Moscow (Russian Federation); Makarov, V.V.; Afanasyev, A.V. [Experimental and Design Organization Gidropress (Russian Federation)

    2007-07-01

    The report covers the methods and results of the latest analytical and experimental studies of fretting corrosion and natural vibrations of a VVER-1000 reactor fuel assemblies (FA). The process of fretting-corrosion was investigated using a multi-specimen facility that simulated fragments of fuel rod-to-spacer grid and lower support grid mating units. A computational model was developed for vibrations in the mechanical system of a fuel rod fragment and a spacer grid fragment. A calculational and experimental modal analysis of a FA was performed. Natural frequencies, modes and decrements of FA vibrations were determined and a satisfactory coincidence of analytical and experimental results was obtained. The assessment of fretting-corrosion process dynamics was made and its dependences on operational factors were obtained. (authors)

  7. Analytical and experimental studies of fretting-corrosion and vibrations of fuel assemblies of a VVER-1000 water cooled and water moderated power reactor

    International Nuclear Information System (INIS)

    The report covers the methods and results of the latest analytical and experimental studies of fretting corrosion and natural vibrations of a VVER-1000 reactor fuel assemblies (FA). The process of fretting-corrosion was investigated using a multi-specimen facility that simulated fragments of fuel rod-to-spacer grid and lower support grid mating units. A computational model was developed for vibrations in the mechanical system of a fuel rod fragment and a spacer grid fragment. A calculational and experimental modal analysis of a FA was performed. Natural frequencies, modes and decrements of FA vibrations were determined and a satisfactory coincidence of analytical and experimental results was obtained. The assessment of fretting-corrosion process dynamics was made and its dependences on operational factors were obtained. (authors)

  8. Analysis of thorium/U-233 lattices and cores in a breeder/burner heavy water reactor

    International Nuclear Information System (INIS)

    Due to the inevitable dwindling of uranium resources, advanced fuel cycles in the current generation of reactors stand to be of great benefit in the future. Heavy water moderated reactors have much potential to make use of thorium, a currently unexploited resource. Core fuelling configurations of a Heavy Water Reactor based on the self-sufficient thorium fuel cycle were simulated using the DRAGON and DONJON reactor physics codes. Three heterogeneously fuelled reactors and one homogeneously fuelled reactor were studied. (author)

  9. Stability monitoring of the Dodewaard boiling-water reactor

    International Nuclear Information System (INIS)

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

  10. 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

  11. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Heavy water and nonproliferation

    International Nuclear Information System (INIS)

    This report begins with a historical sketch of heavy water. The report next assesses the nonproliferation implications of the use of heavy water-moderated power reactors; several different reactor types are discussed, but the focus is on the natural uranium, on-power fueled, pressure tube reactor CANDU. The need for and development of on-power fueling safeguards is discussed. Also considered is the use of heavy water in plutonium production reactors as well as the broader issue of the relative nuclear leverage that suppliers can bring to bear on countries with natural uranium-fueled reactors as compared to those using enriched designs. The final chapter reviews heavy water production methods and analyzes the difficulties involved in implementing these on both a large and a small scale. It concludes with an overview of proprietary and nonproliferation constraints on heavy water technology transfer

  15. 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

  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. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  18. Balancing passive and active systems for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

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

    International Nuclear Information System (INIS)

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

  20. 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.

  1. 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

  2. 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.

  3. Fuel cycle flexibility in Advanced Heavy Water Reactor (AHWR) with the use of Th-LEU fuel

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) is being designed for large scale commercial utilization of thorium (Th) and integrated technological demonstration of the thorium cycle in India. The AHWR is a 920 MW(th), vertical pressure tube type cooled by boiling light water and moderated by heavy water. Heat removal through natural circulation and on-line fuelling are some of the salient features of AHWR design. The physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Our recent efforts have been directed towards a case study for the use of Th-LEU fuel cycle in a once-through mode. The discharged Uranium from Th-LEU cycle has proliferation resistant characteristics. This paper gives the initial core, fuel cycle characteristics and online refueling strategy of Th-LEU fuel in AHWR. (author)

  4. Heavy water reactor user requirement document status and path forward

    International Nuclear Information System (INIS)

    This is a Power Point presentation containing: -1. Outline; - 2. Background - Stages and the HWR-URD; - 3. Background - What do we have now?; - 4. Draft 'D0' of the HWR-URD; - 5. The 2001 April Consultancy; - 6. The Path Forward. - 7. Summary. The 2001 April consultancy addressed the following items: - 1. Status of the URD: - The document structure is substantially complete. Appendices or paragraphs need to be added on some topics; - The URD covers the overall requirements for the design of the nuclear island (NI) and interfaces to the BOP for future HWRs. It contains policies, high-level requirements and important requirements for key areas that are of interest to HWR users; - The draft focuses on horizontal-pressure-tube, heavy water moderated and cooled HWRs. When and if other types need to be considered, the TWG will identify these and direct how they are to be addressed; - The level of detail in the draft and its treatment of the international aspects of the topic are appropriate; - 2. Areas needing further consideration: - While intended for future reactors, it is recognized that regulators may wish to use the URD as a benchmark for evaluating existing or replicate reactors; - The international aspects of the URD require detailed review at each stage of its development; - The EUR and EPRI-URD have had targeted reviews by the regulators. This may be appropriate for the HWR-URD but would add 6 to 12 months to the schedule; - These items will be the focus of the AGM planned for 2002 January. The Path Forward section pinpoints the terms: - 2001 August, implying the task, incorporate comments from April Consultancy, producing draft D1; - 2002 January, implying AGM in Vienna, namely, focus on areas needing further consideration: - Ensure requirements are clearly differentiated from desirable features; - Confirm international aspects are appropriately considered; - Establish need for additional step - regulatory review; - 2002, implying revise URD to reflect

  5. 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 ...

  6. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  7. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  9. 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.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. 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)

  15. 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

  16. 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.

  17. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

    International Nuclear Information System (INIS)

    In the framework of research carried out on a CO2-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors)

  18. 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...

  19. Current developments and future challenges in physics analyses of the NRU heavy water research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, S.; Wilkin, B.; Leung, T., E-mail: nguyens@aecl.ca, E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    The National Research Universal (NRU) reactor is heavy water cooled and moderated, with on-power fueling capability. TRIAD, a 3D two-group diffusion code, is currently used for support of day-to-day NRU operations. Recently, an MCNP full reactor model of NRU has been developed for benchmarking TRIAD. While reactivity changes and flux and power distributions from both methods are in reasonably good agreement, MCNP appears to eliminate a k-eff bias in TRIAD. Beyond TRIAD's capability, MCNP enables the assessment of radiation in the NRU outer structure. Challenges include improving TRIAD accuracy and MCNP performance, as well as performing NRU core-following using MCNP. (author)

  20. Stability of a Steam Cooled Fast Power Reactor, its Transients Due to Moderate Perturbations and Accidents

    International Nuclear Information System (INIS)

    The dynamic behaviour of a steam cooled fast power reactor is investigated with respect to stability, transients due to moderate perturbations at the operating point, and accidents. The studies were performed for a direct cycle, integral plant design for different system pressures, component arrangements and component designs. The stability domain of such a plant is found to be mainly determined by pressure, fuel temperature and coolant density coefficients of reactivity. Other design parameters are of minor influence on stability. The plant is load-following and displays acceptable performance if the reactivity coefficients are not too close to their limiting values. If they are, effective controllers can be designed which ensure good plant operation. The consequences of accidents may be limited by proper design and adequate counteraction

  1. A study of the friction and wear processes of the structural components of fuel assemblies for water-cooled and water moderated power reactors

    International Nuclear Information System (INIS)

    The friction forces affect the fuel assembly (FA) strength at all the stages of its lifecycle. The paper covers the methods and the results of the pre-irradiation experimental studies of the static and dynamic processes the friction forces are involved in. These comprise the FA assembling at the manufacturer, fuel rod flow-induced vibration and fretting-wear in the fuel rod-to-cell friction pairs, rod cluster control assembly (RCCA) movement in the FA guide tubes, FA bowing, FA loading-unloading into the core, irradiation-induced growth and thermal-mechanical fuel rod-to-spacer grid interaction. (authors)

  2. Innovative Pressure Tube Light Water Reactor with Variable Moderator Control

    International Nuclear Information System (INIS)

    The features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide the reactor with considerable flexibility for continuous design improvements and developments. This paper presents the development of innovative pressure tube light water reactor, which has the ability to advance the current pressure tubes reactors. The proposed design is aimed to simplify the pressure tubes reactors by: - replacing heavy water by a light water as a coolant and moderator, - adopting batch refueling instead of on-line refueling. Furthermore, the design is based on proven technologies, existing fuel and structure materials. Therefore, it is reasonable to expect significant capital cost savings, short licensing and introduction period of the proposed concept into the power production grid. The basic novelty of the proposed design is based on an idea of variable moderator content in the core and 'breed and burn' mode of operation. Both concepts were extensively investigated and reported in the past (2) (3) (4). In order to evaluate a practical reactor design build on proven technology, several features of the advanced CANDU reactor (ACR-1000) were adopted. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The proposed design is basically pressure tube light water reactor with variable moderator Control (PTVM LWR). This paper presents a detailed description of the PTVM core design and demonstrates the reactivity control and the 'breed and burn' mode of operation, which are implemented by the variation of the moderator in the core, from a

  3. 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.)

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

    International Nuclear Information System (INIS)

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

  5. 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)

  6. 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)

  7. 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.)

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

    International Nuclear Information System (INIS)

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

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. 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.

  12. 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.

