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Sample records for pwr type reactors

  1. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  2. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  3. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  4. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Ailloud, Jean; Monteil, Marcel.

    1978-01-01

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

  5. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  6. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  7. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    2008-12-01

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  8. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  9. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  10. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  11. Method of stopping operation of PWR type reactor

    International Nuclear Information System (INIS)

    Ueno, Takashi; Tsuge, Ayao; Kawanishi, Yasuhira; Onimura, Kichiro; Kadokami, Akira.

    1989-01-01

    In PWR type reactors after long period of l00 % power operation, since boiling is caused in heat conduction pipes and water is depleted within the intergranular corrosion fracture face in the crevis portion to result in a dry-out state, impregnation and concentration of corrosion inhibitors into the intergranular corrosion fracture face are insufficient. In view of the above, the corrosion inhibitor at a high concentration is impregnated into the intergranular corrosion fracture face by keeping to inject the corrosion inhibitor from l00 % thermal power load by way of the thermal power reduction to the zero power state upon operatioin shutdown. That is, if the thermal power is reduced to or near the 0 power upon reactor shutdown, feedwater in the crevis portion is put to subcooled state, by which the steam present in the intergranular corrosion fracture face are condensated and the corrosion inhibitor at high concentration impregnated into the crevis portion are penetrated into the intergranular corrosion fracture face. (K.M.)

  12. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  13. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  14. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  15. Problems of control of WWER-type pressurized water reactors (PWR's)

    International Nuclear Information System (INIS)

    Drab, F.; Grof, V.

    1978-01-01

    The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

  16. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  17. Programme of hot points eradication (Co-60) led on French PWR type reactors

    International Nuclear Information System (INIS)

    Rocher, A.; Ridoux, P.; Anthoni, S.; Brun, C.

    1998-01-01

    The question of hot points (pellets rich in cobalt 59 or in cobalt 60 in a PWR type reactor), is studied from the radiation protection point of view. The purpose is to see how to optimize the radiation protection, the elimination of these hot points can bring an improvement. (N.C.)

  18. New generation nuclear power units of PWR type integral reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

    1997-01-01

    Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

  19. Development of a computer code for transients simulation in PWR type reactors

    International Nuclear Information System (INIS)

    Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

    1981-01-01

    A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

  20. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  1. Natural vibrations of a core banel of a PWR type reactor by elements of revolution shell

    International Nuclear Information System (INIS)

    Barcellos, C.S. de.

    1980-01-01

    Aim to estimate the behavior of the cove barrel of PWR type reactors, submitted to several load conditions, their dynamic characteristic, were determined. In order to obtain the natural modes and frequencies of the core barrel, the CYLDYFE comprete code based in the finite element method, was developed. The obtained results are compared with results obtained by other programs such as SAP, ASKA and STRUDL/DYNAL and by other analytical methods. (M.C.K.) [pt

  2. Automatic welding processes for reactor coolant pipes used in PWR type nuclear power plant

    International Nuclear Information System (INIS)

    Hamada, T.; Nakamura, A.; Nagura, Y.; Sakamoto, N.

    1979-01-01

    The authors developed automatic welding processes (submerged arc welding process and TIG welding process) for application to the welding of reactor coolant pipes which constitute the most important part of the PWR type nuclear power plant. Submerged arc welding process is suitable for flat position welding in which pipes can be rotated, while TIG welding process is suitable for all position welding. This paper gives an outline of the two processes and the results of tests performed using these processes. (author)

  3. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  4. Nonlinear punctual dynamic applied to simulation of PWR type reactors

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-01-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)

  5. Dynamic power behavior of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Moreira, F.J.

    1984-01-01

    A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

  6. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  7. Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

    International Nuclear Information System (INIS)

    Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

    1979-01-01

    Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

  8. Contribution to the study of the conversion PWR type reactors to the thorium cycle

    International Nuclear Information System (INIS)

    Martins Filho, J.R.

    1980-01-01

    The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author) [pt

  9. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Belem, J.A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

  10. A calculation methodology applied for fuel management in PWR type reactors using first order perturbation theory

    International Nuclear Information System (INIS)

    Rossini, M.R.

    1992-01-01

    An attempt has been made to obtain a strategy coherent with the available instruments and that could be implemented with future developments. A calculation methodology was developed for fuel reload in PWR reactors, which evolves cell calculation with the HAMMER-TECHNION code and neutronics calculation with the CITATION code.The management strategy adopted consists of fuel element position changing at the beginning of each reactor cycle in order to decrease the radial peak factor. The bi-dimensional, two group First Order perturbation theory was used for the mathematical modeling. (L.C.J.A.)

  11. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  12. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1994-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  13. Sensitivity analysis on hot channel of PWR type reactors using matricial formalism

    International Nuclear Information System (INIS)

    Maciel, Edisson Savio G.; Andrade Lima, Fernando Roberto de; Lira, Carlos Alberto B.O.

    1995-01-01

    The matricial formalism of the perturbation theory is applied in a simplified model to study the hot channel of PWR reactors. Mass, linear momentum and energy conservation equations and appropriated heat transfer and fluid mechanics correlations describe the discretized system. After calculating system's thermalhydraulic properties, the matricial formalism is applied and the sensitivity coefficients are determined for each case of interest. Comparisons between perturbative method and direct results of the model have shown good agreement which demonstrates that the matricial formalism is an important tool for discretized system analysis. (author). 6 refs, 2 tabs

  14. Study of crack propagation velocity in steel tanks of PWR type reactor

    International Nuclear Information System (INIS)

    Amzallac, C.; Bernard, J.L.; Slama, G.

    1983-05-01

    Description and results of a serie of tests carried out on crack propagation velocity of steels in PWR environment (pressurized high temperature water), in order to examine the effects of metallurgical parameters such as chemical composition of steel, especially sulfur and carbon content, and steel type (laminate or forged steels), effects of mechanical parameters such as loading ratio, cycle form, frequency and application mode of loads and of chemical parameters (anodal dissolution or fatigue with hydrogen) [fr

  15. Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

    Science.gov (United States)

    Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

    2018-02-01

    Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

  16. An intelligent pedagogic tool for teaching the operators of PWR type reactors

    International Nuclear Information System (INIS)

    Cordier, B.; Guillermard, M.

    1990-01-01

    A tool was developed for assisting the instruction of the operators of a PWR type nuclear power plant. For achieving the objectives, an expert system and a simulator were combined. The main objective of the system is to improve the work of the operators in performing remedial actions in case of accident. The simulator applies two IBM PC AT3 and a MC 680 20 microprocessor. The use and the validation of the expert system are presented. The perspectives for the system, implanted on the Tricastin nuclear power plant, are analyzed [fr

  17. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Robbe, M.F.

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  18. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1985-02-01

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation [fr

  19. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor; Modelo simplificado para simulacao do comportamento termohidraulico do canal quente de reator nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Belem, J A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs.

  20. Preventive testing and leakage detection in pipe-lines of steam condensers and generators of a PWR type reactor

    International Nuclear Information System (INIS)

    Canalini, A.; Carvalho, N.C. de

    1985-01-01

    The non-destructive methods: Spum, Helium and Hydrostatic used in leakage detection in condenser pipelines for PWR type reactors are presented. The time, costs, sensitivity, resources necessary and personnel development factors are considered to choose adequated method, in function of nuclear power plant conditions. The leakage tests are applied in pressurized systems or vacuum. Eddy Current testing is used in condensers and steam generators aiming to avoid leakage in these equipments. The spume testing for leakage detection in condenser pipelines - which operation - and hydrostatic testing for leakage detection through reaming with shutdown - were most efficients. The Helium testing applied in pressurized systems or submitted to vacuum systems presented satisfactory results. The Eddy Current testing in condenser and steam generator pipelines reached desired objective, reducing leakage in the first and preserving the integrity in the second. (M.C.K.) [pt

  1. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  2. Study of behaviour of radioactive iodine inorganic compounds in PWR type reactor loops

    International Nuclear Information System (INIS)

    Alm, M.; Johannsen, K.-H.; Dreyer, R.

    1980-01-01

    Compounds of radioactive iodine and its distribution between water and vapour depending on temperature, pressure and water regime of reactor coolant with water under pressure are investigated. The field of variation of parameters indicated is widened as compared with operating reactor parameters (pressure 2-14 MPa, temperature 210-335 deg C). Distribution of iodine compounds has been studied by a statistical method. For WWER-type reactors the following conclusions have been drawn: radioactive iodine in water and vapor in the first and second loops exists in the form of iodide, radioactive iodine concentration in water vapour at constant temperature and pressure mainly is depended on water pH value, radioactive iodine solubility in water vapor at normal parameters of the reactor first loop can be approximately calculated by the equation: Ksub(d)=Csub(g)/Csub(l)=(rhosub(g)/rhosub(l))sup(2), where Ksub(d) is a coefficient of solid distribution between water and vapour, rho is density c is concentration [ru

  3. PWR type reactor equipped with a primary circuit loop water level gauge

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro.

    1990-01-01

    The time of lowering a water level to less than the position of high temperature side pipeway nozzle has been rather delayed because of the swelling of mixed water level due to heat generation of the reactor core. Further, there has been a certain restriction for the installation, maintenance and adjustment of a water level gauge since it is at a position under high radiation exposure. Then, a differential pressure type water level gauge with temperature compensation is disposed at a portion below a water level gauge of a pressurizer and between the steam generator exit plenum and the lower end of the loop seal. Further, a similar water level system is disposed to all of the loops of the primary circulation circuits. In a case that the amount of water contained in a reactor container should decreased upon occurrence of loss of coolant accidents caused by small rupture and stoppage of primary circuit pumps, lowering of the water level preceding to the lowering of the water level in the reactor core is detected to ensure the amount of water. Since they are disposed to all of the loops and ensure the excess margin, reliability for the detection of the amount of contained water can be improved by averaging time for the data of the water level and averaging the entire systems, even when there are vibrations in the fluid or pressure in the primary circuit. (N.H.)

  4. Analytical and sampling problems in primary coolant circuits of PWR-type reactors

    International Nuclear Information System (INIS)

    Illy, H.

    1980-10-01

    Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

  5. Use of standard spectra for the short life radionuclides and ratios for long life radionuclides in the wastes of EDF PWR type reactors

    International Nuclear Information System (INIS)

    Lantes, B.; Bienvenu, Ph.

    2001-01-01

    This paper presents the type of declaration of radioactivity in the wastes of PWR type reactors park. Particularly, it insists on the justification of use of spectra for the declaration of short live radionuclides. It tackles the important developments of methods and measures of radiochemical analysis made by the Cea in order to determine the ratios to declare the long life radioisotopes. (N.C.)

  6. Re-irradiation and limit testing of the fuels PWR type reactors

    International Nuclear Information System (INIS)

    Roche, M.; Molvault, M.

    1978-01-01

    In view of investigating the neutron radiation behavior of PWR fuel pins, the S.P.S. (Services des Piles de Saclay) developed a set of experimental means used at OSIRIS in Saclay Nuclear Research Center. Said devices are shown to be able to meet present problems concerning can failures, power and temperature cycling, remote-control studies. These means can also be used either for statistical studies, they can then receive several samples, or for analytical studies in instrumented devices of large capacity and accelerated irradiation rate [fr

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

    International Nuclear Information System (INIS)

    Esteves, M.M.

    1985-01-01

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

  8. Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor

    International Nuclear Information System (INIS)

    Braga, C.V.M.; Carajilescov, P.

    1981-01-01

    Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt

  9. Results of theoretical studies on power noise of PWR type reactors

    International Nuclear Information System (INIS)

    Meyer, K.

    1979-12-01

    For evaluating noise measurements of neutron flux in power reactors position-dependent effects are taken into account. Based on an effective one-group diffusion model results are obtained for reactor noise due to vibrating control elements and disturbances in the coolant flow

  10. Velocity of crack growing of Inconel-600, sensitized, contaminated with sulphur in PWR type reactors

    International Nuclear Information System (INIS)

    Castano, M. L.; Blazquez, F.; Gomez Briceno, D.; Lagares, A.

    1998-01-01

    The origin of the vessel head penetration cracking of Jose Cabrera NPP has been attributed to an IGA/SCC process in a highly sensitized Alloy 600 assisted by sulphur species, as both acid sulphates and reduced species originated by the thermal breakdown of the cationic resins present in the primary coolant. The thermal degradation of the cationic resins leads sulphonic acid group scission and sulphates. Under the operating conditions the reduction of sulphates to sulphides is produced. The sulphides formed from the reduction of sulphate can precipitate with metallic cations and be incorporated into the oxide layers of the materials, preferably into nickel alloys. Others components at Jose Cabrera NPP are fabricated from sensitized alloy 600, as bottom vessel penetrations. In order to determine the influence of sulphur incorporated to the oxide layers of bottom vessel penetration alloy 600, an experimental work has been performed to obtained crack growth rate data under PWR primary conditions on sensitized alloy 600. (Author) 5 refs

  11. Study of coolant flow distribution within the PWR type reactor vessel

    International Nuclear Information System (INIS)

    Eberle, L.M.M.

    1983-01-01

    The thermohydraulic design of a pressurized water reactor requires the determination of the coolant flow distributions within the reactor vessel, particulary at the core inlet. In this work it is proposed the study of this flow, using potencial flow theory governed by Laplace's equation, nabla 2 φ = O. The solution of the potential field is obtained by the finite element method, which simplifies considerably the treatment of complex geometrical configurations. The equation is solved by the finite element computer code ANSYS, developed and licensed for structural and thermal analysis by using the analogy between steady state heat transfer equation without heat generation, nabla 2 T=O, and Laplace's equation of the velocity potential. The proposed method has been applied to a commercial reactor, and the results are consistent with the available experimental data. (author) [pt

  12. Modelling and simulation the radioactive source-term of fission products in PWR type reactors

    International Nuclear Information System (INIS)

    Porfirio, Rogilson Nazare da Silva

    1996-01-01

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  13. Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

    International Nuclear Information System (INIS)

    Edy-Sulistyono

    1996-01-01

    Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

  14. High temperature filtration of radioactivable corrosion products in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Dolle, L.

    1976-01-01

    A effective limitation to the deposition of radioactive corrosion products in the core of a reactor at power operation, is to be obtained by filtering the water of the primary circuit at a flow rate upper than 1% of the coolant flow rate. However, in view of accounting for more important release of corrosion products during the reactor start-up and also for some possible variations in the efficiency of the system, it is better that the flow rate to be treated by the cleaning circuit is stated at 5%. Filtration must be effected at the temperature of the primary circuit and preferably on each loop. To this end, the feasibility of electromagnetic filtration or filtration through a deep bed of granulated graphite has been studied. The on-loop tests effected on each filter gave efficiencies and yields respectively upper than 90% and 99% for magnetite and ferrite particles in suspension in water at 250 deg C. Such results confirm the interest lying in high temperature filtration and lead to envisage its application to reactors [fr

  15. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  16. Modification method for increasing inner capacity of PWR type reactor container

    International Nuclear Information System (INIS)

    Toda, Shiro; Matsumoto, Mitsutaka

    1998-01-01

    A reactor container has a cylindrical wall portion and a dome-like ceiling portion. The upper portion of the cylindrical wall portion where the cross section is gradually reduced is cut off to from a circular step portion. A steel plate having an arcuate cross section having an inner surface being in contact with the outer circumferential surface of the ceiling portion is placed on the annular step portion, joined by welding to form a raised cylindrical portion. A raised ceiling portion having the same shape as the ceiling portion is constituted by weld-assembling of steel plates on the upper portion of the raised cylindrical portion. Through holes for air ventilation are perforated on the ceiling portion, and concrete is placed on the outer sides of the cylindrical wall portion, the raised cylindrical portion and the raised ceiling portion to form a concrete layer. Since the raised cylindrical portion and the raised ceiling portion can thus be constituted with no access to the inside of an existent reactor container, radiation exposure can be avoided. (I.N.)

  17. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  18. The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

    International Nuclear Information System (INIS)

    Huber, Horacio

    1989-01-01

    Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

  19. Parameter definition for reactor physics calculation of Obrigheim KWO PWR type reactor using the Gels and Erebus codes

    International Nuclear Information System (INIS)

    Faya, A.G.; Nakata, H.; Rodrigues, V.G.; Oosterkamp, W.J.