  13. Calculational experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled and moderated reactor WWER-type NPPs. Final report

    International Nuclear Information System (INIS)

    The results of testing of equipment at Bohunice NPP and pipeline systems at Unit 3 of Kozloduy NPP (WWER-440 type reactors) are presented in this Final Report. These results side by side with experimental values of natural frequencies and decrements also include experimental data about vibration modes of tested equipment and pipelines. For the first time the results of new calculational-experimental examination of equipment seismic resistance at Unit 2 of Armenian NPP are presented. At Kozloduy NPP direction's request the planed additional tests of some selected items were put off on 1997. Instead of postponed tests we carried out detailed analysis of our past inspections of numerous equipment seismic resistance at the Unit 5 of Kozloduy NPP. Experimental data with results of additional analysis are presented

  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. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-19

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

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

    International Nuclear Information System (INIS)

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

  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. 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)

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

    International Nuclear Information System (INIS)

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

  1. Water cooled FBNR nuclear reactor

    International Nuclear Information System (INIS)

    elements from the fuel chamber up into the core. A fixed suspended core is formed in the reactor. In the shut down condition, the suspended core breaks down and the fuel elements leave the core and fall back into the fuel chamber by the force of gravity. The fuel elements are made of UO2 micro spheres embedded in zirconium and cladded by zircaloy. Any signal from any of the detectors, due to any initiating event, will cut-off power to the pump, causing the fuel elements to leave the core and fall back into the fuel chamber, where they remain in a highly subcritical and passively cooled conditions. The fuel chamber is cooled by natural convection transferring heat to the water in the tank housing the fuel chamber. The nest step in the development of FBNR is the construction of its prototype. Efforts are being made to secure participants in such an endeavor. (author)

  2. Topical papers on heavy water, fuel fabrication and reactors

    International Nuclear Information System (INIS)

    A total of four papers is presented. The first contribution of the Federal Republic of Germany reviews the market situation for reactors and the relations between reactor producers and buyers as reflected in sales agreements. The second West German contribution gives a world-wide survey of fuel element production as well as of fuel and fuel element demand up to the year 2000. The Canadian paper discusses the future prospects of heavy-water production, while the Ecuador contribution deals with small and medium-sized nuclear power plants

  3. A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results

    International Nuclear Information System (INIS)

    A flashing-driven passive moderator cooling system is being developed at AECL for CANDU reactors. Preliminary simulations and experiments showed that the concept was feasible at normal operating power. However, flow instabilities were observed at low powers under conditions of variable and constant calandria inlet temperatures. This finding contradicted code predictions that suggested the loop should be stable at all powers if the calandria inlet temperature was constant. This paper discusses a series of separate-effects tests that were used to identify the sources of low-power instabilities in the experiments, and it explores methods to avoid them. It concludes that low-power instabilities can be avoided, thereby eliminating the discrepancy between the experimental and code results. Two factors were found to be important for loop stability: (1) oscillations in the calandria outlet temperature, and (2) flashing superheat requirements, and the presence of nucleation sites. By addressing these factors, we could make the loop operate in a stable manner over the whole power range and we could obtain good agreement between the experimental and code results. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-12-01

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

  5. Hydrodynamically induced dryout and post dryout important to heavy water reactors: A yearly progress report

    International Nuclear Information System (INIS)

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed to study the CBF induced by various flow instabilities. Details of this test loop are presented. Initial test results have been presented. The two-phase flow regimes and hydrodynamic behaviors in the post dryout region have been studied under propagating rewetting conditions. The effect of subcooling and inlet velocity on flow transition as well as on the quench front propagation was investigated. The test liquid was Freon 113 which was introduced into the bottom of the quartz test section whose walls were maintained well above the film boiling temperature of the test liquid, via a transparent heat transfer fluid. The flow regimes observed down stream of the upward moving quench front were the rough wavy, the agitated, and the dispersed droplet/ligaments. A correlation for the flow regime transition between the inverted annular and the dispersed droplet/ligament flow patterns was developed. The correlation showed a marked dependence on the void fraction at the CBF location and hence on the flow regime encountered in the pre-CBF region

  6. 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)

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

    International Nuclear Information System (INIS)

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

  8. 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

  9. Improving economics and safety of water cooled reactors. Proven means and new approaches

    International Nuclear Information System (INIS)

    Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

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

    International Nuclear Information System (INIS)

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

  11. Advanced reactor design and safety objectives - The heavy water reactor perspective

    International Nuclear Information System (INIS)

    This paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for Advanced Heavy Water reactor Systems are based on systematic reviews of current operating CANDU stations to identify opportunities for generic improvements in reliability, operability, maintainability and to address emerging licensing or safety issues. Hence these requirements represent those for the evolutionary development of the Advanced Heavy Water Reactor systems factoring in the considerable operating experience of the CANDU stations. This evolutionary approach to the development of advanced heavy water reactors will be consistent with a philosophy of minimizing the risk to future reactor owners whose requirements are for a reliable, low cost unit

  12. 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

  13. 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

  14. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. 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

  17. 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)

  18. 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.)

  19. 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)

  20. Boiling Water Reactor Loading Pattern Optimization Using Simple Linear Perturbation and Modified Tabu Search Methods

    International Nuclear Information System (INIS)

    An automated system for designing a loading pattern (LP) for boiling water reactors (BWRs) given a reference LP and control rod (CR) sequence has been developed. This system employs the advanced nodal code SIMULATE-3 and a BWR LP optimization code FINELOAD-3, which uses a simple linear perturbation method and a modified Tabu search method to select potential optimized LP candidates. Both of these unique methods of FINELOAD-3 were developed to achieve an effective BWR LP optimization strategy and to have high computational efficiency. FINELOAD-3 also adjusts deep CR positions to compensate for the core reactivity deviation caused by fuel shuffling. The objective function is to maximize the end-of-cycle core reactivity while satisfying the specified thermal margins and cold shutdown margin constraints. This optimization system realized the practical application for real BWR LP design. Computer time needed to obtain an optimized LP for a typical BWR/5 octant core with 15 depletion steps is ∼4 h using an engineering workstation. This system was extensively tested for real BWR reload core designs and showed that the developed LPs using this system are equivalent or better than the manually optimized LPs

  1. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

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

  2. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

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

  3. Study of new structures adapted to gas-graphite and gas-heavy water reactors

    International Nuclear Information System (INIS)

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

  6. 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

  7. Heavy Water Components Test Reactor Decommissioning

    International Nuclear Information System (INIS)

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D and D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  8. Transient subchannel analysis of turbine trip without bypass with IC of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Full text: The proposed advanced heavy water reactor (AHWR) is a 750 MWth vertical pressure tube type boiling light water cooled and heavy water moderated reactor. Safety analysis is carried out for various postulated initiating events (PIE) for AHWR. Turbine trip without bypass is one such PIE. Turbine trip results in rise in system pressure. Absence of bypass flow may lead to actuation of relief and safety valves depending upon system characteristics. System pressure rise activates various protective and engineered safeguard systems like reactor trip, isolation condenser and advanced accumulator limiting the consequences of the event. In this paper, a turbine trip without bypass is analysed, using a two-fluid code RELAPS/MOD3.2 [Fletcher, 1995] and subchannel analysis code COBRA IV-I. [Wheeler, 1976] Global analysis does not take into account the effect of different fuel pins with different peaking factors, effect of axial and radial cross flow mixing between adjacent subchannels. Combination of these two codes gives a better insight into the problem. The maximum rated reactor channel which houses 54 pin fuel bundle is modeled for COBRA IV-I simulation. The transient forcing function option of COBRA IV-I, validated by [Iwamura, 1994] with their flow-power transient experiments, has been used for transient thermal hydraulic parameter prediction for this study. Transient inlet pressure, inlet mass flow rate, inlet temperature, outlet enthalpy and the heat flux/power obtained from RELAPS /MOD3.2 simulation are among the boundary conditions employed on the COBRA-IV-I simulation. The thermal-hydraulic parameters predicted by RELAPS/MOD3.2 along with the effect of subchannel flow on the fuel element temperatures predicted by COBRA-IV-I are presented and discussed in this paper

  9. Analytical modelling and study of the stability characteristics of the Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    An analytical model has been developed to study the thermohydraulic and neutronic-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 design configuration considered may experience both Ledinegg and Type I and Type II density-wave instabilities depending on the operating condition. Some methods of suppressing these instabilities were found out. In addition, it was found 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 strong influence on the Type I and Type II instabilities. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (author)

  10. 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

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

    International Nuclear Information System (INIS)

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

  12. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  13. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    International Nuclear Information System (INIS)

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  14. 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)

    Full text: The official decision on construction of a Nuclear Power Plant (NPP) in Kazakhstan has been accepted by the Kazakhstan government. The results on the choice of the power reactors projects of the NPP are given in the report. The choice has been carried out with the aim to develop recommendation on reactors of the NPP for construction in Kazakhstan. The choice of the reactors was based on the system comparative analysis of the most advanced power reactors projects using 15 criteria system of the nuclear, radiating and ecological safety and economic competitiveness. Following Pressurized Water Reactor (PWR, WWR) projects have been subjected to the system comparative analysis: 1) Large Sized Reactors (700 MW(el) and up): such as EPR, developed by Germany Siemens and France Framatome companies; CANDU-9, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation, developed by Korea Power Engineering Company, Inc.; APWR, Japanese advanced reactor, developed by Japan Atomic Power Company, Japan, Mitsubishi Heavy Industries, Japan and Westinghouse Electric Company, USA; WWER-1000 (V-392) - development by Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor, development by Westinghouse, USA/Genesi, Italy. 2) Medium Sized Reactors (300 MWe - 700 MWe): AP-600, passive PWR, developed by the Westinghouse company; CANDU-6, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); An-tilde-600, passive PWR, developed by Nuclear Power Institute of China; WWER-640, Russian passive reactor, developed by 0KB ''Gidropress'' Experimental and Design Office, Russian Federation; MS-600, developed by Mitsubishi Company; KSNP-600, developed by Korea Power Engineering Company, Inc., South Korea. 3) Small Sized Reactors (a few MWe- 300 MWe): IRIS, reactor of IV generation, developed by the International Corporation of 13

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

    International Nuclear Information System (INIS)

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

  16. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

  17. 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)...