    1974-01-01

    The main variables for Obrigheim Reactor - KWO diffusion theory calculations, using the EREBUS code were defined. The variables under consideration were: mesh spacing for reactor description, time-step in burn-up calculation, and the temperature in both the moderator and the fuel. The best mesh spacing and time-step were defined considering the relative deviations and the computer time expended in each case. It has been verified that the error involved in the mean fuel temperature calculation (1317 0 K as given by SIEMENS and 1028 0 K as calculated by Dr. Penndorf) does not change substancially the calculation results

  20. From fundamental mode to the PWR type reactors blow off: physical analysis and contribution to the qualification of calculation tools

    International Nuclear Information System (INIS)

    Maghnouj, A.

    1996-01-01

    The work reported in this thesis centres on the resolution of reactor physics problems posed by the use in pressurised water reactors of fuel assemblies containing mixed uranium-plutonium oxide fuel (MOX). The work is essentially dependent on the results of the EPICURE experimental programme carried out between 1988 and 1994 in the reactor EOLE at the Cadarache Research Centre of the CEA. Our contribution to the validation of the computer program APOLLO2 and of its nuclear data library CEA93 shows that this code system satisfactorily calculates the neutronic characteristics of PWR cores. The validation of the experiments has provided useful information concerning the modifications required to be made to the library CEA93, which is based on the basic library of evaluated nuclear data, JEF2. This approach should now be extended to a wider basis of reactor experimental data. The studies of methods for calculating coolant voiding coefficients has made it possible to select suitable methods based on the available deterministic methods of transport theory in 2 ad 3 dimensions. These schemes have given results in satisfactory agreement with the measurements made in EPICURE programme for both local and total coolant voiding. It would now be worth while to validate the chosen methods by comparisons with calculations made using continuous energy Monte Carlo methods. (author)

  1. Expert system for assisting the repair operations on the control racks of the control rods assembly in a 900 MW PWR type reactor

    International Nuclear Information System (INIS)

    Monnier, B.; Doutre, J.L.; Franco, A.

    1990-01-01

    The expert system presented was developed for assisting the repair operations on the control equipment of the control rod assembly in a PWR type reactor. The expert system allows the representation of expert knowledge and diagnostic reasoning. The objective of the expert system is to achieve the most precise diagnostic and localizing of the breakdown elements, by processing the data acquired during breakdown. The development steps, the structure and the applications of the expert system are summarized. The expert system operates in an IBM PC equipped with a AMAIA 8 Mo card. A time schedule of 18 months is predicted [fr

  2. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  3. Modular simulation of the dynamics of a 925 MWe PWR electronuclear type reactor and design of a multivariable control algorithm

    International Nuclear Information System (INIS)

    Mansouri, S.

    1985-06-01

    This work has been consecrated to the modular simulation of a PWR 925 MWe power plant's dynamic and to the design of a multivariable algorithm control: a mathematical model of a plant type was developed. The programs were written on a structured manner in order to maximize flexibility. A multivariable control algorithm based on pole placement with output feedback was elaborated together with its correspondent program. The simulation results for different normal transients were shown and the capabilities of the new method of multivariable control are illustrated through many examples

  4. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  5. Contribution to fuel depletion study in PWR type reactors, reactor core with three and four regions of enrichment

    International Nuclear Information System (INIS)

    Teixeira, M.C.C.

    1977-03-01

    The main methods for calculation of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Power Plant are presented and discussed. (author)

  6. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  7. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

    International Nuclear Information System (INIS)

    Oliveira, A.C.J.G. de.

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

  8. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  9. Radiation shield for PWR reactors

    International Nuclear Information System (INIS)

    Esenov, Amra; Pustovgar, Andrey

    2013-01-01

    One of the chief structures of a reactor pit is a 'dry' shield. Setting up a 'dry' shield includes the technologically complex process of thermal processing of serpentinite concrete. Modern advances in the area of materials technology permit avoiding this complex and demanding procedure, and this significantly decreases the duration, labor intensity, and cost of setting it up. (orig.)

  10. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  11. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  12. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    International Nuclear Information System (INIS)

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  13. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  14. Assessment of subcriticality during PWR-type reactor refueling; Evaluation de la sous-criticite lors des operations de chargement d'un reacteur nucleaire REP

    Energy Technology Data Exchange (ETDEWEB)

    Verdier, A

    2005-04-15

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  15. Response of pressurized water reactor (PWR) to network power generation demands

    International Nuclear Information System (INIS)

    Schreiner, L.A.

    1991-01-01

    The flexibility of the PWR type reactor in terms of response to the variations of the network power demands, is demonstrated. The factors that affect the transitory flexibility and some design prospects that allow the reactor fits the requirements of the network power demands, are also discussed. (M.J.A.)

  16. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  17. Occupational radiation exposures at PWR type reactors: international comparison of some global indicators between 1975 and 1988

    International Nuclear Information System (INIS)

    Benedittini, M.; Tabare, M.

    1989-12-01

    This report presents occupational radiation exposures at PWRs for 1988 in eight countries, i.e. Belgium, Federal Republic of Germany, Finland, France, Japan, Sweden, Switzerland, United-States. It updates the CEPN report no 145 which covered the 1975-1987 period. Only reactors in commercial operation since july 1974 are included in the analysis. As for the previous years, there is a general decrease of the mean annual collective dose per reactor: 2.42 man-Sv per reactor in 1988, while it was 2.56 man-Sv in 1987 and 2.83 man-Sv in 1986. The most important reduction concerns the US reactors (46% decrease since 1982) and those of the Federal Republic of Germany (49% decrease since 1981). This general trend is closely related to the dose reduction programs adopted in countries, combined with the positive impact of the past experience. However, despite this tendency, one may note a small increase of the collective occupational exposure for four out of the eight countries during the 1987-1988 period. The gap between lower and higher values is also continuously reducing: from respectively 0.67 man-Sv (Finland) and 3.66 man-Sv (US) in 1987 (factor of 5.5), to 0.88 man-Sv and 3.08 man-Sv in 1988 (factor of 3.5). Excluding Finland, the difference is reduced to a factor 2. These results show a clear tendency towards an uniformization of practices and operations. Annual collective dose results presented as a function of the number of operating years confirm previous results: for all units, a first period with a collective doses increase, followed by a relative stability or even a reduction, and then for some of them, a second increase. In 1988, the average annual collective dose per reactor for all PWRs is about 1.27 man-Sv after one operating year, 2.58 man-Sv after three years, and respectively 4.16, 3.86 and 3.96 man-Sv after seven, ten and fourteen years [fr

  18. Analysis of reactivity insertion accidents in PWR reactors

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1978-06-01

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt

  19. Economic targets for small PWR reactor designs

    International Nuclear Information System (INIS)

    Board, J.

    1991-01-01

    Small reactors are likely to be less economic than large reactors, but the lower financial exposure with small reactors may be attractive to utilities contemplating a restart to a nuclear programme. New nuclear plant can be economic, but success will depend more on how the plant are built, rather than what type or size is built. A target for new plant for operation early in the next century should be a generation cost of 3p to 3.5 p/kWh. This corresponds to an overnight capital cost of Pound 1000/kWh to Pound 1100/kWh. (author)

  20. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  1. Thermohydraulic behaviour of the hot channel in a PWR type reactor under loss-of-coolant accident conditions (LOCA)

    International Nuclear Information System (INIS)

    Costa, J.R.

    1978-12-01

    An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt

  2. Evaluation of uncertainties of key neutron parameters of PWR-type reactors with slab fuel, application to neutronic conformity

    International Nuclear Information System (INIS)

    Bernard, D.

    2001-12-01

    The aim of this thesis was to evaluate uncertainties of key neutron parameters of slab reactors. Uncertainties sources have many origins, technologic origin for parameters of fabrication and physical origin for nuclear data. First, each contribution of uncertainties is calculated and finally, a factor of uncertainties is associated to key slab parameter like reactivity, isotherm reactivity coefficient, control rod efficiency, power form factor before irradiation and life-time. This factors of uncertainties were computed by Generalized Perturbations Theory in case of step 0 and by directs calculations in case of irradiation problems. One of neutronic conformity applications was about fabrication and nuclear data targets precision adjustments. Statistic (uncertainties) and deterministic (deviations) approaches were studied. Then, neutronics key slab parameters uncertainties were reduced and so nuclear performances were optimized. (author)

  3. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    Energy Technology Data Exchange (ETDEWEB)

    Bakalov, Ivan [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Berlin (Germany); Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany)

    2018-02-15

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  4. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    International Nuclear Information System (INIS)

    Bakalov, Ivan; Sonnenkalb, Martin

    2018-01-01

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  5. Stress corrosion cracking of iron-nickel-chromium alloys in primary circuit environment of PWR-type reactors

    International Nuclear Information System (INIS)

    Boursier, Jean-Marie

    1993-01-01

    Stress corrosion cracking of Alloy 600 steam generator tubing is a great concern for pressurized water reactors. The mechanism that controls intergranular stress corrosion cracking of Alloy 600 in primary water (lithiated-borated water) has yet to be clearly identified. A study of stress corrosion cracking behaviour, which can identify the main parameters that control the cracking phenomenon, was so necessary to understand the stress corrosion cracking process. Constant extension rate tests, and constant load tests have evidenced that Alloy 600 stress corrosion cracking involves firstly an initiation period, then a slow propagation stage with crack less than 50 to 80 micrometers, and finally a rapid propagation stage leading to failure. The influence of mechanical parameters have shown the next points: - superficial strain hardening and cold work have a strong effect of stress corrosion cracking resistance (decrease of initiation time and increase of crack growth rate), - strain rate was the most suitable parameter for describing the different stage of propagation. The creep behaviour of alloy 600 has shown an increase of creep rate in primary water compared to air, which implies a local interaction plasticity/corrosion. An assessment of the durations of the initiation and the propagation stages was attempted for the whole uniaxial tensile tests, using the macroscopic strain rate: - the initiation time is less than 100 hours and seems to be an electrochemical process, - the durations of the propagation stage are strongly dependent on the strain rate. The behaviour in high primary water temperature of Alloys 690 and 800, which replace Alloy 600, was studied to appraise their margin, and validate their choice. Then the last chapter has to objective to evaluate the crack tip strain rate, in order to better describe the evolution of the different stages of cracking. (author) [fr

  6. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  7. Experience feedback examination in PWR type reactors operating for the 1997-1999 period; Examen du retour d'experience en exploitation des reacteurs a eau sous pression pour la periode 1997-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The present report is relative to the examination that the permanent group has made on the experience feedback in operation for PWR type reactors for the period 1997-1999 that was on eleven themes chosen by the Nuclear Safety and Radiation Protection Authority. It used analysis reports made by I.R.S.N. in support of four meetings of the permanent group devoted to this examination from April 2001 to June 2002. The different themes were operating uncertainties, machining to vibrations, analysis of incidents and gaseous releases, circuits, human factors, behaviour of electric batteries, risk of cold source loss. (N.C.)

  8. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model; Aplicacao da teoria de perturbacao para calculos de sensibilidade em nucleos de reatores PWR, usando um modelo de dois canais

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, A.C.J.G. de

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs.

  9. Microstructural characterization of stainless steel 17-4 PH used in the control element of PWR-Type reactors submitted to different heat treatments

    International Nuclear Information System (INIS)

    Ferreira, Douglas F.A.; Rezende, Renato P.; Turcarelli, Tiago

    2017-01-01

    The Control Element is a set of mechanical components of pressurized water cooled nuclear reactors (PWR), with the function of modifying the reactivity of the nucleus by insertion and withdrawal of neutron absorptive rod, in order to change the flow of neutrons (power) to the necessary and desired levels. The control element also has a safety function when there is a need to have negative reactivity available to shut down the reactor in normal operating or accident situations. In this situation, the control element descends instantly and inserts the rods with absorptive material into the fuel element thus shutting down the reactor. The control element consists of control rods, which carry the neutron absorption material and is supported by the spider, pin, spring and spring retainer assembly. The control element has some components that need to have high resistance to impacts when the safety function is activated, so the material of this component must have high mechanical strength and toughness. One of the materials in which can be specified for this application is martensitic stainless steel 17- 4PH (UNS 17400). This steel, when subjected to the aging heat treatment, has its mechanical properties altered due to the precipitation of dispersed intermetallic compounds in the matrix. In all heat treatments performed the predominant microstructure is lath martensite. The heat treatment of the 620 °C / 4 h presented lower hardness when compared to the other treatments and when increase time and temperature the material presents Nb precipitates that increase the hardness. (author)

  10. Microstructural characterization of stainless steel 17-4 PH used in the control element of PWR-Type reactors submitted to different heat treatments

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Douglas F.A.; Rezende, Renato P.; Turcarelli, Tiago, E-mail: ferreira@marinha.mil.br, E-mail: renato.rezende@marinha.mil.br, E-mail: tiago.turcarelli@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (DDNM/CTMSP), São Paulo, SP (Brazil). Diretoria de Desenvolvimento Nuclear da Marinha

    2017-07-01

    The Control Element is a set of mechanical components of pressurized water cooled nuclear reactors (PWR), with the function of modifying the reactivity of the nucleus by insertion and withdrawal of neutron absorptive rod, in order to change the flow of neutrons (power) to the necessary and desired levels. The control element also has a safety function when there is a need to have negative reactivity available to shut down the reactor in normal operating or accident situations. In this situation, the control element descends instantly and inserts the rods with absorptive material into the fuel element thus shutting down the reactor. The control element consists of control rods, which carry the neutron absorption material and is supported by the spider, pin, spring and spring retainer assembly. The control element has some components that need to have high resistance to impacts when the safety function is activated, so the material of this component must have high mechanical strength and toughness. One of the materials in which can be specified for this application is martensitic stainless steel 17- 4PH (UNS 17400). This steel, when subjected to the aging heat treatment, has its mechanical properties altered due to the precipitation of dispersed intermetallic compounds in the matrix. In all heat treatments performed the predominant microstructure is lath martensite. The heat treatment of the 620 °C / 4 h presented lower hardness when compared to the other treatments and when increase time and temperature the material presents Nb precipitates that increase the hardness. (author)

  11. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  12. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain

    International Nuclear Information System (INIS)

    Gadea, J. R.; Willke, A.; Regidor, J. J.; Tecnatom, S. A.

    2010-01-01

    The guide EPRI MRP-139, which provides the way forward for the inspection and evaluation of dissimilar butt welds, the primary system of PWR reactors, indicating the type of nondestructive testing to be done in these areas, based on discovered several cases of default in lnconel alloys 600 and 182 in American and European plants. The phenomenon of cracking.

  13. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  14. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  15. Contribution to the study of the evolution of nuclear fuel composition in PWR type reactors. Reactor cores in three and four regions of enrichment

    International Nuclear Information System (INIS)

    Teixeira, M.C.C.

    1976-01-01

    The main methods for calculations of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Plant are presented and discussed. (author) [pt

  16. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-02-01

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

  17. Contribution to fuel depletion study in PWR type reactors, reactor core with three and four regions of enrichment; Contribuicao ao estudo da evolucao da composicao do combustivel em reatores tipo PWR nucleos a tres e a quatro regioes de enriquecimento

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, M C.C.

    1977-03-01

    The main methods for calculation of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR`s version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Power Plant are presented and discussed. (author).

  18. Conversion rate for PWR reactors in thorium cycle

    International Nuclear Information System (INIS)

    Angelkorte, G.M.

    1980-01-01

    This work concerns to the determination of the conversion-rate for a PWR reactor with an enrichment of 7.47%, considering a cell, geometrically equal to Angra I, composed by Thorium and U-238 in a 1:1 relation. The study was performed considering neutrons of one and two groups of energy, according to the suggestion from other authors sup(1,2). It was also performed a study about the production and consumption of fissile material. (author)

  19. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  20. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    Barroso, D.E.G.; Nair, R.P.K.; Vellozo, S.O.

    1981-01-01

    A study about fuel management in PWR reactors, where not only the conventional uranium cycle is considered, but also the thorium cycle as an alternative is presented. The final results are presented in terms of U 3 O 8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.) [pt

  1. Parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

    2015-01-01

    Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

  2. Radiation streaming in power reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

    1979-02-01

    Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

  3. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  4. PWR reactor vessel in-service-inspection according to RSEM

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

  5. Susceptibility to stress corrosion in stainless steels type AISI 321 and 12X18H10T used in PWR type reactors (WWER)

    International Nuclear Information System (INIS)

    Matadamas C, N.