  18. 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...

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

    International Nuclear Information System (INIS)

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

  20. 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)

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

    International Nuclear Information System (INIS)

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

  2. 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. Experimental investigations and seismic analysis for benchmark study of 1000 MW WWER type units (water-cooled and moderated reactor) nuclear power plant Kozloduy. Final report

    International Nuclear Information System (INIS)

    This final report covers the experimental investigations and seismic analysis for benchmark study of WWER-1000 type units of Kozloduy NPP. A description of the research carried out includes terms of reference, criteria, design parameters and method of analysis, description of mathematical models developed, method for verification of anchorage components and results obtained by the mentioned methods. Experimental investigations concerned with verification of seismic safety related components are described with the conclusions drawn

  4. 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)

  5. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

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

    International Nuclear Information System (INIS)

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

  7. 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

  8. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

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

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

    International Nuclear Information System (INIS)

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

  10. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

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

  11. 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)

  12. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

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

    International Nuclear Information System (INIS)

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

  14. 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)

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

    International Nuclear Information System (INIS)

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

  16. 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)

  17. In-air PIXE for analyzing heavy metals in water boiled in pans

    International Nuclear Information System (INIS)

    The release rates of heavy metals from pans were measured for boiling water as well as for an acidic solution prior to an investigation on the release or sorption of trace elements due to cooking of food by boiling. The boiled samples were condensed and analyzed by means of in-air PIXE. The release of heavy metals was measured for five kinds of pans. For all pans the release rates were considerably more increased by boiling of a 5% solution of acetic acid. Furthermore it was found that by using the alumina coated aluminum pan (alumina pan) the respective release rates of Fe, Cu and Zn were all less than 50 μg per 100 cm2 of the pan surface dipped in the solution, and that monitoring of the contents of aluminum in the boiled solution enabled the estimation of the contribution of metal elements from the pan wall. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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

  8. Boiling water reactor off-gas systems evaluation

    International Nuclear Information System (INIS)

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

  9. Mixed oxide fuel for water cooled reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Widener, S.K.

    1986-01-01

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

  12. Problems related with the power regulation of reactors by physico-chemical methods, and the behaviour of water and heavy water in nuclear reactors

    International Nuclear Information System (INIS)

    Experience of the CEA heavy water reactors and a systematic study of the radiolytic decomposition of water in the core of swimming-pool reactors are described. Setting up of reactivity control by physico-chemical methods. Reactivity control by homogeneous poisoning of the reactor A comparison of the evolution of xenon poisoning with the residual anti reactivity of the poison in solution during its nuclear consumption establishes the programme which must govern the variation in its concentration if the exact compensation is to be produced The behaviour of the poison towards the reactor materials under the particular operational conditions must be taken into account. Radiolytic decomposition of water in the reactors in the presence of soluble poisons: A study of the effect of certain chemically inert salts, present in small concentrations in the water, on its radiolytic decomposition rate, has led to some new results which are discussed. The choice of a soluble poison is justified on the basis of the above results. Reactivity control by the use of a gaseous absorbent The use of a gas control rod circuit for compensation purposes, in place of solid control rods is described. The use of soluble poisons in the moderator to compensate the xenon effect, and of a gaseous absorbent in a circuit known as a gas control rod form original aspects of the reactivity control in the reactor EL 4. (authors)

  13. 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

  14. 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

  15. 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

  16. Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

    International Nuclear Information System (INIS)

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

  17. Thermohydraulic relationships for advanced water cooled reactors, and the role of IAEA

    International Nuclear Information System (INIS)

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1999. It was included into the IAEA's Programme following endorsement in 1995 by the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote international information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. (authors)

  18. 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

  19. 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

  20. Study on hydrodynamically induced dryout and post dryout important to heavy water reactors

    International Nuclear Information System (INIS)

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed. A study was conducted on the CHF induced by various flow instabilities. Details of this test loop are presented

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

    International Nuclear Information System (INIS)

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

  2. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-08

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

  5. Plant life management processes and practices for heavy water reactors

    International Nuclear Information System (INIS)

    In general, heavy water reactor (HWR) nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. Their decisions are depending on essentially business model. They involve the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. Continued plant operation, including operation beyond design life, called 'long term operation, depends, among other things, on the material condition of the plant. This is influenced significantly by the effectiveness of ageing management. Key attributes of an effective plant life management program include a focus on important systems, structure and components (SSCs) which are susceptible to ageing degradation, a balance of proactive and reactive ageing management programmes, and a team approach that ensures the co-ordination of and communication between all relevant nuclear power plant and external programmes. Most HWR NPP owners/operators use a mix of maintenance, surveillance and inspection (MSI) programs as the primary means of managing ageing. Often these programs are experienced-based and/or time-based and may not be optimised for detecting and/or managing ageing effects. From time-to-time, operational history has shown that this practice can be too reactive, as it leads to dealing with ageing effects (degradation of SSCs) after they have been detected. In many cases premature and/or undetected ageing cannot be traced back to one specific reason or an explicit error. The root cause is often a lack of communication, documentation and/or co-ordination between design, commissioning, operation or maintenance organizations. This lack of effective communication and interfacing frequently arises because, with the exception of major SSCs, such as the fuel channels or steam generators, there is a lack of explicit

  6. Boiling water reactor uranium utilization improvement potential

    International Nuclear Information System (INIS)

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

  7. Safety analysis for boiling water reactors

    International Nuclear Information System (INIS)

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

  8. Department of Energy's team's analyses of Soviet designed VVERs (water-cooled water-moderated atomic energy reactors)

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document contains apprendices A through P of this report. Topics discussed are: a cronyms and technical terms, accident analyses reactivity control; Soviet safety regulations; radionuclide inventory; decay heat; operations and maintenance; steam supply system; concrete and concrete structures; seismicity; site information; neutronic parameters; loss of electric power; diesel generator reliability; Soviet codes and standards; and comparisons of PWR and VVER features. (FI)

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

    CERN Document Server

    Golikov, V V; Shabalin, E P

    2002-01-01

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

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

    CERN Document Server

    Golikov, V V; Shabalin, E P

    2002-01-01

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

  11. Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

  12. 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

  13. 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)

  14. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

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

  15. Advanced reactor design and safety objectives. The heavy water reactor perspective

    International Nuclear Information System (INIS)

    The development objectives of advanced heavy water reactors (AHWRs) should be guided by the requirements of the operating utilities. The paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for AHWR systems are based on systematic reviews of current operating CANDU stations. Hence these requirements represent those for the evolutionary development of AHWR systems, factoring in the considerable operating experiences of the CANDU stations. The requirements for the new HWR designs can be summarized under economic objectives, safety objectives, operational objectives and other utility requirements. 2 refs

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  18. 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

  19. 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)

  20. 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.

  1. 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.

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. 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.)

  7. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  8. Efficient Water Management in Water Cooled Reactors

    International Nuclear Information System (INIS)

    number of the countries that have recently begun to consider the introduction of nuclear power are in water scarce regions, which would certainly limit the possibility for deployment of nuclear power plants, in turn hindering these countries' development and energy security. Thus, there is a large incentive to enhance efforts to introduce innovative water use, water management practices and related technologies. Water management for nuclear power plants is gaining interest in IAEA Member States as an issue of vital importance for the deployment of nuclear power. Recent experience has shown that some nuclear power plants are susceptible to prolonged drought conditions, forcing reactors to be shut down or power to be reduced to a minimal level. In some cases, environmental issues have resulted in regulations that limit the possibility for water withdrawal as well as water discharge. Regarding the most common design for cooling nuclear power plants, this has led to a complicated siting procedure for new plants and expensive retrofits for existing ones. The IAEA has already provided its Member States with reports and documents that address the issue. At the height of nuclear power expansion in the 1970s, the need for guidance in the area resulted in publications such as Thermal Discharges at Nuclear Power Stations - Their Management and Environmental Impact (Technical Reports Series No. 155) and Environmental Effects of Cooling Systems (Technical Reports Series No. 202). Today, amid the so-called nuclear renaissance, it is of vital importance to offer guidance to the Member States on the issues and possibilities that nuclear power water management brings. Management of water at nuclear power plants is an important subject during all phases of the construction, operation and maintenance of any nuclear power plant. Water management addresses the issue of securing water for condenser cooling during operation, for construction (during the flushing phase), and for inventory

  9. Defense-in-depth evaluation model development strategy for pressurized heavy water reactor low power and shutdown operations

    International Nuclear Information System (INIS)

    The objective of defense-in-depth evaluation is to assess the level of defense-in-depth maintained during the various plant maintenance activities. Especially for shutdown and outage operations, the defense-in-depth might be challenged due to the reduction in redundancy and diversity resulting from the maintenance. Outage Risk Indicator of Nuclear Power Plants(ORION) which is a outage risk monitor for Korean NPPs is under development. ORION has the ability to assess the outage risk qualitatively using safety function assessment trees and deterministic margins such as time to boil. For pressurized heavy water reactors in Korea, defense-in-depth evaluation model development strategy to reflect the unique characteristics of pressurized heavy water reactors. The strategy and development steps were discussed in this paper

  10. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  11. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    International Nuclear Information System (INIS)

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution

  12. Fast neutron flux in heavy water reactors

    International Nuclear Information System (INIS)

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author)

  13. 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.