    1995-01-01

    Titanium stabilized stainless steels have been utilized in sovietic pressurized water reactors (VVER) for avoid the susceptibility to Intergranular Corrosion (IGC) present in other austenitic stainless steels. However the Intergranular Corrosion resistance of this kind of materials has been questioned because of Intergranular Stress Corrosion Cracking failures (IGSCC) have been reported. This paper study the electrochemical behavior of the AISI 321 stainless steel in a H 3 BO 3 Solution contaminated with chlorides and its susceptibility to Intergranular Corrosion.Electrochemical prediction diagrams of the stainless steels AISI 321 and 12X18H10T (sovietic) sensitized (600 Centigrade, 3 h.) were compared. Cylindrical and conical samples were used in Slow Strain Rate Tests (SSRT), to determine the susceptibility to Stress Corrosion Cracking (SCC) in AISI 321 and 12X18H10T stainless steels. The results obtained showed that the temperature of the solution is a very important factor to detect this susceptibility. Fractography studies on the fracture surfaces of the samples obtained in the SSRT at high temperature were realized. Corrosion velocities of both AISI 321 and 12X18H10T stainless steels were determined using conical samples in the CERT system at high temperature. E.D.A.X. analysis was employed in both AISI 321 and 12X18H10T stainless steels in order to explain the degree of sensitization. (Author)

  6. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  7. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  8. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  9. Fine numerical modelling of thermohydraulic phenomena in EDF PWR reactors

    International Nuclear Information System (INIS)

    Boulot, F.

    1993-01-01

    Over the last 20 years, EDF has developed a family of 2D and 3D industrial thermohydraulics software to solve problems encountered in existing PWR power plants and to design new reactors for the future. The equations used in the models are the averaged Navier-Stokes and energy equations. A brief description is given of the four main codes developed for single-phase and two-phase water-steam flows, some of which use finite differences or finite volumes methods, while others make use of finite elements methods. An example of application is given for each code. (author). 4 figs., 4 refs

  10. Reactors of different types in the world nuclear power

    International Nuclear Information System (INIS)

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

  11. Deboration in nuclear stations of the PWR type

    International Nuclear Information System (INIS)

    1978-01-01

    Reactivity control in nuclear power stations of the PWR type is realised with boric acid. A method to concentrate boric acid without an evaporator has been studied. A flow-sheet with reverse osmosis is proposed. (author)

  12. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  13. United Kingdom and USSR reactor types

    International Nuclear Information System (INIS)

    Lewins, Jeffery

    1988-01-01

    The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

  14. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2007-07-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  15. Design study of a PWR of 1.300 MWe of Angra-2 type operating in the thorium cycle

    International Nuclear Information System (INIS)

    Andrade, E.P.; Carneiro, F.A.N.; Schlosser, G.J.

    1984-01-01

    The utilization of the thorium-highly enriched uranium and thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. The PWR of 1300 MWe from KWU (Angra-2 type) is taken as the reference reactor for the study. Reactor core design calculations for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM 2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. (Author) [pt

  16. Vibrations measurement in fast and PWR reactor study

    International Nuclear Information System (INIS)

    Tigeot, Y.; Epstein, A.; Hareux, F.

    1975-01-01

    In the past severe damages have occured in several nuclear reactors, by structural vibrations induced by the primary cooling flow. To avoid this kind of troubles, the SEMT makes studies for two different types of reactors. For the light pressurized water reactors, some tests have been made on the SAFRAN test loop which is a three loop 1/8 scale internal model of a 900 MWe reactor. This study is actually undertaken jointly with Framatome. Elsewhere, measurements have been made on the Phenix fast breeder sodium reactor, and studies are planned for the Super Phenix reactor [fr

  17. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  18. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

  19. Application of directional solidification ingot (LSD) in forging of PWR reactor vessel heads

    International Nuclear Information System (INIS)

    Benhamou, C.; Poitrault, I.

    1985-09-01

    Creusot-Loire Industrie uses this type of ingot for manufacture of Framatome 1300 and 1450 MW 4-loop PWR reactor vessel heads. This type of ingot offers a number advantages: improved internal soundness; greater chemical, structural and mechanical homogeneity of the finished part; simplified forging process. After a brief description of the pouring and solidification processes, this paper presents an analysis of the results of examinations performed on the prototype forging, as well as review of results obtained during industrial fabrication of dished heads from LSD ingots. The advantages of the LSD ingot over conventional ingots are discussed in conclusion

  20. Seismic analysis of the reactor coolant system of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Borsoi, L.; Sollogoub, P.

    1986-01-01

    For safety considerations, seismic analyses are performed of the Reactor Coolant System (R.C.S.) of PWR Plants. After a brief description of the R.C.S. and R.C.S. operation, the paper presents the two types of analysis used to determine the effect of earthquake on the R.C.S.: modal spectral analysis and nonlinear time history analysis. The paper finally shows how seismic loadings are combined with other types of loadings and illustrates how the consideration of seismic loads affects R.C.S. design [fr

  1. Modelling and simulation the radioactive source-term of fission products in PWR type reactors; Modelagem e simulacao do termo-fonte radioativo de produtos de fissao em reatores nucleares do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Porfirio, Rogilson Nazare da Silva

    1996-07-01

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  2. Bituminization of simulated PWR type reactor wastes, boric evaporator bottons and ion exchange resins, carried out in CNEN/SP using commercial bitumen available in the Brazilian market

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.

    1986-01-01

    The first results of the study of bituminization of simulated PWR wastes, boric evaporator bottons and spent ion-exclange resins (OH - , H + ) and incinerated ash-wates are presented and discussed. The study consisted of characterization of the commercial bitumen, locally available and bitumen wastes products of varying whight compositions. The characterization was carried out using standard analysis methods of ABNT and ASTM, and included measurement of, penetration, softening point and flash point. In addition, the bitumen samples were analized for their resin and asphaltene contents. For leaching studies, wastes products of bitumen and resin loaded with 134 Cs were utilized. The method used was according to the ISO norms. The simulation of the industrial process was carried out using an extruder-evaporator typically used in the plastic industries offered by Industria de Maquinas Miotto Ltda., Sao Bernardo do Campo - SP. (Author) [pt

  3. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.

    1978-09-01

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  4. Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author) [pt

  5. A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

    The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

  6. A method to determine the dampening system of control rod drop mechanism for PWR reactors

    International Nuclear Information System (INIS)

    Trindade, C.E.; Mattos, J.R.L. de; Perrotta, J.A.

    1988-08-01

    A method to determine the Control Assembly damping drop system (dashpot/guide tube) was developed. It's presented a theoretical model, an experimental device and the procedures to determine this system, which is used in PWR reactors. (author) [pt

  7. System for stress corrosion conditions tests on PWR reactors

    International Nuclear Information System (INIS)

    Castro, Andre Cesar de Jesus

    2007-01-01

    The study of environmentally assisted cracking (EAC) involves the consideration and evaluation of the inherent compatibility between a material and the environment under conditions of either applied or residual stress. EAC is a critical problem because equipment, components and structure are subject to the influence of mechanical stress, water environment of different composition, temperature and different material history. Testing for resistance to EAC is one of the most effective ways to determine the interrelationships among this variables on the process of EAC. Up to now, several experimental techniques have been developed worldwide, which address different aspects of environmental caused damage. Constant loading of CT specimens test is a typical example of test, which is used for the estimation of parameters of stress corrosion cracking. To assess the initiation stages and kinetics of crack growth, the testing facility should allow active loading of specimens in the environment that is close to the actual operation conditions of assessed component. This paper presents a testing facility for stress corrosion cracking to be installed at CDTN, which was designed and developed at CDTN. The facility is used to carry out constant load tests under simulated PWR environment, where temperature, water pressure and chemistry are controlled, which are considered the most important factors in SCC. Also, the equipment operational conditions, its applications, and restrictions are presented. The system was developed to operate at temperature until 380 degree C and pressure until 180 bar. It consists in a autoclave stuck at a mechanical system, responsible of producing load , a water treatment station, and a data acquisition system. This testing facility allows the evaluation of cracking progress, especially at PWR reactor. (author) operational conditions. (author)

  8. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  9. Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

    International Nuclear Information System (INIS)

    Afanas'ev, A.

    1999-01-01

    The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

  10. Contribution to the qualification of Gd calculation in PWR reactors

    International Nuclear Information System (INIS)

    Chaucheprat, Patrick.

    1982-06-01

    This thesis presents the state of knowledge on gadolinium and the advantages of its use as burnable poison. A study on the behaviour of gadolinium makes it possible to bring out the essential parameters to which it is sensitive. The most important part of this work is devoted to the measurements by oscillations carried out in Minerve in 1981. The conceiving and implementation of this campaign are reported. The experimental results and the amending factors linked to the interpretation are presented. To complete this study at zero time, it seemed useful to process configurations with fuel clusters of UO 2 - Gd 2 O 3 in order to see the effect of UO 2 - Gd 2 O 3 rods in interaction. To this end, efficiency determinations of UO 2 - Gd 2 O 3 rod clusters were carried out in the Melodie lattice. The second part of this work involves the change in the gadolinium. Two main points are tackled here. The first concerns the determinations by oscillations of ''reconstituted'' samples that are composed of two concentric rings with various 235 U enrichments and gadolinium levels so as to simulate irradiated UO 2 - Gd 2 O 3 fuel. The second point is devoted to the description of the GEDEON experiment. UO 2 - Gd 2 O 3 rods will be irradiated in a 13 x 13 lattice of which the spectrum is representative of that of a PWR. This experiment will take place in the centre of the Melusine reactor at Grenoble [fr

  11. Development of thermohydraulic software for PWR reactors with natural circulation

    International Nuclear Information System (INIS)

    Chasseur, Alfredo F.; Rauschert, A.; Delmastro, Dario F.

    2009-01-01

    The basics concepts about the development of software for steady state analysis of a reactor with natural circulations, in the primary circuit, are exposed. The reactor type is pressurized light water. The equations, correlations and flux diagrams of the source code of the software developed are shown. The source code of the software was written in FORTRAN 77 making use of modular technique, this save development effort and release of news versions is simplified. (author)

  12. Research activity on thermohydraulic problems of PWR reactors

    International Nuclear Information System (INIS)

    Szabados, L.

    1976-06-01

    The general review of the experimental and theoretical research works on thermohydraulic investigation of pressurized water type reactors being done in the Central Research Institute for Physics is given. The main results of the theoretical and theoretical-numerical research are summarized. The most important result of the past years is the construction of the High Pressure Water Cooled Loop (NVH) thermohydraulic loop. Another significant achievement was the development of the reactor thermohydraulic program system. (Sz.N.Z.)

  13. The new electricity of France PWR: calculation scheme of neutron leakages from the reactor cavity

    International Nuclear Information System (INIS)

    Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.; Champion, G.

    1987-04-01

    A new calculation scheme is adapted to evaluate neutron fluxes in the reactor cavity and the containment of next french PWR. In this scheme a large part is given to Monte Carlo method, coupled with SN-method, in order to take into account multiple neutron diffusions and the complexity of the reactor geometry

  14. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  15. Steam generator for PWR type reactor

    International Nuclear Information System (INIS)

    Baba, Iwao; Hiyama, Nobuyuki.

    1994-01-01

    A steam generator of the present invention comprises a primary coolant chamber having primary coolants circulating therein, a secondary coolants chamber having secondary coolants and steams circulating therein, which are isolated from each other by a partition wall, and heat pipes disposed being passed through the partition wall. The heat pipes are disposed having an evaporation portion in the primary coolants chamber, a condensation portion in the secondary coolants chamber, and an intermediate heat insulating portion in the partition wall. Since the primary coolants containing radioactivity and the secondary coolants not containing radioactivity does not transfer heat directly by a heat transfer wall, a leakage accident of radioactivity to the secondary coolants can be prevented. Moreover, since the heat pipes are used, a great amount of heat can be transferred by a slight temperature difference by using steams of the heat transfer medium itself, latent heat due to coagulation, and capillary phenomenon. Since neither transferring power nor pumps are required, heat of the primary coolants can effectively be transferred to the secondary coolants. (N.H.)

  16. Abnormality diagnosis device for PWR type reactor

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The device of the present invention rapidly detects a small amount of primary coolants leaked from a heat transfer pipes in a steam generator to estimate the scale of the rupture and forecast the transition of the amount of the leakage. That is, a gamma-ray spectrum and dose rate measuring system having a bypass line capable of decaying short half-time nuclides disposed to each of second main steam pipelines and primary coolant pipelines. Data obtained from the measuring systems are compared, to judge the presence of abnormal symptom due to leakage and further, radioactivity concentration in main steams is calculated based on the dose rate at each of the measuring points. Further, radioactivity concentration of the leakage from the secondary main steam pipelines is calculated by change with passage of time. With such procedures, when primary coolants are leaked to the secondary main steam pipelines, the dose rate of the measured system is changed, thereby enabling to recognize the state at the abnormal point. The transition of abnormalities can be forecast with reference to the dose rate by change with passage of time. Further, state of the abnormal point can be recognized based on the gamma-ray spectrum, which is inherent upon occurrence of fuel rupture. (T.S.)

  17. Strategies of development of reactor types

    International Nuclear Information System (INIS)

    Bacher, P.

    2004-01-01

    The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

  18. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    Gueray, B.S.

    2001-01-01

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  19. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  20. Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

    International Nuclear Information System (INIS)

    Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

    1979-01-01

    The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

  1. Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

    International Nuclear Information System (INIS)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G.

    2007-01-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  2. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

  3. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  4. Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

    2000-03-01

    At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

  5. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the PWR for use in the naval reactor programme and the subsequent construction and operation of the nuclear power plant at Shippingport in 1957. The development of the BWR in 1954 and its selection for the plant at Dresden in 1959 established this concept as the other major reactor type in the US nuclear power programme. The subsequent growth profile is presented. A significant operating record has been accumulated concerning the availability of each of these reactor types. In addition, the use and performance of BWRs and PWRs in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to ensure effective safeguards at nuclear power installations; current measures are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. Both fuel cycles are examined in terms of: fuel burnup experience and prospects for improvement; natural uranium resources; enrichment capacity; reprocessing and recycle; and the interrelationships among the latter three factors. High-level waste management currently involving on-site storage of spent fuel is discussed in terms of available capacity and plans for expansion. The US electric utility industry viewpoint regarding an ultimate programme for waste management is outlined. Finally, the current economics and future cost trends of nuclear power plants are evaluated. (author)

  6. Application of the BEACON-TSM system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2011-01-01

    BEACON-TSM is an advanced system of the operation support of PWR reactors that combines the capabilities of an advanced nodal neutronic model and the measures of the instrumentation available in plant to determine, accurately and continuously, the distribution of power in the core and the available margins to the limits of the beak factors.

  7. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  8. Tank type reactor

    International Nuclear Information System (INIS)

    Otsuka, Fumio.

    1989-01-01

    The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

  9. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  10. Polynomial parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, Joao Claudio B.

    2015-01-01

    The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 U 92 enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K inf , generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  11. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  12. Tank type LMFBR type reactors

    International Nuclear Information System (INIS)

    Shimizu, Hiroshi

    1985-01-01

    Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

  13. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  14. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

  15. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  16. Study of PWR reactor efficiency as a function of turbine steam extractions

    International Nuclear Information System (INIS)

    Rocha, Janine Gandolpho da; Alvim, Antonio Carlos Marques; Martinez, Aquilino Senra

    2002-01-01

    The objective of this work is to optimize the extractions of the low-pressure turbine of a PWR nuclear reactor, in order to obtain the best thermodynamic cycle efficiency. We have analyzed typical data of a 1300 MW PWR reactor, operating at 25%, 50%, 75% and 100% capacities, respectively. The first stage of this study consists of generating a mathematical model capable of describing the reactor behavior and efficiency at any power level. The second stage of this study consists of to combine the generated mathematical model in an optimization computer program that optimize the extractions flow of the low-pressure turbine until it finds the optimal system efficiency. This work does not alter the nuclear facility project in any way. (author)

  17. Study for identification of control rod drops in PWR reactors at any burnup step

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2013-01-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  18. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  19. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  20. Control of PWR reactor energy supplied to a stream turbine

    International Nuclear Information System (INIS)

    Petetrot, J.F.; Parent, Pierre.

    1981-01-01

    This patent presents a process for regulating the power provided by a pressurized water nuclear reactor to a steam turbine, by moving the control rods absorbing the neutrons in the reactor core and by diverting a fraction of the steam produced by the reactor, outside the turbine circuit, by opening by-pass valves [fr

  1. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  2. Conversion ratio and consumption of fissile material in PWR reactors

    International Nuclear Information System (INIS)

    Tiba, C.