  14. 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

  15. 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

  16. Thermophysical properties of materials for water cooled reactors

    International Nuclear Information System (INIS)

    The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs

  17. Vaporization Rate Analysis of Primary Cooling Water from Reactor PUSPATI TRIGA (RTP) Tank

    International Nuclear Information System (INIS)

    Primary cooling system consists of pumps, heat exchangers, probes, a nitrogen-16 diffuser and associated valves is connected to the reactor TRIGA PUSPATI (RTP) tank by aluminium pipes. Both the primary cooling system and the reactor tank is filled with demineralized light water (H2O), which serves as a coolant, moderator as well as shielding. During reactor operation, vaporization in the reactor tank will reduce the primary water and contribute to the formation of vapor in the reactor hall. The vaporization may influence the function of the water subsequently may affect the safety of the reactor operation. It is essential to know the vaporization rate of the primary water to ensure its functionality. This paper will present the vaporization rate of the primary cooling water from the reactor tank and the influence of temperature of the water in the reactor tank to the vaporization rate. (author)

  18. Development of light water reactors and cooling systems

    International Nuclear Information System (INIS)

    Development of Light Water Reactors (LWR) started in US. Japan imported this technology from US, and its construction and improvement had been done by adding Japanese original technologies. This article outlined development history of LWR, its plant system and main components, ECCS and accident management. Most operating LWR were those rapidly developed from 1960 to 1970 and associated accident response was so designed for all assumed conditions at that time. At the Fukushima Daiichi nuclear power station, needed works could not be down well due to all losses of AC and DC power, inability to assure plant state at control room under power stoppage, work interruption caused by aftershock and road blockage by Tsunami drift, and accident management was not effective and accident was enlarged. After the Fukushima nuclear accident, enhanced measures against more stringent conditions such as external events should be prepared to assure safety of nuclear power station using latest knowledge. (T. Tanaka)

  19. 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)

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

    International Nuclear Information System (INIS)

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

  1. 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

  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. 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)

  4. Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

    International Nuclear Information System (INIS)

    Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 233U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

  5. 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...

  6. Status and future program of research and development on reduced-moderation water reactors

    International Nuclear Information System (INIS)

    The reduced-moderation water reactor (RMWR) aims at effective utilization of uranium resource, multiple recycling of plutonium and high burn-up and long operation cycle. To realize the RMWR, it is required to establish high neutron energy spectrum and to keep void reactivity coefficient negative and higher conversion ratio greater than one. Technical items for validation of the design are heat removal characteristics of tight lattice core, reactor core physics performance and MOX fuel and cladding cans integrity under irradiation. Development of fuel cans for high burn-up and low-cost reprocessing technology of MOX spent fuels are needed to improve economics. Related R and D works have been performed under the collaboration of industry, academia and national institute as one of technical developments of innovative nuclear systems in Japan. Based on results on these confirmation tests, technology demonstration reactor facilities have been proposed to construct and finally test for the deployment of the RMWR. (T. Tanaka)

  7. Behavior of tritium in heavy water reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. 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.)

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

    International Nuclear Information System (INIS)

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

  11. 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...

  12. Neutron spectrum calculation and safety analysis for supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    The supercritical water reactor is one of the six reactors recommended by Generation Ⅳ International Forum. Compared with existing light water reactors, the supercritical water reactor has advantages of high thermal efficiency, simplified system structure and low cost. The physical model of the supercritical water reactor is established with MCNP program in this paper, which solves the problem of intricate geometry of fuel assembly. The change of coolant density along the axis is considered and the neutron spectrum distribution of different regions of the core is calculated. The safety in loss of coolant accident for the supercritical water reactor and the effect of missing coolant in different regions on the reactivity and effective multiplication factor analyzed. The results show the supercritical water reactor core has high security. The countermeasures of loss of coolant accident is studied and the effectiveness of boron water cooling is validated. The research not only provide important reference for the construction and security analysis of the supercritical water reactor, but also has great significance for the application and development of the supercritical water reactor. (authors)

  13. Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Holly; Pencer, Jeremy; Yetisir, Metin; Leung, Laurence [Atomic Energy of Canada Limited, Ontario (Canada)

    2012-03-15

    Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR.

  14. Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

    International Nuclear Information System (INIS)

    Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR

  15. Progress in development and design aspects of advanced water cooled reactors

    International Nuclear Information System (INIS)

    The objective of the Technical Committee Meeting (TCM) was to provide an international forum for technical specialists to review and discuss technology developments and design work for advanced water cooled reactors, safety approaches and features of current water cooled reactors and to identify, understand and describe advanced features for safety and operational improvements. The TCM was attended by 92 participants representing 18 countries and two international organizations and included 40 presentations by authors of 14 countries and one international organization. A separate abstract was prepared for each of these presentations. Refs, figs, tabs

  16. The major species of heavy metal aerosol resulting from water cooling systems and spray dryer systems during incineration processes

    Science.gov (United States)

    Wey; Yang; Wei

    1998-11-01

    Trace toxic metals in municipal solid waste may escape from the incineration process in flue gas, in dry collected ash, in wet scrubbed ash, or as a suspended aerosol. Therefore, understanding the behavior of heavy metals in the flue gas and the best controls in the air pollution control equipment are important and necessary. The control conditions of water cooling and spray dryer systems during incineration processes significantly influence the formation of heavy metal compounds. The formation of chromium (Cr), lead (Pb), and cadmium (Cd) species under various control conditions (water cooling tower and spray dryer reactor) was investigated in this study. The object of the experiment is to understand the effects of water cooling and spray dryer systems individually on the formation of heavy metal species. The operating parameters that are evaluated include different control systems, control temperatures, and chlorine content. A thermodynamic equilibrium model was also used to evaluate experimental data. In order to match real incineration conditions, a two-stage simulation was performed in this experiment. The results showed that the relationship of speciation between the simulation prediction and X-ray diffraction (XRD) analysis is consistent for Cr compounds; both indicated that Cr2O3 is the major species. The relationship is almost the same for Cd compounds, but not for Pb compounds. PMID:9846130

  17. Methods and technologies for cost reduction in the design of water cooled reactor power plants

    International Nuclear Information System (INIS)

    The Specialists Meeting was organized in the framework of the IAEA International Working Group on Advanced Technologies for Water-Cooled Reactors. Its purpose was to provide an international forum for review and discussion on recent results in research and development on different methods and technologies of current and advanced water-cooled reactor power plants, which can lead to reduced investment and operation, maintenance and fuel-cycle costs of the plants. 27 specialists representing 10 countries and the IAEA took part in the meeting. 10 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  18. Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Water: Performance and Stability

    Science.gov (United States)

    Lisowski, Darius D.

    This experimental study investigated the thermal hydraulic behavior and boiling mechanisms present in a scaled reactor cavity cooling system (RCCS). The experimental facility reflects a ¼ scale model of one conceptual design for decay heat removal in advanced GenIV nuclear reactors. Radiant heaters supply up to 25 kW/m2 onto a three parallel riser tube and cooling panel test section assembly, representative of a 5° sector model of the full scale concept. Derived similarity relations have preserved the thermal hydraulic flow patterns and integral system response, ensuring relevant data and similarity among scales. Attention will first be given to the characterization of design features, form and heat losses, nominal behavior, repeatability, and data uncertainty. Then, tests performed in single-phase have evaluated the steady-state behavior. Following, the transition to saturation and subsequent boiling allowed investigations onto four parametric effects at two-phase flow and will be the primary focus area of remaining analysis. Baseline conditions at two-phase flow were defined by 15.19 kW of heated power and 80% coolant inventory, and resulted in semi-periodic system oscillations by the mechanism of hydrostatic head fluctuations. Void generation was the result of adiabatic expansion of the fluid due to a reduction in hydrostatic head pressure, a phenomena similar to flashing. At higher powers of 17.84 and 20.49 kW, this effect was augmented, creating large flow excursions that followed a smooth and sinusoidal shaped path. Stabilization can occur if the steam outflow condition incorporates a nominal restriction, as it will serve to buffer the short time scale excursions of the gas space pressure and dampen oscillations. The influences of an inlet restriction, imposed by an orifice plate, introduced subcooling boiling within the heated core and resulted in chaotic interactions among the parallel risers. The penultimate parametric examined effects of boil-off and

  19. 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

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

    International Nuclear Information System (INIS)

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

  1. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Science.gov (United States)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  2. 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

  3. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-01

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

  5. 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

  6. High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

    International Nuclear Information System (INIS)

    The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

  7. Novel Diagnostics and Computational Methods of Neutron Fluxes in Boiling Water Reactors

    OpenAIRE

    Loberg, John

    2010-01-01

    The focus in this thesis is to improve knowledge of the BWR related uncertainties void, channel bow, and control rods. The presence of void determines the moderation of neutrons in BWRs. A high void fraction is less efficient in moderating neutrons than a low one. As a consequence, the ratio of thermal to fast neutrons is dependent on the surrounding void fraction. In this thesis, calculations with 2D/3D codes corroborate this dependence, the void correlation, to be linear and very robust to ...

  8. Pin cooling and dryout in steady local boiling

    International Nuclear Information System (INIS)

    A study is presented of pin cooling and dryout mechanisms in steady local boiling, with the particular objective of understanding the substantial dryout margins observed in the KNS local boiling experiments. Mechanisms for the entry of liquid into the voided region are discussed, and pin cooling by draining liquid films deduced to be likely. The conditions required for interruption of the film flow, and hence for dryout, are examined, with particular attention to vapour/liquid interactions causing film breakdown, inhibition of rewetting and film flooding. This leads to the hypothesis that dryout occurs when a critical vapour velocity is reached, which is shown to be consistent with the limited data on dryout conditions in steady boiling. (orig.)

  9. 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

  10. 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

  11. CEA program on future generation light water modular reactors and gas cooled reactors

    International Nuclear Information System (INIS)

    The CEA programme on 'Future Generation Reactors and Fuel Cycles' aims at studying and developing the mean and long term most promising options for nuclear reactors, fuels and reprocessing. These options should contribute to make the nuclear energy a major source of the sustainable development. The program also aims at maintaining at the highest level of competency the technologies with which the CEA will be able to bring to national achievements or international projects in the next decades, projects whose specifications and calendar are today unknown. These studies on the 'Future Generation Reactors and Fuel Cycles' constitute a field privileged for international collaboration. The corresponding researches are structured in four main axes: Innovations for LWR; Systems of 4th generation; Sodium-cooled reactors; Systems which are the object for survey or exploratory studies. Studies on future nuclear gas technologies are mainly covered by the 4th generation programme (Gen IV). Within this context, the goals pursued, in particular the minimization of the production of long lived waste and the saving of resources (i.e.: the optimised utilisation of fissile and fertile nuclear fuels), could justify an evolution towards hard neutron spectra and high temperatures, to cover applications other than the electricity production, e.g.: hydrogen production, desalination, cogeneration. The main R and D axis for these long-term objectives currently the area of Gas Cooled Reactors (GCR). The corresponding program is structured through eight R and D projects details of which are presented within the paper. (author)

  12. Boiling induced mixed convection in cooling loops

    International Nuclear Information System (INIS)

    This article describes the SUCO program performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. In case of a core melt accident, the sump cooling concept realises a decay heat removal system that is based on passive safety features within the containment. The article gives, first, results of the experiments in the 1:20 linearly scaled SUCOS-2D test facility. The experimental results are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. Second, the real height SUCOT test facility with a volume and power scale of 1:356 that is aimed at investigating the mixed single-phase and two-phase natural circulation flow in the reactor sump, together with first measurement results, are discussed. Finally, a numerical approach to model the subcooled nucleate boiling phenomena in the test facility SUCOT is presented. Physical models describing interfacial mass, momentum and-heat transfer are developed and implemented in the commercial software package CFX4.1. The models are validated for an isothermal air-water bubbly flow experiment and a subcooled boiling experiment in vertical annular water flow. (author)

  13. Outlook for Heavy-Water Reactors

    International Nuclear Information System (INIS)

    For large capacity units heavy-water reactors have already established an economic position competitive with any other nuclear or conventional electric generators in certain quite normal circumstances. With continued development they are expected to maintain this position and contribute a large fraction of the world's power demand into the indefinite future and extend to other large-scale uses such as desalting water. The grounds for this forecast are shown to be firmly based on a low-fuelling cost despite any likely rise in the price of uranium, together with extreme flexibility to adapt the fuel cycle to meet any changes in the market of fuel supply and reprocessing. Appropriate use of thorium with uranium is likely in the long term. As the scale of use expands, the cost of fabricating fuel and of producing heavy water will fall. Very few large heavy-water power reactors have yet been built and as experience is gained the cost of construction will fall. An important near-term development will be the practice of taking heat from the fuel direct to generate the working steam. In the longer term as improved materials are developed for higher temperatures the thermal to electrical conversion efficiency will be raised. With all these improvements in view, a long-term competitive position is foreseen. Details are given of several fuel cycle changes showing the inventories of fuel and heavy water involved in meeting market changes and the expectation of net fuel cycle costs being held in the range 0.3 to 0.6 mill/kWh in terms of to-day's dollar values despite a rise in the cost of uranium. The analysis shows possible interaction with other types of reactor including breeders. Reasons are given for welcoming any possible future competition or partnership from these types in meeting the growing world demand for harnessed energy. (author)

  14. Behaviour of heavy water in nuclear reactors of the CEA

    International Nuclear Information System (INIS)

    In the two heavy water reactors of the CEA: Zoe and P-2, we do: A) the supervision of the isotopic composition of the heavy water; B) the supervision of gases released by the decomposition of the heavy water under radiation, and to their recombination; C) periodic analyses of impurities. (M.B.)

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

    International Nuclear Information System (INIS)

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

  16. 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

  17. 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.)

  18. 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)

  19. Thermo-mechanical behaviour of coolant channels for heavy water reactors under accident conditions

    International Nuclear Information System (INIS)

    The objective of nuclear safety research programme is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off normal conditions. Indian Pressurised Heavy Water Reactors (PHWRs) are tube type of reactors. The coolant channel assemblies, being one of the most important components, need detailed analysis under all operating conditions as well as during postulated conditions of accidents for its thermo-mechanical behaviour. One of the postulated accident scenarios for heavy water moderated pressure tube type of reactors i.e. PHWRs is Loss Of Coolant Accident (LOCA) coincident with Loss Of Emergency Core Cooling System (LOECCS). In this case, even though the reactor is tripped, the decay heat may not be removed adequately due to low or no flow condition and inventory depletion of primary side. Since the emergency core cooling system is presumed to be not available, the cooling of the fuel pins and the coolant channel assembly depends on the moderator cooling system, which is assumed to be available. Moderator cooling system is a separate system in PHWRs. In PHWRs, the fuel assembly is surrounded by pressure tube, an annulus insulating environment and a concentric calandria tube. In this postulated accident scenario, a structural integrity evaluation has been carried out to assess the modes of deformation of pressure tube-calandria tube assembly in a tube type nuclear reactor. The loading of pressure and temperature causes the pressure tube to sag/balloon and come in contact with the outer cooler calandria tube. The resulting heat transfer could cool and thus control the deformation of the pressure tube thus introducing inter-dependency between thermal and mechanical contact behaviour. The amount of heat thus expelled significantly depends on the thermal contact conductance and the nature of contact between the two tubes. Deformation of pressure tube creates a heat removal path to the relatively

  20. 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)

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

    International Nuclear Information System (INIS)

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

  2. Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

  3. Uncertainties treatment in the water-cooled nuclear research reactor-thermal design and analysis

    International Nuclear Information System (INIS)

    This paper describes methods of uncertainties and its calculationprocedures for the water-cooled nuclear research reactor (i. e. WWR-M2) with a 10 MWth, in the thermal design and analysis, where the uncertainties are due to the reactor fuel coolant channel design fabrication defects (fuel meat and clad thickness uncertainties). The results show: (1) The effect of fuel meat and cladding thickness may have great influence on the distribution of the axial temperatures (cladding surface, and fuel centerline) and other parameters in the reactor core and more intense in the reactor thermal design, (2) The selection of new reactor core operating conditions and parameters due to the fuel coolant channel fabrication defects, and (3) Calculation values of the hot spot and temperatures of the WWR-M2 reactor by using different methods.(author)

  4. Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data

    International Nuclear Information System (INIS)

    Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author)

  5. Contribution to the study of the stability of water-cooled reactors

    International Nuclear Information System (INIS)

    This work is devoted to the study of the stability of reactors cooled by water subjected only to natural convection. It is made up of two parts, a theoretical study and experimental work, each of these parts being devoted to a consideration of linear and non-linear conditions: - calculation of the transfer function of the reactor using neutronic and hydrodynamic linear equations with the determination of the instability threshold; - demonstration of the existence of the limiting oscillation cycle in the case of a linear feedback using MALKIN'S method; - measurement and interpretation of the reactor's transfer functions and of the hydrodynamic transfer functions; and - analysis of the noise due to boiling. (author)

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

    International Nuclear Information System (INIS)

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

  7. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yanghyun; Kim, Keonsik; Park, Jeongyong; Yang, Yongsik; Kim, Hyungkyu; In, Wangkee; Song, Kunwoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR.

  8. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    International Nuclear Information System (INIS)

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR

  9. Evolutionary water cooled reactors: Strategic issues, technologies and economic viability. Proceedings of a symposium

    International Nuclear Information System (INIS)

    Symposium on evolutionary water cooled reactors: Strategic issues, technologies and economic viability was intended for managers in utilities, reactor design organizations and hardware manufacturing companies and for government decision makers who need to understand technological advances and the potential of evolutionary water cooled reactors to contribute to near and medium term energy demands. The topics addressed include: strategic issues (global energy outlook, the role of nuclear power in sustainable energy strategies, power generation costs, financing of nuclear plant projects, socio-political factors and nuclear safety requirements); technological advances (instrumentation and control, means od improving prevention and mitigation of severe accidents, development of passive safety systems); keys to economic viability (simplification, standardization, advances in construction and project management, feedback of experience from utilities into new designs, and effective management of plant operation)

  10. The chemistry of the PSGHWR heavy water moderator system

    International Nuclear Information System (INIS)

    The PSGHWR heavy water moderator and helium overblanket circuits, and the heavy water purification plant are described. Details of the deuterisation and dedeuterisation of ion exchange resin charges for the purification plant are given, and a pattern of consistency over seventeen deuterisations of anion and cation beds is shown. The chemistry specification for the moderator system is given, and the changing philosophy of ion exchange plant operation to maintain this specification, from early operation with continuous flow through cation and anion clean up columns, to the current mode of minimum intermittent flow, is explained. Anionic and cationic impurities and radio nuclides found in the moderator are listed, and their sources examined. The inter-relation between the ion exchange plant as a source as well as a remover of impurities and the part played by peroxide is examined. Radiolysis data is presented and an attempt is made to explain the correlation between radiolysis rate and impurity concentrations, ion exchange column operation and other plant parameters. A soluble poison, enriched 10boron, is used for balancing excess fuel reactivity. The boron cycle and method of control of boron concentration are explained, and experience of anion exchange resin capacity for boron at different input concentrations is given. Experience covering 26 regenerations of the anion beds, mostly using deutero chemicals, and details of the method of recovery of deuteroboric acid and potassium deuteroxide by electrodialysis are given. Finally, heavy water losses and isotopic degradation of D2O during 8 years of operation are examined. (author)

  11. Summary of the 3rd workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  12. Operation and Licensing of Mixed Cores in Water Cooled Reactors

    International Nuclear Information System (INIS)

    Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

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

    International Nuclear Information System (INIS)

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

  14. 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)

  15. 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.)