    1977-01-01

    It has been shown that the uranium resources will be insufficient for future projected demand. The many solutions to this problem are considered and, in particular, the effect of enrichment on the conversion ratio and hence total uranium comsumption is studied. The developed computacional method employs the one-group neutron diffusion theory. The model is verified by calculating typical burn-up, conversion ratio, U-235 comsumption and plutonium production values in PWR's, and comparing results with those in the published literature. The associated costs of U and U-Pu fuel cycles are also studied for various enrichment values [pt

  3. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

    Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

  4. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  5. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

  6. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  7. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  8. The AMEBA PWR, a new concept in the technology of nuclear reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A

    2000-05-01

    AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO{sub 2}. The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up.

  9. The AMEBA PWR, a new concept in the technology of nuclear reactor safety

    International Nuclear Information System (INIS)

    Novelli, A.

    2000-01-01

    AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO 2 . The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up

  10. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  11. The PWR fuel cycle. Utilization of uranium in a reactor

    International Nuclear Information System (INIS)

    Mignot, E.

    After having briefly described the core of a pressurized water reactor, the fuel is examined and, in particular, the change in reactivity that governs the renewal of the fuel. The present French nuclear units are taken as example and it is shown that with the development of the nuclear complex, it is no longer possible to reason on the basis of an isolated reactor, since the running of a reactor is set by the network and its working constraints become a priority. The optimization of the fuel control must therefore cover the total cost [fr

  12. Monitoring PWR reactor vessel liquid level with SPNDs during LOCAs

    International Nuclear Information System (INIS)

    Adams, J.P.

    1982-01-01

    Data from in-core self-powered neutron detectors taken during two nuclear loss-of-coolant accident simulations have been correlated with core moderator density changes. The detector current attenuation has been calculated during blowdown and reflood phases of the simulation. Based on these data, it is concluded that these detectors could be used to monitor reactor vessel liquid level during loss-of-coolant accidents in pressurized water reactors

  13. Non-linear punctual kinetics applied to PWR reactors simulation

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-11-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model for the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and CSMP using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (Author) [pt

  14. Study on Reactor Physics Characteristic of the PWR Core Using UO2

    International Nuclear Information System (INIS)

    Tukiran Surbakti

    2009-01-01

    Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

  15. PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Schwirian, R.; Robison, G.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

  16. Study of power peak migration due to insertion of control bars in a PWR reactor

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Costa, Danilo Leite; Borges, Diogo da Silva; Lava, Deise Diana; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown

  17. Improvements of nuclear fuel management in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1978-07-01

    The severe variations to which the different elements contributing to the determination of the fuel cycle cost are subjected have led to a reopening of the problem of ''optimization'' of nuclear fuel management. The increase in costs of uranium ore, isotope separation work units (swu), reprocessing, the political implications of proliferation associated with the employment of reprocessing operations have been at the origin of a reassessment of present-day management. It therefore appeared to be appropriate to study variants with respect to a reference mode represented by the management of the PWR 900 MWe systems, without burnable poison in the cycle at equilibrium (Case 3 of Table 1). In order to obtain a complete view of impacts of such modifications, computations were carried out as far as the appraisal of the cycle cost and with reprocessing. There has likewise been added to this the estimate of the gain anticipated from certain improvements in the neutron balance contributed at the level of the lattice

  18. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  19. Study on external reactor vessel cooling capacity for advanced large size PWR

    International Nuclear Information System (INIS)

    Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

    2014-01-01

    External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

  20. Critical heat flux correlation analysis for PWR reactors with low mass flow

    International Nuclear Information System (INIS)

    Carajilescov, Pedro

    1996-01-01

    The major limit in the thermalhydraulic design of water cooled reactors consists in the occurrence of critical heat flux, which is verified by correlation of large range of validity. In the present work, the major design correlations were analyzed, through comparisons with experimental data, for utilization in PWR with low mass flux in the core. The results show that the EPRI correlation, with modifications, gives conservative results, from the safety point of view, with lower data spreading, being the most indicated for the reactor thermal design. (author)

  1. The application of neural networks for optimization of the configuration of fuel assemblies in PWR reactors

    International Nuclear Information System (INIS)

    Sadighi, M.; Setayeshi, S.; Salehi, A.A.

    2002-01-01

    This paper presents a new method to solve the problem of finding the best configuration of fuel assemblies in a PWR (Pressurized Water Reactor) core. Finding an optimum solution requires a huge amount of calculations in classical methods. It has been shown that the application of continuous Hop field neural network accompanied by the Simulated Annealing method to this problem not only reduces the volume of the calculations, but also guarantees finding the best solution. In this study flattening of neutron flux inside the reactor core of Brusher NPP is considered as an objective function. The result shows the optimum core configuration which is in agreement with the pattern proposed by the designer

  2. Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.

    1987-01-01

    The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author) [pt

  3. Earthquake resistance of residual heat removed (RHR) pump for pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, K.; Honma, T.; Matsubayashi, H.; Inazuka, H.

    1980-01-01

    The present paper deals with the earthquake resistance of the residual heat removed (RHR) pump of single stage double volute type, which is one of the structurally simplest pumps used for pressurized water reactors (PWR). The results of the study can be summarized as follows: (1) Any trouble which can give effect on the functions of the pump at earthquake does not become a problem so long as each part of the pump is of aseismatically rigid structure. (2) Aseismatic tolerance test in the pump's operating condition has shown that the earthquake resistance of the pump at its location has a tolerance about five times the dynamic design acceleration. (3) The pump is provided with an impeller-casing wear ring at the pressure boundary between the suction side pressure and discharge side pressure. This wear ring acts as an underwater bearing when the pump is in operation, and improves the vibration characteristics, particularly damping ratio, of the pump shaft to a great extent to make the pump more aseismatic. (4) In the evaluation of the underwater bearing characteristics of the wear ring, the evaluation accuracy of the vibration characteristics of the pump shaft can be improved by taking into consideration the pressure loss in the wear ring part from the head of the single stage of the pump due to the rotation of the impeller. (author)

  4. Safety aspects in decontamination operations: Lessons learned during the decommissioning of a small PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Ponnet, M.; Emond, O.

    2002-01-01

    Decontamination operations are generally executed during the decommissioning of nuclear installations for different objectives: decontamination of loops or large pieces to reduce the dose rate inside a contaminated plant or decontamination to minimize the amount of radioactive waste. These decontamination operations raise safety issues such as radiological exposure, classical safety, environmental releases, production and management of secondary waste, management of primary resources, etc. This paper presents the return of experience from decontamination operations performed during the dismantling of the BR3 PWR reactor. The safety issues are discussed for 3 types of decontamination operations: full system decontamination of the primary loop with a chemical process to reduce the dose rate by a factor of 10; thorough decontamination with an aggressive chemical process of dismantled pieces to reach the unconditional clearance values; and thorough decontamination processes with physical processes of metals and of concrete to reach the unconditional clearance values. For the protection of the workers, we must consider the ALARA aspects and the classical safety issues. During the progress of our dismantling operations, the dose rate issue was becoming less important but the classical safety issues were becoming preponderant due to the use of very aggressive techniques. For the protection of the environment, we must take all the precautions to avoid any leakages from the plant and we must use processes which minimize the use of toxic products and which minimize the production of secondary wastes. We therefore promote the use of regenerative processes. (author)

  5. Lifetime assessment on PWR reactor vessel internals in Korea

    International Nuclear Information System (INIS)

    Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

  6. Study of PWR reactor efficiency as a function of turbine steam extractions; Estudo da otimizacao da eficiencia de reator PWR em funcao das extracoes de vapor da turbina

    Energy Technology Data Exchange (ETDEWEB)

    Rocha, Janine Gandolpho da; Alvim, Antonio Carlos Marques; Martinez, Aquilino Senra [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    The objective of this work is to optimize the extractions of the low-pressure turbine of a PWR nuclear reactor, in order to obtain the best thermodynamic cycle efficiency. We have analyzed typical data of a 1300 MW PWR reactor, operating at 25%, 50%, 75% and 100% capacities, respectively. The first stage of this study consists of generating a mathematical model capable of describing the reactor behavior and efficiency at any power level. The second stage of this study consists of to combine the generated mathematical model in an optimization computer program that optimize the extractions flow of the low-pressure turbine until it finds the optimal system efficiency. This work does not alter the nuclear facility project in any way. (author)

  7. Aging management of PWR reactor internals in U.S. plants

    International Nuclear Information System (INIS)

    Amberge, K.J.; Demma, A.

    2015-01-01

    This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

  8. Determination of welding parameters for execution of weld overlayer on PWR nuclear reactor nozzles

    International Nuclear Information System (INIS)

    Ribeiro, Gabriela M.; Lima, Luciana I.; Quinan, Marco A.; Schvartzman, Monica M.

    2009-01-01

    In the PWR reactors, nickel based dissimilar welds have been presented susceptibilities the stress corrosion (S C). For the mitigation the problem a deposition of weld layers on the external surface of the nozzle is an alternative, viewing to provoke the compression of the region subjected to S C. This paper presents a preliminary study on the determination of welding parameters to obtain these welding overlayers. Welding depositions were performed on a test piece welded with nickel 182 alloy, simulating the conditions of a nozzle used in a PWR nuclear power plant. The welding process was the GTAW (Gas Tungsten Arc Welding), and a nickel 52 alloy as addition material. The overlayers were performed on the base metals, carbon steel an stainless steel, changing the welding parameters and verifying the the time of each weld filet. After that, the samples were micro structurally characterized. The macro structures and the microstructures obtained through optical microscopy and Vickers microhardness are presented. The preliminary results make evident the good weld quality. However, a small weld parameters influence used in the base material microstructure (carbon steel and stainless steel). The obtained results in this study will be used as reference in the construction of a mock up which will simulate all the conditions of a pressurizer nozzle of PWR reactor

  9. FBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Tsugio.

    1986-01-01

    Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

  10. Loop type LMFBR reactor

    International Nuclear Information System (INIS)

    Ito, Hiroyuki

    1989-01-01

    In conventional FBR type reactors, primary coolants at high temperature uprise at a great flow rate and, due to the dynamic pressure thereof, the free surface is raised or sodium is partially jetted upwardly and then fallen again. Then, a wave killing plate comprising a buffer plate and a deflection plate is disposed to the liquid surface of coolants. Most of primary sodium uprising from the reactor core along the side of the upper mechanism during operation collide against the buffer plate of the wave killing plate to moderate the dynamic pressure and, further, disperse radially of the reactor vessel. On the other hand, primary sodium passing through flowing apertures collides against the deflection plate opposed to the flowing apertures to moderate the dynamic pressure, by which the force of raising the free surface is reduced. Thus, uprising and waving of the free surface can effectively be suppressed to reduce the incorporation of cover gases into the primary sodium, so that it is possible to prevent in injury of the recycling pump, abrupt increase of the reactor core reactivity and reduction of the heat efficiency of intermediate heat exchangers. (N.H.)

  11. Influence of FIMA burnup on actinides concentrations in PWR reactors

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

  12. The construction of a PWR power station reactor building liner

    International Nuclear Information System (INIS)

    Skirving, N.; Goulding, J.S.; Gibson, J.A.

    1991-01-01

    Cleveland Bridge and Engineering Co Ltd (CBE) are constructing the Reactor Building Liner Plate containment of the Sizewell 'B' Power Station for Nuclear Electric Ltd. This has entailed extensive offsite prefabrication of components and their subsequent erection at Sizewell. It has been necessary to engineer temporary supporting mechanisms to enable manufacture and erection to proceed, yet also to withstand wet concrete forces during the progressive construction. The Reactor Building Liner Plate is a safety related system and as such, in addition to strict compliance with the ASME code, the Quality Assurance (QA) requirements of BS 5882 are applicable. A dedicated Project Team was established by CBE to control and direct the work. Equally important as satisfying the rigorous Q.A. requirements has been the need to meet programme and budget. This paper details CBE execution of the Project. (author)

  13. Meeting the next generation PWR safety requirements: The EPR Reactor

    International Nuclear Information System (INIS)

    Salhi, Othman

    2008-01-01

    The development process pursued the harmonization of technical solutions and the integration of all the lessons learned from earlier nuclear plants built by both vendors. As far as safety more specifically is concerned, the basic choice for the EPR was to adopt an evolutionary approach based on experience feedback from the reactors built by Areva, which at the time already amounted to nearly 100. This philosophy makes today's Areva EPR the natural descendant of the most advanced French N4 and German Konvoi power reactors currently in operation. EPR design choices affecting safety were motivated by a continuous quest for higher levels of safety. A two-fold approach was followed: 1. improvement of the measures aimed at further reducing the already very low probability of core melt 2. incorporation of measures aimed at further limiting the consequences of a severe accident, in the knowledge that its probability of occurrence has been considerably reduced. Through its filiations with French N4 and German Konvoi power reactors, the EPR benefits from the uninterrupted, evolutionary innovation process that has supported the development of PWRs since their introduction into the market place. This is especially true for safety where the EPR brings a unique combination of both tried and tested and innovative features that further improve the prevention of severe accidents and their mitigation

  14. Requirements on PWR reactor design with respect to seismic effects

    International Nuclear Information System (INIS)

    Novak, J.; Pecinka, L.

    1981-01-01

    From the seismic point of view the individual parts of a nuclear power plant must be built such as to allow the shutdown of the reactor up to the safe shutdown earthquake level, the removal of after-heat and the prevention of uncontrolled release of radioactivity into the environment. To the level of operating basic earthquake the plant must be designed such as to allow the operation of the reactor for a period of 100 hours from the seismic event without exceeding the permissible annual dose to personnel and population. The possibility of a loss-of-coolant accident owing to a seismic event is reduced mainly by the integrated performance of the primary circuit, the high-strength structure, the insulation of the main components from the shift of the foundations and the use of floating structures. The pressure vessel of the WWER-1000 reactor is therefore pAaced in a shaft on a support ring and is locked by another support ring. (Z.M.)

  15. Water Chemistry Control in Reducing Corrosion and Radiation Exposure at PWR Reactor

    International Nuclear Information System (INIS)

    Febrianto

    2006-01-01

    Water chemistry control plays an important role in relation to plant availability, reliability and occupational radiation exposures. Radiation exposures of nuclear plant workers are determined by the radiation rate dose and by the amount maintenance and repair work time Water chemistry has always been, from beginning of operation of power Pressurized Water Reactor, an important factor in determining the integrity of reactor components, fuel cladding integrity and minimize out of core radiation exposures. For primary system, the parameters to control the quality of water chemistry have been subject to change in time. Reactor water coolant pH need to be optimally controlled and be operated in range pH 6.9 to 7.4. At pH lower than 6.9, cause increasing the radiation exposure level and increasing coolant water pH higher than 7.4 will decrease radiation exposure level but increasing risk to fuel cladding and steam generator tube. Since beginning 90 decade, PWR water coolant pH tend to be operated at pH 7.4. This paper will discuss concerning water chemistry development in reducing corrosion and radiation exposure dose in PWR reactor. (author)

  16. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  17. A method to calculate spatial xenon oscillations in PWR reactors

    International Nuclear Information System (INIS)

    Ronig, H.

    1976-01-01

    The new digital computer programme SEXI for the calculation of spatial Xe oscillations is described. A series expansion of the flux density and the particle densities following the geometrical eigenfunctions of a homogeneous block reactor is chosen as an approach to the solution of the system of differential equations describing this feedback process between neutron flux density and Xe particle density. To calculate the neutron flux density, the time-dependent form of the diffusion equation is used instead of the more common stationary form. Integration is carried out using formal time differential quotients of the Fourier coefficients. (orig./RW) [de

  18. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    Barroso, D.E.G.

    1980-01-01

    A comparitive study of fuel recycling in Pressurized Water Reactors was developed, considering not only the conventional uranium cycle, but also the use of thorium as an alternative. The use of thorium was done by varying its conoentration in the homogeneous mixture with uranium in the fuel from 30% up to 90%. The U-233 produced is incorporated within the isotopic composition of irradiated uranium. Various fractions of irradiated recycled fuel to be reprocessed and recycled was considered. Various alternatives of recycling were outlined and a final comparison in the tests done, is furnished in terms of U 3 O 8 and UTS requirements and approximated costs of fuel cycle stages involved. The recycled fuel is considered to be uniformly distributed in the fuel element rods introduced in the nucleus. The influence of the utilization of thorium was also considered for the development of an optimum fuel cycle, regarding the safeguards against nuclear proliferation when utilizing plutonium. A zero-dimensional cellular model was adopted to represent the reactor and the calculus of microscopic cross-sections for the homogenized cell was done by the computer code LEOPARD. A digital computer program was develped for neutronic and fuel depletion calculus and to simulate the refueling of various cycles. (Author) [pt

  19. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  20. FBR type reactor

    International Nuclear Information System (INIS)

    Jinbo, Masakazu; Kawakami, Hiroto; Nagaoka, Kazuhito.

    1996-01-01

    In a LMFBR type reactor, a liquid level control means is disposed for lowering a level of liquid metal present in an annular gap along with temperature elevation of the liquid metal after the level is once elevated upon start-up of the reactor. In addition, a liquid level measuring means is disposed for measuring the level of the liquid metal present in the annular gap so as to intermittently lower the liquid level. Thus, temperature gradient in the vertical direction of the container can be moderated compared with the case where the liquid level is not changed or the case where temperature is changed together with the elevation of the liquid level. As a result, the change of difference of thermal expansion is decreased to reduce stresses generated in the circumferential direction thereby preventing occurrence of a liquid level heat ratchet phenomenon. Even if the liquid level control means should stop during operation, the liquid level lowers and does not cause a sharp heat gradient as in the case where the liquid level is elevated, and since the temperature of the liquid level is lowered even after shut down of the reactor, generated stresses are not increased. Safety of an intermediate heat exchanger vessel is ensured and observation from a control chamber is enabled. (N.H.)

  1. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  2. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  3. Aging assessment of PWR [Pressurized Water Reactor] Auxiliary Feedwater Systems

    International Nuclear Information System (INIS)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab

  4. Sensitivity of risk parameters to human errors in reactor safety study for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.K.; Hall, R.E.; Swoboda, A.L.

    1981-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study (RSS) for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study. The code employed point estimate approach and ignored the smoothing technique applied in RSS. It computed the point estimates for the system unavailabilities from the median values of the component failure rates and proceeded in terms of point values to obtain the point estimates for the accident sequence probabilities, core melt probability, and release category probabilities. The sensitivity measure used was the ratio of the top event probability before and after the perturbation of the constituent events. Core melt probability per reactor year showed significant increase with the increase in the human error rates, but did not show similar decrease with the decrease in the human error rates due to the dominance of the hardware failures. When the Minimum Human Error Rate (M.H.E.R.) used is increased to 10 -3 , the base case human error rates start sensitivity to human errors. This effort now allows the evaluation of new error rate data along with proposed changes in the man machine interface

  5. Study on dynamic characteristics of reduced analytical model for PWR reactor internal structures

    International Nuclear Information System (INIS)

    Yoo, Bong; Lee, Jae Han; Kim, Jong Bum; Koo, Kyeong Hoe

    1993-01-01

    The objective of this study is to establish the procedure of the reduced analytical modeling technique for the PWR reactor internal(RI) structures and to carry out the sensitivity study of the dynamic characteristics of the structures by varying the structural parameters such as the stiffness, the mass and the damping. Modeling techniques for the PWR reactor internal structures and computer programs used for the dynamic analysis of the reactor internal structures are briefly investigated. Among the many components of RI structures, the dynamic characteristics for CSB was performed. The sensitivity analysis of the dynamic characteristics for the reduced analytical model considering the variations of the stiffnesses for the lower and upper flanges of the CSB and for the RV Snubber were performed to improve the dynamic characteristics of the RI structures against the external loadings given. In order to enhance the structural design margin of the RI components, the nonlinear time history analyses were attempted for the RI reduced models to compare the structural responses between the reference model and the modified one. (Author)

  6. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  7. Qualification according to PDI's techniques UT EPRI methodology Phased Array for the inspection of vessels of PWR reactor with advanced robotic equipment; Cualificacion segun metodologia PDI de EPRI de te cnicas UT Phased Array para la inspeccion de vasijas de reactor PWR con eq uipos roboticos avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Gonzalez, P.; Fernandez, F.

    2014-07-01

    The techniques and procedures qualified in the program EPRI PDI are directly applicable in plants whose reference code is ASME XI - specifically the Appendix VIII-, mainly USA and countries in which it is established American PWR technology. While countries with reactors in operation technology ABB (Sweden) or type VVER (Finland and Eastern countries) requires a qualification of specific technical type ENIQ, PDI qualification is a valuable reference since it allows to deal with such qualifications with guarantees. (Author)

  8. Loading pattern optimization of PWR reactors using Artificial Bee Colony

    International Nuclear Information System (INIS)

    Safarzadeh, O.; Zolfaghari, A.; Norouzi, A.; Minuchehr, H.

    2011-01-01

    Highlights: → ABC algorithm is comparable to the canonical GA algorithm and PSO. → The performance of ABC shows that the algorithm is quiet promising. → The final band width of search fitness values by ABC is narrow. → The ABC algorithm is relatively easy to implement. - Abstract: In this paper a core reloading technique using Artificial Bee Colony algorithm, ABC, is presented in the context of finding an optimal configuration of fuel assemblies. The proposed method can be used for in-core fuel management optimization problems in pressurized water reactors. To evaluate the proposed technique, the power flattening of a VVER-1000 core is considered as an objective function although other variables such as K eff , power peaking factor, burn up and cycle length can also be taken into account. The proposed optimization method is applied to a core design optimization problem previously solved with Genetic and Particle Swarm Intelligence Algorithm. The results, convergence rate and reliability of the new method are quite promising and show that the ABC algorithm performs very well and is comparable to the canonical Genetic Algorithm and Particle Swarm Intelligence, hence demonstrating its potential for other optimization applications in nuclear engineering field as, for instance, the cascade problems.

  9. Apparatus for localizing disturbances in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Sykora, D.

    1989-01-01

    The invention according to CS-PS 177386, entitled ''Apparatus for increasing the efficiency and passivity of the functioning of a bubbling-vacuum system for localizing disturbances in nuclear power plants with a pressurized water reactor'', concerns an important area of nuclear power engineering that is being developed in the RGW member countries. The invention solves the problems of increasing the reliability and intensification during the operation of the above very important system for guaranteeing the safety of the standard nuclear power plants of Soviet design. The essence of the invention consists in the installation of a simple passively operating supplementary apparatus. Consequently, the following can be observed in the system: first an improvement and simultaneous increase in the reliability of its function during the critical transition period, which follows the filling of the second space with air from the first space; secondly, elimination of the hitherto unavoidable initiating role of the active sprinkler-condensation device present; thirdly, a more effective performance and subjection of the elements to disintegration of the water flowing from the bubbling condenser into the first space; and fourthly, an enhanced utilization of the heat-conducting ability of the water reservoir of the bubbling condenser. Representatives of the supplementary apparatus are autonomous and local secondary systems of the sprinkler-sprayer without an insert, which spray the water under the effect of gravity. 1 fig

  10. Management routes for materials arising from the decommissioning of a PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Demeulemeester, Y.; Moers, S.; Ponnet, M.

    2001-01-01

    The management of wastes from decommissioning is described for the on-going dismantling of the BR3 PWR small reactor. The incentive is put on the radionuclides characterization, the description of the various waste streams, the conditioning techniques for low radioactive waste (LAW) to high radioactive waste (RAW), the alternative evacuation routes (recycling in the nuclear, free release by decontamination) and the minimization of secondary wastes during dismantling. Finally, some considerations are given on the overall dismantling cost and on the relative costs of the various evacuation routes. (author)

  11. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  12. The application of modern nodal methods to PWR reactor physics analysis

    International Nuclear Information System (INIS)

    Knight, M.P.

    1988-06-01

    The objective of this research is to develop efficient computational procedures for PWR reactor calculations, based on modern nodal methods. The analytic nodal method, which is characterised by the use of exact exponential expansions in transverse-integrated equations, is implemented within an existing finite-difference code. This shows considerable accuracy and efficiency on standard benchmark problems, very much in line with existing experience with nodal methods., Assembly powers can be calculated to within 2.0% with just one mesh per assembly. (author)

  13. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  14. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

    International Nuclear Information System (INIS)

    Monteiro, Iara Arraes

    1999-02-01

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  15. Evaluation on the habitability of a reactor control room for a 1300 MWe PWR following a LOCA

    International Nuclear Information System (INIS)

    Chang, Si Young; Ha, Chung Woo

    1988-01-01

    An evaluation on the habitability of a reactor control room for a French 1300 MWe P'4 type PWR following a LOCA has been performed through exposure dose assessment for a reactor operator. A computer code COREX calculating the time-integrated exposure dose has been developed to provide a reasonable basis in this evaluation. Using COREX the exposure dose reduction factors in the reactor control room, the time--integrated radioactivities released into the atmosphere and the time-integrated exposure dose up to 30 days following the LOCA can be also calculated. From the exposure dose assessment, the time-integrated exposure dose to whole body and thyroid of a reactor operator were 0.36 mSv(0.036 rem) and 480 mSv(48.0 rem), respectively after 30 days following the LOCA. The thyroid dose of 480 mSv was nearly 10 times greater than the dose equivalent limit of 50 mSv(5.0 rem) set by the ICRP. Regarding the habitability of a reactor control room, this exceeding thyroid exposure dose could be reduced to 1.2 mSv(0.12 rem), which is 400 times less than the original, by considering the practical 4 work-shifts a day, and by improving the iodine removal efficiency of the filtration system n the reactor control room through the reinforcement of charcoal bed filters for iodine removal. The radiological habitability of a reactor control room, therefore, could be assured by comparing with the dose equivalent limit of the ICRP

  16. Dissolution process for advanced-PWR-type fuels

    International Nuclear Information System (INIS)

    Black, D.E.; Decker, L.A.; Pearson, L.G.

    1979-01-01

    The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

  17. Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

    2014-01-01

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  18. Tank type nuclear reactors

    International Nuclear Information System (INIS)

    Naito, Kesahiro; Shimoyashiki, Shigehiro; Yokota, Norikatsu; Takahashi, Kazuo.

    1985-01-01

    Purpose: To improve the seismic proofness and the radiation shielding of LMFBR type reactors by providing the reactor with a structure reduced in the size and the weight, excellent in satisfactory heat insulating property and having radioactive material capturing performance. Constitution: Two sheets of ceramic plate members (for instance, mullite, steatite, beryllium ceramics or the like) which can be fabricated into plate-like shape and have high heat insulating property are overlapped with each other, between which magnetic heat-insulating material with magnetizing magnetic ceramics (for example, Lisub(0.5)Fesub(2.5)O 4 , Ni-Fe 2 O 4 , Fe-Fe 2 O 4 ) are sandwiched and the whole assembly is covered with metal coating material (for example, stainless steels). The inside of the coating material is evacuated or filled with an inert gas with low heat-conductivity (argon) at a pressure less than 1 kg/cm 2 abs, considering that the temperature goes higher and the inner pressure increases upon operation. In this way, the size of the laminated structure can be reduced to about 1/7 of the conventional case. The magnetic heat insulating materials can capture the magnetic impurities in sodium. (Kawakami, Y.)

  19. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    Fairclough, M.P.; Tymons, B.J.

    1985-06-01

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  20. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  1. Effects of aging in containment spray injection system of PWR reactor containment

    International Nuclear Information System (INIS)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L.

    2014-01-01

    This paper presents a contribution to the study of the components aging process in commercial plants of Pressurized Water Reactors (PWR). The analysis is done by applying the method of Fault trees, Monte Carlo Method and Fussell-Vesely Importance Measurement. The study on the aging of nuclear plants, is related to economic factors involved directly with the extent of their operational life, and also provides important data on issues of safety. The most recent case involving the process of extending the life of a PWR plant can be seen in Angra 1 Nuclear Power Plant by investing $ 27 million in the installation of a new reactor cover. The corrective action generated an extension of the useful life of Angra 1 estimated in twenty years, and a great savings compared to the cost of building a new plant and the decommissioning of the first, if it had reached the operation time out 40 years. The extension of the lifetime of a nuclear power plant must be accompanied by special attention from the most sensitive components of the systems to the aging process. After the application of the methodology (aging analysis of Containment Spray Injection System (CSIS)) proposed in this paper, it can be seen that increasing the probability of failure of each component, due to the aging process, generate an increased general unavailability of the system that contains these basic components. The final results obtained were as expected and can contribute to the maintenance policy, preventing premature aging in nuclear power systems

  2. ALIBABA, an assistance system for the detection of confinement leaks in a PWR reactor

    International Nuclear Information System (INIS)

    Bedier, P.O.; Libmann, M.

    1995-01-01

    The objective of the Crisis Technical Center (CTC) of the French Institute for Nuclear Protection and Safety (IPSN) is to estimates the consequences of a given nuclear accident on the populations and the environment. ALIBABA is a data processing tool available at the CTC and devoted to the detection of confinement leaks in 900 MWe PWR reactors using the activity values measured by the captors of the installation. The heart of this expert system is a structural and functional representation of the different components directly involved in the leak detection (isolating valves, ventilation systems, electric boards etc..). This tool can manage the availability of each component to make qualitative and quantitative balance-sheets. This paper presents the ALIBABA software, an industrial prototype realized with the SPIRAL knowledge base systems generator at the CEA Reactor Studies and Applied Mathematics Service (SERMA) and commercialized by CRIL-Ingenierie Society. It describes the techniques used for the modeling of PWR systems and for the visualization of the survey. The functionality of the man-machine interface is discussed and the means used for the validation of the software are summarized. (J.S.). 6 refs

  3. INETEC new system for inspection of PWR reactor pressure vessel head

    International Nuclear Information System (INIS)

    Nadinic, B.; Postruzin, Z.

    2004-01-01

    INETEC Institute for Nuclear Technology developed new equipment for inspection of PWR and VVER reactor pressure vessel head. The new advances in inspection technology are presented in this article, as the following: New advance manipulator for inspection of RPVH with high speed of inspection possibilities and total automated work; New sophisticated software for manipulator driving which includes 3D virtual presentation of manipulator movement and collision detection possibilities; New multi axis controller MAC-8; New end effector system for inspection of penetration tube and G weld; New eddy current and ultrasonic probes for inspection of G weld and penetration tube; New Eddy One Raster scan software for analysis of eddy current data with mant advanced features which allows easy and quick data analysis. Also the results of laboratory testing and laboratory qualification are presented on reactor pressure vessel head mock, as well as obtained speed of inspection and quality of collected data.(author)

  4. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  5. Technical report: fabrication of PWR type rodlet fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Uno, Hisao; Sasajima, Hideo

    1990-06-01

    With respect to the simulated reactivity initiated accident (RIA) experiments with pre-irradiated LWR type fuel rods at nuclear safety research reactor (NSRR), there were principally three technical difficulties which should be overcome: (1) Fabrication of the rodlet fuel; Fuel rods from the commercial power reactors had an active column length by 3.6m. To utilize this for NSRR pulse experiment, rodlet fuel having an active column length by 0.12m (reduced to one thirtieth) is requested to fabricate without changing the inside fuel conditions. (2) Development of in-core instrumentations: During pre-irradiation stages, a long-sized fuel rod had dimensional changes by waterside corrosion, bowing, creep down and so on. The fuel also had greater amount of radioactive fission products. This condition is significant to in-core instrumentations to be attached to the fuel rods. Well characterized data to be obtained from these, however, are quite necessary and important from research point of view. Remote handling techniques to attach the rod pressure sensor, the cladding extensometer, the fuel extensometer, and the cladding surface thermocouple to pre-irradiated fuel rods are, therefore, requested to develop. (3) Installation of PIE equipments for pulsed rodlet fuels: PIE on the pulsed rodlet fuels are necessary to better understanding the fuel performance detaily. Equipments which can easily detect the data related to PCMI type fuel failure are matter of concern. Since 1986, the technical difficulties have been tried to overcome by all staffs belonging to Reactivity Accident Laboratory, NSRR Operation Division, Department of Reactor Fuel Examination and Hot Laboratory. This report describes the technical achievements obtained through four years work. (author)

  6. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  7. Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

    International Nuclear Information System (INIS)

    El-Sahlamy, N.M.

    2017-01-01

    One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

  8. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  9. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  10. Utilization of nuclear energy for generating electric power in the FRG, with special regard to LWR-type reactors

    International Nuclear Information System (INIS)

    Vollradt, J.