  16. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

  17. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

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

  18. 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

  19. Thermal hydraulic analysis due to the changes in heat removal for advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor is a natural circulation light water cooled and heavy water moderated pressure tube reactor. Changes in heat removal by primary heat transport system of the reactor have significant impact on various important system parameters like pressures, qualities, reactor power and flows. Increase in heat removal leads to the cooldown of the system subsequently reducing pressure, void increase and changes in power and flows of the system. Decrease in heat removal leads to warm-up of the system subsequently raising pressure, void collapse, and changes in power and flows of the system. The behaviour is complex as system under consideration is natural circulation system. This article presents the results of simulations made with the RELAP5-MOD3.2 code that show first the impact of a decrease in feed water temperature on fluid temperature, steam drum pressure, core exit void, reactivity, reactor power, core flow, steam flow and clad temperature and secondly the impact of a loss of normal feed water flow on steam drum pressure, channel flow, core quality, clad surface temperature. For lowering of feed water temperature transient and in isolation condensers cold water injection, the reactor power increases and the reactor trips on the high power signal. Simultaneous flow increment due to the 2 phase natural circulation characteristic has caused the clad temperature to limit to their steady state value. In case of loss of feed transient the reactor trips on high pressure. The clad surface temperature rise from steady state operating value is marginal and it is well within the safety limit as per the acceptance criteria

  20. Good practices in heavy water reactor operation

    International Nuclear Information System (INIS)

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

  1. 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)

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

    International Nuclear Information System (INIS)

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

  3. Technological readiness of evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Nuclear energy has evolved to a mature industry that supplies over 16% of the world's electricity, and it represents an important option for meeting the global energy demands of the coming century in an environmentally acceptable manner. New, evolutionary water cooled reactor designs that build on successful performance of predecessors have been developed; these designs have generally been guided by wishes to reduce cost, to improve availability and reliability, and to meet increasingly stringent safety objectives. These three aspects are important factors in what has been called technological readiness for an expanded deployment of nuclear power; a major increase in utilization of nuclear power will only occur if it is economically competitive, and meets safety expectations. To this end, the industry will also have to maintain or improve the public perception of nuclear power as a benign, economical and reliable energy source. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  5. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  6. Static seals and their application in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    Information relative to six types of static seals commonly used in the primary cooling systems of nuclear reactors is compiled. This information includes a description of each type of seal, its material of construction, design features, operating experience, and advantages and disadvantages. The types covered include spiral-wound asbestos-filled gaskets, hollow metallic O-rings, Belleville spring type of gasketed joints, integrated elastomer and metal retainer gaskets, and solid metal gaskets with heavy cross sections. Omega, canopy, and lip seals are discussed briefly, and information on flange design for gasketing is also presented

  7. Safety implications of an integrated boiling curve model for water-cooled divertor channels

    International Nuclear Information System (INIS)

    The international fusion community is actively researching advanced heat transfer methods for removal of high thermal loads from next-generation divertor assemblies. Such advanced techniques may indeed optimize the operational and economical performance of future divertor designs. However, with its extensive operational database, water-cooling remains as one of the optimum choices for near-term divertor designs. Critical heat flux (CHF) is the maximum heat flux that water, at a given set of inlet conditions, can remove via fully developed nucleate boiling. Accordingly, an accurate CHF calculation is of the utmost importance for maintaining adequate safety margins in divertor operation. This paper uses the integrated boiling curve model developed at Sandia National Laboratories to examine the safety implications of calculating the CHF. In particular, this paper focuses on the influence of the finite element peaking factor (FEPF) that converts the heat flux predicted by CHF correlations into a plasma heat flux that can be measured. The analyses illustrate that the FEPF is proportional to the plasma heat flux and thus accurate calculation of the CHF requires the use of the appropriate FEPF for the given water conditions and plasma heat flux. It is shown that using a geometric peaking factor is inadequate since the true peak factor is dependent upon the plasma heat flux. The conclusion is that a finite element analysis incorporating an integrated boiling curve is required for accurate calculation of the CHF

  8. Thermophysical properties database of materials for light water reactors and heavy water reactors. Final report of a coordinated research project 1999-2005

    International Nuclear Information System (INIS)

    The IAEA Coordinated Research Project (CRP) on the Establishment of a Thermo-physical Properties Database for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs) started in 1999. It was included in the IAEA's Nuclear Power Programme following endorsement in 1997 by the IAEA's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and the TWG-HWR). Furthermore, the TWG on Fuel Performance and Technology (TWG-FPT) also expressed its support. This CRP was conducted as a joint task within the IAEA's project on technology development for LWRs and HWRs in its nuclear power programme. Improving the technology for nuclear reactors through better computer codes and more accurate materials property data can contribute to improved economics of future plants by helping to remove the need for large design margins, which are currently used to account for limitations of data and methods. Accurate representations of thermo-physical properties under relevant temperature and neutron fluence conditions are necessary for evaluating reactor performance under normal operation and accident conditions. The objective of this CRP was to collect and systematize a thermo-physical properties database for light and heavy water reactor materials under normal operating, transient and accident conditions and to foster the exchange of non-proprietary information on thermo-physical properties of LWR and HWR materials. An internationally available, peer reviewed database of properties at normal and severe accident conditions has been established on the Internet. This report is intended to serve as a useful source of information on thermo-physical properties data for water cooled reactor analyses. The properties data have been initially stored in the THERSYST data system at the University of Stuttgart, Germany, which was subsequently developed into an internationally available Internet database named THERPRO at Hanyang University, Republic of Korea

  9. 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)

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

    International Nuclear Information System (INIS)

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

  11. 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

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

    International Nuclear Information System (INIS)

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

  13. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, 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, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  14. Fluid-phase transitions in light water and heavy water

    International Nuclear Information System (INIS)

    Phase transitions are important phenomena in water-cooled nuclear-fission reactors. In this study, liquid-vapour phase transitions and critical-point behaviour are considered. Special attention is given to the van der Waals model, asymptotic critical scaling, thermodynamic potentials and steam properties. Models and data are shown for light water and heavy water. The relevance to advanced Generation III reactors and proposed Generation IV reactors is assessed. (author)

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

    International Nuclear Information System (INIS)

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

  16. 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...

  17. 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 ...

  18. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

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

  19. The possible use of cermet fuel in the DIDO and PLUTO heavy-water research reactors

    International Nuclear Information System (INIS)

    International restrictions on the supply of highly enriched uranium have resulted in the requirement to fuel research reactors with a lower-enrichment uranium fuel. A study has been made of the feasibility of using low-enrichment fuels of a new type in the DIDO and PLUTO reactors. This work has been done as a contribution to the studies currently being carried out internationally on the implications of using lower-enrichment fuels in heavy-water-moderated research reactors. The uranium content of the U/Al alloy at present used cannot be increased sufficiently to maintain the requisite U235 content without undesirable effects on the physical properties of the alloy. A different type of fuel will therefore be required to maintain the desired nuclear characteristics. A possible solution to the problem is the use of a cermet (U3O8/Al) fuel material. Cermet fuel has poorer thermal conductivity than metallic fuel, and may also contain particles of the ceramic of a size that approaches the total thickness of the cermet core. We therefore have to consider both the average temperature of the centre of the fuel and whether large particles of the ceramic may be significantly hotter than the average. This paper describes a preliminary study of the feasibility of this concept from the heat-transfer and safety viewpoints. Calculations have been made for a cermet of 20%-enrichment 2.3g U/cm3, used in a high-power element in a DIDO-type reactor. To accommodate the cermet, the cladding has been reduced in thickness to 0.318mm (0.0125 in) the core increasing to 1.044mm, but the fuel geometry is otherwise unchanged. It is concluded that from the heat-transfer viewpoint there is no problem during normal operation or the maximum credible power transient in these reactors. (author). 10 refs, 6 figs, 2 tabs

  20. 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.)

  1. Modelling containment passive safety systems in advanced water cooled reactors

    International Nuclear Information System (INIS)

    Most designs of advanced passive reactors incorporate Passive Containment Cooling Systems (PCCS) relying on steam condensation to cope with possible pressure increase that would result in the case of a postulated accident. As a consequence, experimental and analytical research programmes have been launched worldwide to investigate new configurations and conditions involved in these new scenarios. This paper summarises the major outcomes of the joint research of CIEMAT, UPV, and UW in developing predictive models to address anticipated conditions in the Simplified Boiling Water Reactors (CIEMAT-UPV) and in the AP600 (CIEMAT-UW). Even though both models share some of their fundamental characteristics (such as being mass/heat transfer analogy based), samples of their validation against independent databases illustrate their intrinsic differences in formulation according to the scenarios addressed by each one. Relative importance of condensate film or gas mixture velocity are discussed, and the effect of key factors such as noncondensable gas presence and pressure are stated. Experimental data from University of Berkeley (UCB) and from University of Wisconsin - Madison (UW) will be used to support comparisons and discussions held in the paper. In short, this work demonstrates that heat/mass transfer analogy-based models, particularly those relying on diffusion film modelling to account for noncondensable gas presence, are extremely useful in test interpretation and result in good agreement with reliable databases. (author)

  2. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  3. 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

  4. 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

  5. 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

  6. 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

  7. Designing a plant for tritium and protium removal from heavy water of the PIK reactor

    International Nuclear Information System (INIS)

    The project of the facility for tritium and protium isotopic removal (FIR) from the PIK reactor heavy water is described. The FIR principle scheme, wherein different methods for isotopes separation and management of deuterium, contaminated by tritium are applied, is presented. The experimental facilities created for obtaining the initial data required for the FIR project development are described. In particular, the facility for the hydrogen isotopes separation is created on the basis of the isotope exchange between water and hydrogen and water electrolysis

  8. Summary of the 4th workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  9. Summary of the 4th workshop on the reduced-moderation water reactor

    International Nuclear Information System (INIS)

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  10. 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