    1977-01-01

    Comments on interdependencies in energy industry and energy generation as seen by energy supply utilities, stating that the generation of electric power in Germany can only be based on coal and nuclear energy in the long run, are followed by the most important, fundamental, nuclear-physical, technological and in part political interdependencies prevailing in the starting situation of 1955/58 when the construction of nuclear power plant reactors began. Then the development ranging to the 28000 MW nuclear power output to be expected in 1985 is outlined, totalling in 115000 MW electric power in the FRG. Finally, using the respectively latest order, the technical set up of each of the reactor types with 1300 MWe unit power offered by German manufacturers are described: BBC/BBR PWR-type reactor Neupotz, KWU-PWR-type reactor Hamm and KWU PWR-type reactor double unit B+C Gundremmingen. (orig.) [de

  11. Comparison of radioactive doses after the last protection layer insight the reactor structure for Russian VVER-1000 and German PWR-1300 reactors

    International Nuclear Information System (INIS)

    Rahimi, A.; Mansourshaiflu, N.; Alizadeh, M. R.

    2004-01-01

    In pressurized reactors (VVER and PWR), various protections layers are used for reducing the output core doses. At any protection layer, some amount of neutron and gamma doses is reduced. In this project the axial flux of neutron and gamma beams have been evaluated at various protection layers in the operation state the German PWR-1300 and Russian VVER-1000 reactors by the MCNP computer code. For the purpose of effective use of the MCNP code and assuring its correct performance about of fluxed beams common and series of scientific answers and bench marks should be considered and the results obtained by the MCNP code, be compared with this answers. Then by using appropriate method, for reducing the flux variants of neutron and gamma beams at various protection layers of German PWR-1300 and Russian VVER-1000 reactors of the operation state of both reactors have been accelerated. In this projects, bench marks are computations and numbers existing in PSAR's present at Bushehr nuclear power plant. At the end, by using the results obtained and the standard doses, the time which a person can have work activity at the reactor wall (after the last protection layer), was compared for the operation status of the German PWR-1300 and Russian VVER-1000 reactors

  12. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  13. Fusion reactors - types - problems

    International Nuclear Information System (INIS)

    Schmitter, K.H.

    1979-07-01

    A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

  14. Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

    International Nuclear Information System (INIS)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-01-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  15. Sodium fast reactor: an asset for a PWR UOX/MOX fleet - 5327

    International Nuclear Information System (INIS)

    Tiphine, M.; Coquelet-Pascal, C.; Girieud, R.; Eschbach, R.; Chabert, C.; Grosman, R.

    2015-01-01

    Due to its low fissile content, Pu from spent MOX fuels is sometimes regarded as not recyclable in LWR. Based on the existing French nuclear infrastructure (La Hague reprocessing plant and MELOX MOX manufacturing plant), AREVA and CEA have evaluated the conditions of Pu multi recycling in a 100% LWR fleet. As France is currently supporting a Fast Reactor prototype project, scenario studies have also been conducted to evaluate the contribution of a 600 MWe SFR in the LWR fleet. These scenario studies consider a nuclear fleet composed of 8 PWR 900 MWe, with or without the contribution of a SFR, and aim at evaluating the following points: -) the feasibility of Pu multi-recycling in PWR; -) the impact on the spent fuels storage; -) the reduction of the stored separated Pu; -) the impact on waste management and final disposal. The studies have been conducted with the COSI6 code, developed by CEA Nuclear Energy Direction since 1985, that simulates the evolution over time of a nuclear power plants fleet and of its associated fuel cycle facilities and provides material flux and isotopic compositions at each point of the scenario. To multi-recycle Pu into LWR MOX and to ensure flexibility, different reprocessing strategies were evaluated by adjusting the reprocessing order, the choice of used fuel assemblies according to their burn-up and the UOX/MOX proportions. The improvement of the Pu fissile quality and the kinetic of Pu multi-recycling in SFR depending on the initial Pu quality were also evaluated and led to a reintroduction of Pu in PWR MOX after a single irradiation in SFR, still in dilution with Pu from UOX to maintain a sufficient fissile quality. (authors)

  16. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

    Science.gov (United States)

    Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

    2017-09-01

    The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  17. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  18. Pool-type reactor

    International Nuclear Information System (INIS)

    Hopkins, S.R.

    1977-01-01

    This invention relates to a pool nuclear reactor fitted with a perfected system to raise the buckets into a vertical position at the bottom of a channel. This reactor has an inclined channel to guide a bucket containing a fuel assembly to introduce it into the reactor jacket or extract it therefrom and a damper at the bottom of the channel to stop the drop of the bucket. An upright vertically movable rod has a horizontally articulated arm with a hook. This can pivot to touch a radial lug on the bucket and pivot the bucket around its base in a vertical position, when the rod moves up [fr

  19. Investigating the cooling ability of reactor vessel head injection in the Maanshan PWR using CFD simulation

    International Nuclear Information System (INIS)

    Tseng Yungshin; Lin Chihhung; Wan Jongrong; Shih Chunkuan; Tsai, F. Peter

    2011-01-01

    In order to reduce the crack growth rate on the welding of penetration pipe, Pressurized Water Reactor (PWR) of Maanshan nuclear power plant (NPP) uses vessel head injection to cool vessel lid and control rod driving components. The injection flow from the cold leg is drained by the pressure difference between cold leg and upper internal components. In this study, 10 million meshes model with 4 sub-models have been developed to simulate the thermal-hydraulic behavior by commercial CFD program FLUENT. The results indicate that the injection nozzles can provide good cooling ability to reduce the maximum temperature for lid on the vessel head. The maximum temperature of vessel lid is about 293.81degC. Based on the simulated temperature, ASME CODE N-729-1 was further used to recount the effective degradation years (EDY) and reinspection years (RIY) factors. It demonstrates that the EDY and RIY factors are still less than 1.0. Therefore, the re-inspection period for Maanshan PWR would not be significantly affected by the miner temperature difference. (author)

  20. An evaluation of debris mobility within a PWR reactor coolant system during the recirculation mode

    International Nuclear Information System (INIS)

    Andreychek, T.S.

    1987-01-01

    To provide for the long-term cooling of the nuclear core of a Pressurized Water Rector (PWR) following a hypothetical Loss-of-Coolant Accidnet (LOCA), water is drawn from the containment sump and pumped into the reactor coolant system (RCS). It has been postulated that debris from the containment, such as dirt, sand, and paint from containment walls and in-containment equipment, could be carried into the containment sump due to the action of the RCS coolant that escapes from the breach in the piping and then flows to the sump. Once in the sump, this debris could be pumped into the Safety Injection System (SIS) and ultimately the RCS itself, causing the performance of the SIS to be degraded. Of particular interest is the potential for core blockage that may occur due to debris transport into the core region by the recirculating flow. This paper presents a method of evaluating the potential for debris from the sump to form core blockages under recirculating flow conditions following a hypothetical LOCA for a PWR

  1. R and D relative to the serious accidents in the PWR type reactors: assessment and perspectives; R and D relative aux accidents graves dans les reacteurs a eau pressurisee: bilan et perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Bentaib, A.; Bonneville, H.; Caroli, H.; Chaumont, B.; Clement, B.; Cranga, M.; Koundy, V.; Laurent, B.; Micaelli, J.C.; Meignen, R.; Pichereau, F.; Plassart, D.; Van-Dorsselaere, P. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Ducros, G.; Journeau, Ch.; Magallon, D. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Durin, M.; Studer, E. [CEA Saclay 91 - Gif sur Yvette (France); Seiler, J.M. [CEA Grenoble, 38 (France); Ranval, W. [Electricite de France (EDF), 75 - Paris (France)

    2006-07-01

    This document presents the current state of the research relative to the grave accidents realized in France and abroad. It aims at giving the most exhaustive possible and objective vision of this original field of research. He allows to contribute to the identification and to the hierarchical organization of the needs of R and D, this hierarchical organization in front of, naturally, to be completed by a strong lighting on needs in terms of safety analyses associated with the different risks and the physical phenomena, in particular with the support of probability evaluations of safety level 2, whose the level of sharpness must be sufficient not to hide, by construction, physical phenomena of which the limited knowledge leads to important uncertainties. Let us note that neither the safety analyses, nor the E.P.S. 2 are presented in this document. This report presents the physical phenomena which can arise during a grave accident, in the reactor vessel and in the reactor containment, their chain and the means allowing to ease the effects. The corresponding scenarios are presented to the chapter 2. The chapter 3 is dedicated to the progress of the accident in the reactor vessel; the degradation of the core in reactor vessel (3.1), the behavior of the corium in bottom of reactor vessel (3.2) the break of the reactor vessel (3.3) and the fusion in pressure (3.4) are thus handled there. The chapter 4 concerns the phenomena which can lead to a premature failure of the containment, namely the direct heating of gases of the containment (4.1), the hydrogen risk (4.2) and the vapor explosion (4.3). The phenomenon which can lead to a delayed failure from the containment, namely the interaction corium-concrete, is approached on the chapter 5. The chapter 6 is dedicated to the problems connected to the keeping back and to the corium cooling in reactor vessel and out of reactor vessel, namely the keeping back in reactor vessel by re-flooding of the primary circuit or by re

  2. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  3. FBR type reactor

    International Nuclear Information System (INIS)

    Yamaoka, Mitsuaki

    1988-01-01

    Purpose: To enable to increase the burning period by enabling to decrease the reduction of burning reactivity and unifying the irradiation amount of fast neutrons. Constitution: A cylindrical reactor core made of fissile material-enriched fuel is constituted so as to form a plurality of layer-like enriched regions in which the enrichment degree of the fissile material is increased from the center to the radial and axial directions. Then, the ratio between the average enrichment degree for all of the enrichment regions other than the region at the reactor core center with the lowest enrichment degree and the enrichment degree of the enriched region formed at the center of the reactor core is made greater by 5 % or 20 % than the ratio at the initial burning stage where the power distribution of the reactor core is most flattened. (Kawakami, Y.)

  4. BWR type reactors

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka

    1983-01-01

    Purpose: To decrease the control rod exchanging frequency by increasing the working life of control rods for ordinary operation with large neutron irradiation dose, to thereby decrease the exposure dose for operators performing exchanging work, as well as decrease the amount of radioactive wastes resulted upon exchange of the control rods. Constitution: Hafnium solid metal is employed as the neutron absorber of control rods for usual operation inserted into and withdrawn from fuel assemblies for the reactor power control over the entire cycle of the ordinary reactor operation and boron carbide powder is employed as the neutron absorber for emergency control rods to be inserted between the fuel assemblies only upon reactor scram or shutdown, whereby the working life of the control rods for ordinary reactor operation with greater neutron irradiation dose can be improved. Accordingly, the control rod exchanging frequency can be reduced to decrease the exposure dose to the operator for conducting the exchanging work. (Yoshihara, H.)

  5. Design study of a PWR of 1300 MWe of Angra-2 type operating in the thorium cycle

    International Nuclear Information System (INIS)

    Andrade, E.P.; Carneiro, F.A.N.; Schlosser, J.G.

    1984-01-01

    The utilization of the thorium-highly enriched uranium and of the thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. Reactor core design calculations were performed for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM-2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. No technical problem that precluded the utilization of any of the options analyzed was found. The savings in uranium ore introduced by the thorium cycle with fuel recycling ranges from 13% to 52% as compared with the usual uranium once-through cycle; the SWU savings goes from 13% to 22%. (Author) [pt

  6. LWR type reactor

    International Nuclear Information System (INIS)

    Kato, Kiyoshi.

    1993-01-01

    A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

  7. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  8. Theoretical-experimental modelling of the momentum equation for PWR reactor steam generators

    International Nuclear Information System (INIS)

    Rodrigues, L.A.H.

    1994-01-01

    A mathematical model in steady-state conditions of the momentum equation at the secondary side of a vertical U-tube steam generator with recirculation is presented. The U-tube test section was the 150 bar - Circuito Termoidraulico Experimental - CTE-150. This facility is a Experimental Thermal-hydraulic Circuit and operates at the same conditions (pressure and temperature) of a typical PWR reactor. A comparison between the Homogeneous and Separate Flow models was done. those models were verified and compared with experimental data for several operational conditions. The results show that the model fits very well the experimental data and seems to be appropriate to study water recirculation of a steam generator secondary side. (author)

  9. Manufacturing and properties of closure head forging integrated with flange for PWR reactor pressure vessel

    International Nuclear Information System (INIS)

    Tomoharu Sasaki; Iku Kurihara; Etsuo Murai; Yasuhiko Tanaka; Koumei Suzuki

    2003-01-01

    Closure head forging (SA508, Gr.3 Cl.1) integrated with flange for PWR reactor pressure vessel has been developed. This is intended to enhance structural integrity of closure head resulted in elimination of ISI, by eliminating weld joint between closure head and flange in the conventional design. Manufacturing procedures have been established so that homogeneity and isotropy of the material properties can be assured in the closure head forging integrated with flange. Acceptance tensile and impact test specimens are taken and tested regarding the closure head forging integrated with flange as very thick and complex forgings. This paper describes the manufacturing technologies and material properties of the closure head forging integrated with flange. (orig.)

  10. In-service inspection of sub-coating defects in PWR reactor vessel tubes

    International Nuclear Information System (INIS)

    Birac, A.; Frappier, J.C.; Saglio, Robert.

    1982-08-01

    Since the presence of cracks under the coating of the tubes of certain PWR reactor vessels were noted during manufacture, the need emerged to develop a nondestructive testing method to guarantee the detection of existing cracks and to determine their potential evolution. An ultrasonic testing method was developed for the purpose. In Part 1, the choice of ultrasonic transducers is justified from the theoretical and practical standpoints. In Part 2, the results obtained on test specimens containing artificial defects are presented in accordance with the different parameters involved. In Part 3, covering parts with a large number of real defects, the results of real defect/recorded signal correlations are given, with respect to both detection and dimensions. Examples of automatic data processing are analyzed [fr

  11. Computer-generated vibratory signatures for EDF PWR reactor vessel internals

    International Nuclear Information System (INIS)

    Trenty, A.; Lefevre, F.; Garreau, D.

    1992-07-01

    This paper presents a device for generation of characteristic signatures for normal or faulty vibrations on EDF PWR internal structures. The objective is to test the efficiency of methods for diagnosing faults in these structures. With this device, it is possible to build an entire PSD in several phases: choice of a general basic shape, localized addition of several kinds of background noise, generation of peaks of variable shapes, adjustment of local or global amplifications... It also offers the possibility of distorting real PSDs acquired from the reactor: shifting frequency or modifying peak shape, eliminating or adding existing shapes or shapes to be created, smoothing curves... One example is given of simulated loss of function in a hold-down spring on a computer-generated PSD of ex-core neutron noise. The device is now being used to test the potential of neural networks in recognizing faults on internal structures

  12. Research on 3D power distribution of PWR reactor core based on RBF neural network

    International Nuclear Information System (INIS)

    Xia Hong; Li Bin; Liu Jianxin

    2014-01-01

    Real-time monitor for 3D power distribution is critical to nuclear safety and high efficiency of NPP's operation as well as the control system optimization. A method was proposed to set up a real-time monitor system for 3D power distribution by using of ex-core neutron detecting system and RBF neural network for improving the instantaneity of the monitoring results and reducing the fitting error of the 3D power distribution. A series of experiments were operated on a 300 MW PWR simulation system. The results demonstrate that the new monitor system works very well under condition of certain burnup range during the fuel cycle and reconstructs the real-time 3D distribution of reactor core power. The accuracy of the model is improved effectively with the help of several methods. (authors)

  13. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor

    International Nuclear Information System (INIS)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes

    2011-01-01

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  14. BWR type reactors

    International Nuclear Information System (INIS)

    Hayashi, Katsuhisa; Watanabe, Shigeru.

    1983-01-01

    Purpose: To simplify the structure of control rod driving systems, as well as improve the safety and maintainability thereof. Constitution: Control-rod-guide tubes are disposed vertically above the reactor core and control-rod drives are disposed further thereabove, by which the control rods are moved upwardly and downwardly from above the reactor core through the guide tubes. Further, a partitioning cylinder is provided between the inner cirumferential wall at the upper portion of a pressure vessel and the control-rod-guide tubes and a gas-liquid separator is disposed to the space between the partitioning cylinder and the pressure vessel wall, to which steams generated in the reactor core are introduced. In such a structure of the reactor, since all of the control rods are inserted or extracted by the control rod drive system from above the reactor core, if the control rod drives or the likes should fail and accidentally drop the control rods, they exert in the direction of suppressing the nuclear reaction, whereby the safety can be improved. (Sekiya, K.)

  15. FBR type reactors

    International Nuclear Information System (INIS)

    Maemoto, Junko.

    1985-01-01

    Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

  16. Sub-critical crack growth and clad integrity in a PWR reactor pressure vessel

    International Nuclear Information System (INIS)

    Tice, D.R.; Foreman, A.J.E.; Sharples, J.K.

    1987-10-01

    The possibility of in-service growth of sub-critical defects in a PWR reactor pressure vessel to a critical size which could result in vessel failure was addressed in both the 1976 and 1982 reports of the Light Water Reactor Study Group (LWRSG), under the Chairmanship of Dr W Marshall (now Lord Marshall). An addendum to this report was published by UKAEA in April 1987. The section of the addendum dealing with subcritical crack growth and the related issue of integrity of the stainless steel cladding on the inner vessel surface is reproduced in this report. This section of the LWRSG addendum provides a review of the current status of fatigue crack growth and environmentally assisted cracking research for pressure vessel steels in light water reactor environments, as well as a review of developments in crack growth assessment methods. The review concludes that the alternative assessment procedures now being developed give a more realistic prediction of in service crack growth than the ASME Section XI Appendix A fatigue crack growth curves. (author)

  17. FBR type reactor

    International Nuclear Information System (INIS)

    Inoue, Kotaro; Kawashima, Katsuyuki; Zuketen, Atsushi.

    1982-01-01

    Purpose: To flatten the power distribution of a reactor core and shorten the breeding time. Constitution: The reactor core comprises a core region having fission products, an outer blanket region surrounding the outer side of the core region and having fertile material and an inner blanket region disposed within the core region and having fertile material. The axial thickness of the inner blanket region is made greater at the central portion and smaller at the peripheral portion of the inner blanket region, and the outermost peripheral end at the peripheral portion of the inner blanket region is opposed by way of the core region to the outer blanket region. In such an arrangement, the power decrease in the peripheral portion of the core region can be suppressed to thereby flatten the power distribution in the reactor core and shorten the breeding time. (Moriyama, K.)

  18. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  19. Natural convection type reactor

    International Nuclear Information System (INIS)

    Nakayama, Takafumi; Horiuchi, Tetsuo; Moriya, Kimiaki; Matsumoto, Masayoshi; Akita, Minoru.

    1988-01-01

    Purpose: To improve the reliability by decreasing the number of dynamic equipments and safely shutdown the reactor core upon occurrence of accidents. Constitution: A pressure relief valve and a pressurizing tank or gravitational water falling tank disposed to the main steam pipe of a reactor are installed in combination. Upon loss-of-coolant accident, the pressure relief valve is opened to reduce the pressure in the reactor pressure vessel to the operation pressure for each of the tanks, thereby enabling to inject water in the pressurizing tank at first and, thereafter, water in the gravitational water falling tank successively to the inside of the pressure vessel. By utilizing the natural force in this way, the reliability can be improved as compared with the case of pumped water injection. Further, by injecting an aqueous boric acid to a portion of a plurality of tanks, if the control rod insertion becomes impossible, aqueous boric acid can be injected. (Takahashi, M.)

  20. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR [pressurized-water-reactor] plants

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1988-01-01

    Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs

  1. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  2. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  3. Results of safety analysis on PWR type nuclear power plants with two and three loops

    International Nuclear Information System (INIS)

    1979-01-01

    The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

  4. Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

    International Nuclear Information System (INIS)

    Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

    2008-01-01

    As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

  5. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  6. FBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Azekura, Kazuo; Inoue, Kotaro.

    1981-01-01

    Purpose: To decrease power fluctuations due to burning of blanket fuel element clusters by partially replacing the fertile materials in the blanket fuel element clusters with fissile materials. Constitution: Fertile materials in the radial blanket fuel element clusters disposed to the outside or inside of the reactor core are partially replaced with fissile materials. Since the power density of the fissile materials is at the maximum in the initial burning stage and decreases as the burning proceeds, the power density of the materials which is smaller in the initial burning stage and becomes greater with the burning by the neutron-accumulated plutonium is offset. Accordingly, the power fluctuations in the blanket fuel element clusters due to the burning made smaller thereby enable to form a reactor core with less power fluctuations due to burning under the constant coolant flow rate depending on the power in the final burning stage where the blanket power is maximum. (Moriyama, K.)

  7. FBR type reactors

    International Nuclear Information System (INIS)

    Otsuka, Masaya; Yamakawa, Masanori; Goto, Tadashi; Ikeuchi, Toshiaki; Yamaki, Hideo.

    1986-01-01

    Purpose: To prevent thermal deformation and making the container compact by improving the cooling performance of main container walls. Constitution: A pipeway is extended from a high pressure plenum below the reactor core and connected to the lower side of the flow channel at the inside of a thermal shielding layer disposed to the inside of the main container wall. Low pressure sodium sent from the low temperature plenum into the high pressure plenum is introduced to the pipeway, caused to uprise in the inside flow channel, then turned for the direction, caused to descend in the outer side flow channel between the main container and the inside flow channel and then returned to the low temperature plenum. A heat insulating layer disposed with argon gas is installed to the inside of the flow channel to reduce the temperature change applied upon reactor scram. An annular linear induction pump capable of changing the voltage polarity is disposed at the midway of the pipeway and the polarity is switched such that the direction of flow of the liquid sodium is exerted as a braking force upon rated operation, whereas exerted as a pumping force upon reactor scram. (Sekiya, K.)

  8. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

    International Nuclear Information System (INIS)

    Barre, B.; Camarcat, N.

    1995-01-01

    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

  9. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  10. The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

    International Nuclear Information System (INIS)

    Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

    1977-01-01

    The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

  11. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    International Nuclear Information System (INIS)

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-01-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  12. The NCSU [North Carolina State Univ.] freon PWR [pressurized water reactor] loop

    International Nuclear Information System (INIS)

    Caves, J.R.; Doster, J.M.; Miller, G.D.; Wehring, B.W.; Turinsky, P.J.

    1989-01-01

    The nuclear engineering department at North Carolina State University has designed and constructed an operating scale model of a pressurized water reactor (PWR) nuclear steam supply system (NSSS). This facility will be used for education, training, and research. The loop uses electric heaters to simulate the reactor core and Freon as the primary and secondary coolant. Viewing ports at various locations in the loop allow the students to visualize flow regimes in normal and off-normal operating conditions. The objective of the design effort was to scale the thermal-hydraulic characteristics of a two-loop Westinghouse NSSS. Provisions have been made for the simulation of various abnormal occurrences. The model is instrumented in much the same manner as the actual NSSS. Current research projects using the loop include the development of adaptive expert systems to monitor the performance of the facility, diagnose mechanical faults, and to make recommendations to operators for mitigation of accidents. This involves having thermal-hydraulics and core-physics simulators running faster than real time on a mini-supercomputer, with operating parameters updated by communication with the data acquisition and control computer. Further opportunities for research will be investigated as they arise

  13. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

  14. Comparative study T-type and I-type layout of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Eko Rudi Iswanto and Siti Alimah

    2010-01-01

    Determining plant layout is one of the five major stages during the life time of a nuclear power plant. Some important factors that affect in the selecting of plant layout are availability of infrastructure, economic aspects, social aspects, public and environment safety, and also easy to do. Another factor to be considered is requirements as seismic design, which refers to the principles of good security workers, communities and the environment of radiological risks. There are many layout types of nuclear power plant, two of them are T-type layout and I-type layout. Each type of the plant layout has advantage and disadvantage, therefore this study is to understand them. Good layout is able to provide a high level of security against earthquakes. In term of earthquake design, I-type layout has a higher security level than T-type layout. Therefore, I-type layout can be a good choice for PWR nuclear power plants 1000 MWe that will be built in Indonesia. (author)

  15. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  16. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF 2 -ThF 4 -UF 4 ) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate [fr

  17. FBR type reactor

    International Nuclear Information System (INIS)

    Nagai, Fumio.

    1979-01-01

    Purpose: To unify the temperature distribution in a nuclear reactor vessel by the provision of a gas recycle path for pressurizing a cover gas to recycle the cover gas and thus stir the gas in a cover gas chamber. Constitution: A plurality of gas inlet tubes and gas discharge tubes are provided to the wall of a cover gas chamber above the liquid level of coolants in a nuclear reactor vessel and the cover gas is recycled through the tubes. The plurality of gas inlet tubes are each provided at their tops with nozzles opening circumferentially and communicated to the outlet of a compressor. While on the other hand, the plurality of gas discharge tubes are communicated to the inlet of a compressor. Upon operation of the compressor, the pressurized cover gas is jetted out from the nozzles, swirls along the inner circumferential surface of the vessel and interrupts and stirs the vertical thermal convection. The gas, after swirling one-half of the inner circumferential surface of the vessel, automatically flows out of the gas discharging tubes opening behind the nozzles and then flows into the inlet of the compressor. (Seki, T.)

  18. Seismic analysis of a PWR 900 reactor: study of reactor building with soil-structure interaction and evaluation of floor spectra

    International Nuclear Information System (INIS)

    Gantenbein, F.; Aguilar, J.

    1983-08-01

    The purpose of this paper is the evaluation of seismic response and floor spectra for a typical PWR 900 reactor building with respect to soil-structure interaction for soil stiffness). The typical PWR 900 reactor building consists of a concrete cylindrical external building and roof dome, a concrete internal structure (internals) on a common foundation mat as illustrated. The seismic response is obtained by SRSS method and floor spectra directly from ground spectrum and modal properties of the structure. Seismic responses and floor spectra computation is performed in the case of two different ground spectra: EDF spectrum (mean of oscillator spectra obtained from 8 californian records) normalized to 0.2 g, and DSN spectrum (typical of shallow seism) normalized to 0.3 g. The first section is devoted to internals' modelisation, the second one to the axisymmetric model of the reactor, the third one to the seismic response, the fourth one to floor spectra

  19. Separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    Hida, Kazuki.

    1993-01-01

    In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

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

    International Nuclear Information System (INIS)

    1987-11-01

    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

  1. WWER-440 type reactor core

    International Nuclear Information System (INIS)

    Mizov, J.; Svec, P.; Rajci, T.

    1987-01-01

    Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

  2. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  3. BWR type reactors

    International Nuclear Information System (INIS)

    Tsunoyama, Shigeaki; Tanabe, Akira.

    1979-01-01

    Purpose: To provide a main steam pressure shock absorber for reflecting the effect of the pressure propagation to coolants surface in the reactor core. Constitution: An annular shock absorber having near the water level through holes for water level measurement is provided to the gap between the skirt of a steam separator and a pressure vessel. Pressure waves are made the rapid closure of a main steam check valve. If arrived from the dome to the shock absorber, are mostly reflected to the side of the dome and give no substantial effects on the water surface. If the through holes are made small enough, the effects of pressure waves passing through the holes are negligible if they reach the water surface. (Kawakami, Y.)

  4. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part describes the MSBR core (data presented are from ORNL 4541). The principal characteristics of the core are presented in tables together with plane and elevation drawings, stress being put upon the reflector, and loading and unloading. Neutronic, and thermal and hydraulic characteristics (core and reflectors) are more detailed. The reasons why a graphite with a tight graphite layer has been chosen are briefly exposed. The physical properties of the standard graphite (irradiation behavior) have been determined for an isotropic graphite with fine granulometry; its dimensional variations largely ressemble that of Gilsonite. The mechanical stresses computed (Wigner effect) do not implicate in any way the graphite stack [fr

  5. LMFBR type reactors

    International Nuclear Information System (INIS)

    Sakurai, Akio; Matsushita, Kazuo.

    1985-01-01

    Purpose: To surely prevent the vibrational displacement of a rector core, and remove the sliding portions so as to avoid sticking or localized load concentration. Constitution: Cylindrical vibration-damping walls are protruded from the inner surface at the bottom of the main vessel. The vibration-damping walls constituted with outer and inner walls defining a coolant passage surround the reactor core support structures with small gaps between the inner surface of the inner wall and the outer circumferential surface of the core support structures. If the core support structures tend to displace horizontally due to earthquakes or the likes, the small gaps are varied and the coolants flow through the gap. Vibration-damping for the core support structures can be obtained by the flowing resistance to the coolants. (Yoshino, Y.)

  6. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  7. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors

    International Nuclear Information System (INIS)

    Blanc-Tranchant, P.

    1999-11-01

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B 4 C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  8. The pseudo-harmonics method: an application involving perturbations caused by control rod insertion in PWR reactors

    International Nuclear Information System (INIS)

    Claro, L.H.; Alvim, A.C.M.; Thome, Z.D.

    1988-08-01

    The objective of this work is to stydy the effect of intense perturbations, such as control rod insertion in the core of PWR reactors, through a perturbation approach consisting of a modified version of the pseudo-harmonics method. A typical one-dimensional PWR reactor model was used as a reference state, from which two perturbations were imposed, simulation gray and black control rod insertion. In the first case, eigenvalue convergence was achieved with the eighth order of approximation approximation and perturbed fluxes and eigenvalue estimates agreed very well with direct calculation results. The second case tested represents a very intense localized perturbation. Oscillation in keff were observed er of approximation increased and the method failed to converge. Results obtained indicate that the pseudo-harmonics method can be used to compute 2 group fluxes and fundamental eigenvalue of perturbated states resulting from gray control rod insertion in PWR reactors. The method is limited, however, by perturbation intensity, as other perturbation methods are. (author) [pt

  9. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  10. Horizontal loading test by whole model specimen simulating inner concrete structure of PWR type nuclear power plant

    International Nuclear Information System (INIS)

    Furuya, Noriyuki; Sekine, Masataka; Kimura, Kozo; Yamaguchi, Yoshihiro; Yamaguchi, Tsuneo; Takeda, Toshikazu

    1985-01-01

    The Nuclear Power Engineering Test Center has performed a horizontal loading test by a whole model specimen simulating the inner concrete structure of a PWR type nuclear power plant in order to investigate restoring force characteristics of reactor buildings. This report describes the results of examination of applicability to the test results of analysis methods based on elastic theory. The analysis results of elastic stiffness, concrete cracking load, rebar yielding loads and ultimate strength were compared with the test results. According to this examination, it is recognized that even these analysis methods based on elastic theory are comparatively effective for analysis of an inner concrete structure of fairly complex configuration, although there are limits of the scope of applicability. (author)

  11. Assessment of some interfacial shear correlations in a model of ECC bypass flow in PWR reactor downcomer

    International Nuclear Information System (INIS)

    Popov, N.K.; Rohatgi, U.S.

    1987-01-01

    The bypass/refill process in the PWR reactor downcomer, following a large rupture of a cold leg coolant supply pipe, is a complicated thermo-hydraulic two-phase flow phenomenon. Mathematical modeling of such phenomena is always accompanied with a difficult task of selection of suitable constitutive correlations. In a typically hydrodynamic phenomenon, like ECC refill process of the reactor lower plenum is considered, the phasic interfacial friction is the most influential constitutive correlation. Therefore, assessment of the well-known widely-used interfacial friction constitutive correlations in the model of ECC bypass/refill process, is the subject of this paper

  12. PWR [pressurized water reactor] optimal reload configuration with an intelligent workstation

    International Nuclear Information System (INIS)

    Greek, K.J.; Robinson, A.H.

    1990-01-01

    In a previous paper, the implementation of a pressurized water reactor (PWR) refueling expert system that combined object-oriented programming in Smalltalk and a FORTRAN power calculation to evaluate loading patterns was discussed. The expert system applies heuristics and constraints that lead the search toward an optimal configuration. Its rate of improvement depends on the expertise coded for a search and the loading pattern from where the search begins. Due to its complexity, however, the solution normally cannot be served by a rule-based expert system alone. A knowledge base may take years of development before final acceptance. Also, the human pattern-matching capability to view a two-dimensional power profile, recognize an imbalance, and select an appropriate response has not yet been surpassed by a rule-based system. The user should be given the ability to take control of the search if he believes the solution needs a new direction and should be able to configure a loading pattern and resume the search. This paper introduces the workstation features of Shuffle important to aid the user to manipulate the configuration and retain a record of the solution

  13. ASTEC-CATHARE2 benchmarks on French PWR 1300MWe reactors

    International Nuclear Information System (INIS)

    Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

    2009-01-01

    The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe reactors. This PSA-2 is heavily relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, an important series of benchmarks with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe accident scenarios. The present paper details 2 out of the 14 studied scenarios: a 12 inches cold leg Loss of Coolant Accident (LOCA) and a 2 tubes Steam Generator Tube Rupture (SGTR). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and the ASTEC results of the core degradation phase are presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results. (author)

  14. Aging mechanisms in the Westinghouse PWR [Pressurized Water Reactor] Control Rod Drive system

    International Nuclear Information System (INIS)

    Gunther, W.; Sullivan, K.