  11. Toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

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

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼21 Btu/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

  13. 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)

  14. Heavy water reactors: Status and projected development. Part I. Final draft of a report to be published in the IAEA technical reports series. Working material

    International Nuclear Information System (INIS)

    retains a low cost operational condition; to illustrate the short and medium term potential for design evolution of the heavy water reactor type; to describe the basis of economic competitiveness of the HWR, its resistance to severe cost increases and capability for extensive source localization; and to provide a reference document on HWRs and to help guide the activities of the IWG-HWR. Those organizations developing and operating HWRs recognize the potential for development of this line of reactors, and it is the intent of this document to illustrate that potential. Various countries and organizations in the past have explored a number of variants of heavy water reactors, and there is a desire to continue to explore some of these options in the future. Currently the pressurized heavy water cooled, heavy water moderated design is an economically competitive design and will likely continue to dominate the heavy water reactor type for some time. This document concentrates on heavy water moderated reactors for electricity production. Reactors for district heating and research reactors are not discussed, except where historical multi-purpose use was a rationale for developing the concept

  15. Heavy water reactors: Status and projected development. Part II. Final draft of a report to be published in the IAEA technical reports series. Working material

    International Nuclear Information System (INIS)

    retains a low cost operational condition; to illustrate the short and medium term potential for design evolution of the heavy water reactor type; to describe the basis of economic competitiveness of the HWR, its resistance to severe cost increases and capability for extensive source localization; and to provide a reference document on HWRs and to help guide the activities of the IWG-HWR. Those organizations developing and operating HWRs recognize the potential for development of this line of reactors, and it is the intent of this document to illustrate that potential. Various countries and organizations in the past have explored a number of variants of heavy water reactors, and there is a desire to continue to explore some of these options in the future. Currently the pressurized heavy water cooled, heavy water moderated design is an economically competitive design and will likely continue to dominate the heavy water reactor type for some time. This document concentrates on heavy water moderated reactors for electricity production. Reactors for district heating and research reactors are not discussed, except where historical multi-purpose use was a rationale for developing the concept

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

    International Nuclear Information System (INIS)

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

  17. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  18. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  19. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Appendices. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

    Appendices are presented concerning the evaluations of decommissioning financing alternatives; reference site description; reference BWR facility description; radiation dose rate and concrete surface contamination data; radionuclide inventories; public radiation dose models and calculated maximum annual doses; decommissioning methods; generic decommissioning information; immediate dismantlement details; passive safe storage, continuing care, and deferred dismantlement details; entombment details; demolition and site restoration details; cost estimating bases; public radiological safety assessment details; and details of alternate study bases.

  20. Local heat transfer from the corium melt pool to the boiling water reactor pressure vessel wall

    International Nuclear Information System (INIS)

    The present study considers in-vessel accident progression after core melt relocation to the lower head of a Boiling Water Reactor (BWR) and formation of a melt pool containing a forest of Control Rod Guide Tubes (CRGTs) cooled by purging flows. Descending streams of melt that flow along cooled surfaces of CRGT, and impinge on the bottom surface of the vessel wall can significantly increase local heat transfer. The area of enhanced heat transfer enlarges with decreasing of the melt Prandtl (Pr) number, while the peaking value of the heat transfer coefficient is a non-monotone function of Pr number. The melt Pr number depends on the melt composition (fractions of metallic and oxidic melt components) and thus is inherently uncertain parameter of the core melting and relocation scenarios. The effect of Pr number in the range of 1.02 - 0.03 on the local and integral thermal loads on the vessel wall is examined using Computational Fluid Dynamics (CFD). Heat transfer models obtained on the base of CFD simulations are implemented in the Phase-change Effective Convectivity Model (PECM) for simulation of reactor-scale accident progression heat transfer in real 3D geometry of the BWR lower plenum. We found that the influence of the low Pr number on the thermal loads in a big melt pool becomes more significant at later time, than rapid acceleration of the creep in the vessel wall. This result suggests that global vessel failure is insensitive to the melt composition in the considered 0.7 m deep melt pool configuration. However, it is not clear yet if the low Pr number effect has an influence on vessel failure mode in the other possible melt pool configurations. (author)

  1. Indian heavy water programme - challenges and opportunities

    International Nuclear Information System (INIS)

    Discovery of fission of uranium in 1939 opened up hitherto unknown possibilities for utilising the fission energy for use of mankind, mainly for the production of and electrical energy. It was realised that this nuclear energy could be an ideal substitute for the fast depleting fossil fuels which would one day get exhausted. Two main concepts of nuclear power reactor got evolved, one enriched uranium fuelled, ordinary water moderated reactor and another natural uranium fuelled heavy water moderated reactor. The concentration of uranium 235U needed for ordinary water moderated reactors is 3% but the naturally occurring uranium in India contains only 0.7% of 235U. The reactors utilising natural uranium as fuel require Heavy Water as moderator. The processing of uranium ore to achieve from 0.7% to 3% is highly complex. Recognising the fact that India has limited uranium resources but rich thorium resources, Dr. Bhabha formulated a three stage nuclear power generation programme for our country. The first generation reactors can use natural uranium as fuel with heavy water as moderator. Since the technology to generate such large scale heavy water to match the urgent need for nuclear power generation was not indigenously available, the technology available with Canada and France was utilised for installation of first generation heavy water plants in India. However, the peaceful nuclear experiment conducted by India in 1974 caused resentment among the countries that supplied Heavy Water technology to India and they stopped all technological help and assistance in nuclear field. Thereafter, it was the story of India going alone in heavy water production. That made India meets successfully all challenges on the way to installation, commissioning and sustained operation of all plants. Today we have six operating Heavy Water plants, spread all over the country. We have reached a stage, a change from a situation of crunch to a level of not only self sufficiency but to a state

  2. Dry deposition of 88Rb and 138Cs from a boiling water reactor plume

    International Nuclear Information System (INIS)

    Double tracer experiments were made in May 1981 at the Ringhals nuclear power plant in Sweden to investigate atmospheric-dispersion and dose models. Sulphurhexafluoride (SF6) and radioactive noble gases were released simultaneously from a 110-m stack and detected downwind at distances of 3-4 km. The experiments were made under near-neutral conditions. One-hour measurements at ground level yielded cross-wind profiles of SF6 concentrations and gamma radiation from the plume. In-situ gamma spectrometric measurements demonstrated a significant surplus of gamma rays from the noble gas daughters (88Rb and 138Cs) compared with those from the noble gases. This surplus was interpreted as due to dry deposition from the plume, and deposition velocities were estimated at 0.02-0.10 m s-1. These values are very high when compared with values recommended for calculating consequences of nuclear accidents. The high values are believed to be due to the very small size of the daughter particles

  3. AHWR integral test loop scaling philosophy and system description

    International Nuclear Information System (INIS)

    The Indian Advanced Heavy Water Reactor (AHWR) is pressure tube type, boiling light water cooled and heavy water moderated reactor incorporating many passive safety features. To experimentally validate the performance of various inherent safety features and engineering concepts an Integral Test Loop (IIL) has been designed based on the power-to-volume scaling philosophy. The IIL simulated the primary heat transport (PHT) and emergency core cooling system (ECCS) along with the associated control system or AHWR. This report presents the scaling philosophy followed for IIL and detailed description of the system and its various components. (author)

  4. Main control panel design and simulator in advanced boiling water reactor [ABWR

    International Nuclear Information System (INIS)

    The ABWR type main control panel has been developed to enhance the reliability of plant monitoring and operation. This panel consists of large display panels and a compact main console, which enables operators to work from their seated position. Primary plant and system status information is presented on large display panels so that the information can be shared by the entire operating crew. This panel is applied to Kashiwazaki-Kariwa Nuclear Power Station Unit 6 which is scheduled to begin commercial operation in 1996. The ABWR type main control panel is quite different from the conventional one. Consequently the full-scope ABWR simulator is under construction, which will be completed in 1994, so that the operators will be trained sufficiently. (author)

  5. Decontamination of the heavy water system of the RA reactor

    International Nuclear Information System (INIS)

    The heavy water system of the RA reactor was decontaminated of 60Co. The solution used for decontamination was 7% H3PO4 and 3% CrO3. The decontamination factor ranged from 10 to 100. From the results the distribution of 60Co in the heavy water, and on stainless and aluminium parts was determined (author)

  6. Tritium in heat transport and moderator systems of CANDU reactors

    International Nuclear Information System (INIS)

    The production rates of tritium in the heavy-water moderator and heat transport systems of CANDU reactors are calculated from the neutron fluxes generated in reactor-physics analyses of the lattice cell and radiation-physics analyses of the primary radial shields. The concentrations of tritium activity in the heavy water can then be calculated assuming a simple build-up of a decaying radioactive species. This simple treatment has been compared with tritium concentrations measured in the domestic CANDU 6 stations. These comparisons show that the predicted concentrations need to take account of the heavy water management practices of the stations as well as the operating history of the plant. The success of the individual station at segregating the heat transport and moderator heavy water systems also has some impact on the tritium concentrations in the heat transport systems. There is some evidence to show that there are different levels of success at achieving separation. The current build-up of tritium in the heavy water systems shows no evidence for significant tritium production from helium-3, the decay product of tritium, via the 3He(n,p)3H-reaction. The paper introduces explanations for the absence of this effect. The agreement between the predicted concentrations and the measured concentrations is an indirect validation of the thermal fluxes calculated in the lattice and radiation physics codes. The agreement suggests that a comparison between predicted and measured tritium levels in the heavy water systems would allow an operator to monitor the success of the plant at maintaining segregation between the heavy water systems. The paper presents details of the flux levels used to predict tritium production, typical rates of heavy water loss from the plant and the operating histories over a 15-year period. The paper indicates how heavy water recoveries having tritium concentrations between those of the heat transport and moderator systems affect the tritium

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

    International Nuclear Information System (INIS)

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

  8. 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

  9. 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.

  10. Thermal-hydraulics and safety concepts of supercritical water cooled reactors

    International Nuclear Information System (INIS)

    The paper summarizes the status of safety system development for supercritical water cooled reactors and of thermal-hydraulic codes needed to analyze them. While active safety systems are well understood today and expected to perform as required, the development of passive safety systems will still need further optimization. Depressurization transients have successfully been simulated with some codes by a pseudo-two-phase flow simulation of supercritical water. Open issues of thermal-hydraulic codes include modeling of deteriorated heat transfer in one-dimensional system codes and predictions of heat transfer during depressurization transients from supercritical to sub-critical conditions. (author)

  11. The human factor and the part it plays in failures of the normal operation of water-cooled, water-moderated WWER-440 reactors at the Bohunice nuclear power plant

    International Nuclear Information System (INIS)

    The paper presents the findings of a study of the role of the human factor based on the results of the first four years of operation of the Bohunice B-1 nuclear power plant. It describes the method by which plant personnel are trained and the system of maintaining the level of staff skills. It is expected that there will be an improvement in the quality of personnel training and that an analysis of the role of the human factor will be made in the course of subsequent operation. (author)

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

    International Nuclear Information System (INIS)

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

  13. Supercritical water-cooled nuclear reactors: NPP layouts and thermal design options of pressure channels

    International Nuclear Information System (INIS)

    Research activities are currently underway worldwide to develop generation IV nuclear reactor concepts with the objective of improving thermal efficiency and increasing economic competitiveness of generation IV nuclear power plants (NPPs) compared to modern NPPs. The supercritical water-cooled reactor (SCWR) concept is one of six generation-IV options chosen for further investigation and development in many countries worldwide, including Canada. Water-cooled reactors operating at subcritical pressures (7 - 16 MPa) have provided significant electricity production for the past 50 years. However, the thermal efficiency of current NPPs is not very high (30 - 35%). As such, more competitive designs with higher thermal efficiencies, close to those of modern supercritical (Sc) thermal power plants (45 - 50%), need to be developed and implemented. Previous studies have shown that direct cycles with single-reheat and no-reheat configurations are the best options for an SCWR concept. There are a few technical challenges associated with the single-reheat and no-reheat supercritical water (SCW) NPP configurations. The single-reheat cycle requires nuclear steam-reheat, thus increasing the complexity of the reactor core design. Conversely, the major technical challenge associated with an Sc no-reheat turbine is high moisture content in the low-pressure-turbine exhaust. The SCWR-core concept investigated in this paper is based on a generic pressure-tube (pressure-channel) reactor with a 43-element bundle string cooled with supercritical water. The considered 1200-MW el reactor has the following operating parameters: pressure of 25 MPa and reactor inlet/outlet temperatures of 350/625 C. Previous studies have shown that is uranium dioxide (UO2) is used, the fuel centerline temperature might exceed the industry accepted limit of 1850 C. Therefore, this paper investigates a possibility of using uranium carbide (UC), uranium nitride (UN), uranium dicarbide (UC2), uranium dioxide plus

  14. Piping installation for reactor heavy water system

    International Nuclear Information System (INIS)

    Characteristics and main installation steps for the piping of the reactor heavy water loop system were introduced in this paper. According to the system design, equipment accommodation and spot management, main issues with effect on the quality and schedule of pipeline installation were analyzed. Accordingly, some solutions were put forward, which included: work allocation should be made clear in documents; installation preparative such as design checkup and technology communication should be prepared completely; requirements of system cleaning, test items in every experiment, inspection in work and equipment maintenance should be considered in the system design; perfect documents distribution system and stock plan should be built; technology requirements and quality assurance should be claimed in contracts; quality should be controlled by way of external evidence, inspection in manufactory, exterior quality assurance examination, and test during consignment; series of management procedure should be established in detail. (authors)

  15. Innovative concept of Reduced-Moderation Water Reactor (RMWR) for effective fuel utilization through recycling

    International Nuclear Information System (INIS)

    Full text: An innovative water-cooled reactor concept named Reduced-Moderation Water Reactor (RMWR) is under development by JAERI in cooperation with some Japanese utilities and vendors. The reactor aims at achievement of a high conversion ratio more than 1.0 with plutonium (Pu) mixed oxide (MOX) fuel, based on the well-experienced water-cooled reactor technology. Such a high conversion ratio can be attained by reducing the moderation of neutrons, i.e. reducing the water fraction in the core, and is favorable to realize long-term energy supply by effective utilization of the uranium resources, multiple recycling of Pu, or high burn-up / long operation cycle achievement. The reduced neutron moderation with the water results in a similar neutron spectrum to that in a sodium-cooled fast breeder reactor (FBR) even in a water-cooled reactor core. Another important design target for the RMWR is to achieve the negative void reactivity coefficient. This is one of the important characteristics of the currently operated light water reactors, especially from the safety point of view. However, the negative void reactivity coefficient and the high conversion ratio are in the trade-off relation in the reactor design and this gives difficulty to be overcome in the design of the RMWR. Up to the present, we have succeeded in proposing several types of basic design concepts satisfying both the main design targets under both the boiling water reactor (BWR) type concept and the pressurized water reactor (PWR) type one. The common design characteristics are the tight-lattice fuel rod configuration and the short core. The former is to attain the high conversion ratio and the latter is for the negative void reactivity coefficient. Additionally, the axial, i.e. upper, lower or internal, or the radial blankets made of the depleted UO2 are also introduced by necessity for both purposes mentioned above. Since the RMWR is intended to be operated in the fuel cycle with the multiple recycling

  16. Draft layout, containment and performance of the safety system of the European Supercritical Water-Cooled Reactor

    International Nuclear Information System (INIS)

    In Europe, the research on Supercritical Water-Cooled Reactors is integrated in a project called 'High Performance Light Water Reactor Phase 2' (HPLWR Phase 2), co-funded by the European Commission. Ten partners and three active supporters are working on critical scientific issues to determine the potential of this reactor concept in the electricity market. Close to the end of the project the technical results are translated into a draft layout of the HPLWR. The containment and safety system are being explained. Exemplarily, a depressurization event shows the capabilities of the safety system to sufficiently cool the reactor by means of a low pressure coolant injection system. (author)

  17. 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

  18. 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

  19. Technical meeting on advanced fuel pellet materials and fuel rod designs for water cooled reactors. Book of abstracts

    International Nuclear Information System (INIS)

    Heavy Water-Cooled Reactors (TWGLWR and TWGHWR) with a proposal to hold it at the Paul Scherrer Institute, Switzerland. The purpose of the meeting is to provide an overview on the status and perspective of fuel pellet materials development and recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting will cover both light and heavy water reactor fuels with the following main objectives: - Consideration of modern technological and design tools enabling reliable performance of fuels and rod columns in current and planned operational environments; - Analysis of high burnup fuel structure and properties, including RIM effects, thermal behaviour, fission gas release, PCI and PCMI; - Discussion on specific features of MOX fuel, as well as perspectives on advanced fuels like Vibro-pack, Thorium fuel and others. Each of the papers in this book of abstracts has been indexed separately

  20. Analysis of Gamma Activity of Heavy Water at RB Reactor

    International Nuclear Information System (INIS)

    The RB experimental nuclear reactor still works with heavy water obtained in 1959 from the former USSR. Gamma activity of the heavy water was periodically controlled during the past time. In this experiment, measurements were carried out with two samples: D2O taken from the RB reactor and D2O that has been used in the reactor. Two germanium spectrometers were used as detectors. Gamma spectra data were evaluated manually and using several computer codes. results of the experiment show that gamma activity D2O of RB reactor is at the level of background in the Vinca Institute, without contamination with fission products. (author)

  1. Simulation and control of water-gas shift packed bed reactor with inter-stage cooling

    Science.gov (United States)

    Saw, S. Z.; Nandong, J.

    2016-03-01

    Water-Gas Shift Reaction (WGSR) has become one of the well-known pathways for H2 production in industries. The issue with WGSR is that it is kinetically favored at high temperatures but thermodynamically favored at low temperatures, thus requiring careful consideration in the control design in order to ensure that the temperature used does not deactivate the catalyst. This paper studies the effect of a reactor arrangement with an inter-stage cooling implemented in the packed bed reactor to look at its effect on outlet temperature. A mathematical model is developed based on one-dimensional heat and mass transfers which incorporate the intra-particle effects. It is shown that the placement of the inter-stage cooling and the outlet temperature exiting the inter-stage cooling have strong influence on the reaction conversion. Several control strategies are explored for the process. It is shown that a feedback- feedforward control strategy using Multi-scale Control (MSC) is effective to regulate the reactor temperature profile which is critical to maintaining the catalysts activity.

  2. 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)

  3. Advanced concept of reduced-moderation water reactor (RMWR) for plutonium multiple recycling

    International Nuclear Information System (INIS)

    An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

  6. A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  7. A passive emergency heat sink for water cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. Their effective heat transport combines with the large heat-transfer coefficients of tube banks, thereby reducing containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  8. 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

  9. 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)

  10. Experimental investigations and seismic analysis for benchmark study of 1000 MW WWER type units (water-cooled and moderated reactor) nuclear power plant Kozloduy. Final report from 1 November 1994 - 31 October 1995

    International Nuclear Information System (INIS)

    This final report summarizes the research carried out and the results obtained presented in previous progress reports. In includes the description of the research carried out, namely, definition of terms and references, criteria, design parameters and methods of analysis; development of 3-dimensional and 2-dimensional mathematical models for diesel generator stations concerning seismic analyses; supports and anchorages analyses and seismic capacity estimation for main components of the cooling system; methodology for verification of component anchorages; conclusions drawn concerning the seismic safety of Kozloduy NPP