    1991-01-01

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs

  15. Study of chemical additives in the cementation of radioactive waste of PWR reactors

    International Nuclear Information System (INIS)

    Vieira, Vanessa Mota; Tello, Cledola Cassia Oliveira de

    2012-01-01

    In this research it has been studied the effects of chemical admixtures in the cementation process of radioactive wastes. These additives are used to improve the properties of waste cementation process, both of the paste and of the solidified product. However there are a large variety of these materials that are frequently changed or taken out of the market. Then it is essential to know the commercially available materials and their effects. The tests were carried out with a solution simulating the evaporator concentrate waste coming from PWR nuclear reactors. It was cemented using two formulations, A and B, incorporating higher or lower amount of waste, respectively. It was added chemical admixtures from two manufacturers (S and H), which were: accelerators, set retarders and superplasticizers. The experiments were organized by a factorial design 23. The measured parameters were: the viscosity, the setting time, the paste and product density and the compressive strength. The parameter evaluated in this study was the compressive strength at age of 28 days, is considered essential security issues relating to the handling, transport and storage of cemented waste product. The results showed that the addition of accelerators improved the compressive strength of the cemented products. (author)

  16. A multi-agent design for a pressurized water reactor (P.W.R.) control system

    International Nuclear Information System (INIS)

    Aimar-Lichtenberger, M.

    1999-01-01

    This PhD work is in keeping with the complex industrial process control. The starting point is the analysis of control principles in a Pressurized Water Reactor (P.W.R). In order to cope with the limits of the present control procedures, a new control organisation by objectives and means is defined. This functional organisation is based on the state approach and is characterized by the parallel management of control functions to ensure the continuous control of the installation essential variables. With regard to this complex system problematic, we search the most adapted computer modeling. We show that a multi-agent system approach brings an interesting answer to manage the distribution and parallelism of control decisions and tasks. We present a synthetic study of multi-agent systems and their application fields.The choice of a multi-agent approach proceeds with the design of an agent model. This model gains experiences from other applications. This model is implemented in a computer environment which combines the mechanisms of an object language with Prolog. We propose in this frame a multi-agent modeling of the control system where each function is represented by an agent. The agents are structured in a hierarchical organisation and deal with different abstraction levers of the problem. Following a prototype process, the validation is realized by an implementation and by a coupling to a reactor simulator. The essential contributions of an agent approach turn on the mastery of the system complexity, the openness, the robustness and the potentialities of human-machine cooperation. (author)

  17. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  18. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  19. Chemical process measurements in PWR-type nuclear power plants

    International Nuclear Information System (INIS)

    Glaeser, E.

    1978-01-01

    In order to achieve high levels of availability of nuclear power plants equipped with pressurized water reactors, strict standards have to be applied to the purity of coolant and of other media. Chemical process measurements can meet these requirements only if programmes are established giving maximum information with minimum expenditure and if these programmes are realized with effective analytical methods. Analysis programmes known from literature are proved for their usefulness, and hints are given for establishing rational programmes. Analytical techniques are compared with each other taking into consideration both methods which have already been introduced into nuclear power plant practice and methods not yet generally used in practice, such as atomic absorption spectrophotometry, gas chromatography, etc. Finally, based on the state of the art of chemical process measurements in nuclear power plants, the trends of future development are pointed out. (author)

  20. From fundamental mode to the PWR type reactors blow off: physical analysis and contribution to the qualification of calculation tools; Du mode fondamental a la vidange des reacteurs a eau sous pression: analyse physique et contribution a la qualification des outils de calcul

    Energy Technology Data Exchange (ETDEWEB)

    Maghnouj, A

    1996-01-18

    The work reported in this thesis centres on the resolution of reactor physics problems posed by the use in pressurised water reactors of fuel assemblies containing mixed uranium-plutonium oxide fuel (MOX). The work is essentially dependent on the results of the EPICURE experimental programme carried out between 1988 and 1994 in the reactor EOLE at the Cadarache Research Centre of the CEA. Our contribution to the validation of the computer program APOLLO2 and of its nuclear data library CEA93 shows that this code system satisfactorily calculates the neutronic characteristics of PWR cores. The validation of the experiments has provided useful information concerning the modifications required to be made to the library CEA93, which is based on the basic library of evaluated nuclear data, JEF2. This approach should now be extended to a wider basis of reactor experimental data. The studies of methods for calculating coolant voiding coefficients has made it possible to select suitable methods based on the available deterministic methods of transport theory in 2 ad 3 dimensions. These schemes have given results in satisfactory agreement with the measurements made in EPICURE programme for both local and total coolant voiding. It would now be worth while to validate the chosen methods by comparisons with calculations made using continuous energy Monte Carlo methods. (author)

  1. Reactor core design calculations and fuel management in PWR; Izracun projekta sredice in upravljanja z forivom tlacnovodnega reaktorja

    Energy Technology Data Exchange (ETDEWEB)

    Ravnik, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    Computer programs and methods developed at J. Stefan Institute for nuclear core design of Krsko NPP are treated. development, scope, verification and organisation of core design procedure are presented. The core design procedure is applicable to any NPP of PWR type. (author)

  2. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  3. Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

    International Nuclear Information System (INIS)

    Jong, Rudolf Peter de

    2004-01-01

    Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

  4. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  5. Aspects of PWR nuclear power plant secondary cycle relating to reactor safety

    International Nuclear Information System (INIS)

    Mueller, A.E.F.; Leal, M.R.L.V.; Dominguez, D.

    1981-01-01

    A safety study of the main steam system, condensate and feedwater systems and water treatment system that belong to the secondary cooling circuits of a PWR nuclear power plant is presented. (E.G.) [pt

  6. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  7. Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

    International Nuclear Information System (INIS)

    Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

    2017-01-01

    The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

  8. Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

    International Nuclear Information System (INIS)

    Soares, Alexandre de Souza

    2014-01-01

    The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

  9. Study on new-type fuel-related assembly handling tools for PWR NPP

    International Nuclear Information System (INIS)

    Fan Xiumei

    2013-01-01

    This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

  10. Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

    Energy Technology Data Exchange (ETDEWEB)

    Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

  11. Application of the perturbation theory-differential formalism-for sensitivity analysis in steam generators of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Sanders, R.M.G.; Andrade Lima, F.R. de; Alvim, A.C.M.

    1987-06-01

    An homogeneous model which simulates the stationary behavior of steam generators of PWR type reactors and uses the differential formalism of perturbation theory for analysing sensibility of linear and non-linear responses, is presented. The PERGEVAP computer code to calculate the temperature distribution in the steam generator and associated importance function, is developed. The code also evaluates effects of the thermohydraulic parameter variation on selected functionals. The obtained results are compared with results obtained by GEVAP computer code . (M.C.K.) [pt

  12. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  13. Use of 'tail' as spent fuel dilution factor of Angra-1 (PWR) for use in the Embalse (Candu) reactor

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Maiorino, Jose Rubens

    1995-01-01

    This work purposes a process to use the tail of isotopic enrichment as a factor of dilution (blending) for the burned fuel of Angra-I reactor (PWR) for final utilization in the Embalse (Candu). It was made use of the same technic in previous works that used natural uranium. For this purpose, it was made a tail parametrization inside of the traditional limits of enrichment (between 0.2 and 0.3%). The study showed that the tail utilization represents great savings for the uranium supplies and environment and economic advantages. (author). 8 refs, 4 figs, 11 tabs

  14. Probabilistic study of primary pump trip in a P.W.R. reactor: use of response surface methodology

    International Nuclear Information System (INIS)

    Bars, C.; Duchemin, B.; Maigret, N.; Peltier, J.; Rostan, O.; Villeneuve, M.J. de; Lanore, J.M.

    1981-09-01

    This paper describes a probabilistic study about the consequences of the trip or blockage of one of the three PWR reactor primary pumps. The distribution of the input parameters is taken into account and the resulting distribution of the consequence (number of failed fuel rods) is assessed. The necessity to do this study with the response surface methodology and the precautions to take are outlined. The results show that the probability to have failed fuel rods is about 10 -4 for pump trip and 0.16 for blockage with, in this case, a mean of 196 failed rods, that is 0.5 % of total number of rods

  15. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  16. Susceptibility to stress corrosion in stainless steels type AISI 321 and 12X18H10T used in PWR type reactors (WWER); Susceptibilidad a la corrosion bajo esfuerzo de barras de acero inoxidable AISI 321 y 12X18H10T en ambientes utilizados en reactores VVER

    Energy Technology Data Exchange (ETDEWEB)

    Matadamas C, N

    1996-12-31

    Titanium stabilized stainless steels have been utilized in sovietic pressurized water reactors (VVER) for avoid the susceptibility to Intergranular Corrosion (IGC) present in other austenitic stainless steels. However the Intergranular Corrosion resistance of this kind of materials has been questioned because of Intergranular Stress Corrosion Cracking failures (IGSCC) have been reported. This paper study the electrochemical behavior of the AISI 321 stainless steel in a H{sub 3}BO{sub 3} Solution contaminated with chlorides and its susceptibility to Intergranular Corrosion.Electrochemical prediction diagrams of the stainless steels AISI 321 and 12X18H10T (sovietic) sensitized (600 Centigrade, 3 h.) were compared. Cylindrical and conical samples were used in Slow Strain Rate Tests (SSRT), to determine the susceptibility to Stress Corrosion Cracking (SCC) in AISI 321 and 12X18H10T stainless steels. The results obtained showed that the temperature of the solution is a very important factor to detect this susceptibility. Fractography studies on the fracture surfaces of the samples obtained in the SSRT at high temperature were realized. Corrosion velocities of both AISI 321 and 12X18H10T stainless steels were determined using conical samples in the CERT system at high temperature. E.D.A.X. analysis was employed in both AISI 321 and 12X18H10T stainless steels in order to explain the degree of sensitization. (Author).

  17. MTR and PWR/PHWR in-pile loop safety in integration with the operation of multipurpose reactor - GAS

    International Nuclear Information System (INIS)

    Suharno; Aji, Bintoro; Sugiyanto; Rohman, Budi; Zarkasi, Amin S.; Giarno

    1998-01-01

    MTR and PWR/PHWR In-Pile Loop safety analysis in integration with the operation of Multipurpose Reactor - Gas has been carried out and completed. The assessment is emphasized on the function of the interface systems from the dependence of the operation and the evaluation to the possibility of leakage or failure of the in-pile part inside the reactor pool and reactor core. The analysis is refers to the logic function of the interface system and the possibility of leakage or failure of the in-pile part inside reactor pool and reactor core to consider the integrity of the core qualitatively. The results show that in normal and in transient conditions , the interface system meet the function requirement in safe integrated operation of in-pile loop and reactor. And the results of the possibility analysis of the leakage shows that the possibility based on mechanically assessment is very low and the impact to core integrity is nothing or can be eliminated. The possible position for leakage is on the flen on which one meter above the top level of the core, therefore no influence of leakage to the core

  18. OMR type process heat reactor

    International Nuclear Information System (INIS)

    Franzetti, Franco.

    1974-01-01

    A description is given of an OMR type reactor for heat generation. It includes a vessel the upper part of which is shut by a plug. The lower part of the vessel includes a core of fuel elements and is filled with an organic liquid. Over this there is a middle area filled with an inert gas. The plug includes an upper part forming a closure and resting around its edge on the vessel, and a lower part fixed under the closure and composed of a hollow cylindrical tank fitted with a bottom and filled with another organic liquid. The height of the cylindrical tank is such that, increased by the height of the first organic liquid in the lower area and above the core, it provides biological protection. The cooling system includes a heat exchanger and a pump to move the liquid from the lower part of the core and to inject some as spray into that part of the vessel filled with the inert gas. When loading and unloading, after the reactor is shut down, the clear organic liquid contained in the plug is discharged into the reactor vessel in such a way that it does not mix with the opaque organic liquid already contained in the vessel, and in that the opaque organic liquid is emptied out [fr

  19. Overall aspects of control of ISIS-type nuclear reactor

    International Nuclear Information System (INIS)

    Amato, S.; Santinelli, A.

    1996-01-01

    The paper describes the main aspects related to the definition of main controls required to operate an ISIS-type nuclear power reactors. ISIS is a PWR-type intrinsically safe nuclear reactor designed by ANSALDO, based on density lock concept; it presents, between the other safety functions, self-depressurization and core cooling capability for unlimited time. Due to its specific characteristics, the ISIS reactor required to development of new control philosophy (if compared with actual nuclear power reactor) with the implementation of new control functions, for instance the density locks hot/cold interface locations control. This paper describes the main control functions implemented, their rationale, as well as the dynamic simulation performed to verify the adequacy of controls definitions. The dynamic simulations here described refers to a step-wise power ramp of 100-90-100 (% of nominal power) and to a power ramp of 100-50-100 with a slope of 5%/min; the results obtained have shown the ISIS capability to perform such operational transients, despite its innovative design was mainly focused on intrinsically safe behaviour. (author)

  20. Chooz-B1, the new Electricite de France PWR: calculation scheme of neutron leakages from the reactor cavity

    International Nuclear Information System (INIS)

    Champion, G.; Thiriet, A.; Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.

    1987-04-01

    A new calculation scheme has been set up to assess the neutron field characteristics inside French PWR. In order to take into account multiple neutron scattering and the complexity of the reactor geometry, the use of Monte-Carlo methods have been heavily increased. They are coupled with classical SN.-methods. The main goal aimed at was to find out the neutron field characteristics at the level of the reactor pit openings. These radiation reference sources will be used to check the neutron shielding efficiencies. The new calculation scheme has been applied to CHOOZ-B1, the first unit of the new N4 program. The former results have been compared with the measurement results related to PALUEL-I and II PWR, two units of the previous P4 program. Although the core and the geometry are not entirely similar, it is possible to check with confidence the calculation results along the vessel and at the core midplane level with the measurement results at the same locations. It appears that they are in good agreement. Consequently, the new calculation scheme appears reliable

  1. Laboratory results gained from cold worked type 316Ti under simulated PWR primary environment

    International Nuclear Information System (INIS)

    Devrient, B.; Kilian, R.; Koenig, G.; Widera, M.; Wermelinger, T.

    2015-01-01

    Beginning in 2005, intergranular stress corrosion cracking (IGSCC) of barrel bolts made from cold worked type 316Ti (German Material No. 1.4571 K) was observed in several S/KWU type PWRs. This mechanism was so far less understood for PWR primary conditions. Therefore an extended joint research program was launched by AREVA GmbH and VGB e.V. to clarify the specific conditions which contributed to the observed findings on barrel bolts. In the frame of this research program beneath the evaluation of the operational experience also laboratory tests on the general cracking behavior of cold worked type 316Ti material, which followed the same production line as for barrel bolt manufacturing in the eighties, with different cold work levels covering up to 30 % were performed to determine whether there is a specific susceptibility of cold worked austenitic stainless steel specimens to suffer IGSCC under simulated PWR primary conditions. All these slow strain rate tests on tapered specimens and component specimens came to the results that first, much higher cold work levels than used for the existing barrel bolts are needed for IGSCC initiation. Secondly, additional high active plastic deformation is needed to generate and propagate intergranular cracking. And thirdly, all specimens finally showed ductile fracture at the applied strain rates. (authors)

  2. Announcement of recommendations of the Reaktor-Sicherheitskommission. As of 24 July 1997. Joint recommendations of RSK and GPR for safety requirements of future nuclear PWR-type power plants. English versions published in the years 1995 through 1997

    International Nuclear Information System (INIS)

    1997-01-01

    The recommendations, most parts given in English, refer to the European Pressurized Water Reactor (EPR) and have been established by the German RSK (reactor safety commission), the corresponding French organization GPR and the German SSK (radiation protection commission). This publication continues earlier joint recommendations by the national bodies, last published by the German BMU (responsible German ministry) on 5 May 1995, in BAnz. page 7452. The safety recommendations establish the basis for further activities in the Franco-German project for development of the EPR, a PWR type reactor of the next generation. (CB) [de

  3. Reduced scaling of thermal-hydraulic circuits for studies of PWR reactors natural circulation

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1993-01-01

    The Ishii et al. hydrodynamic similarity criteria for natural circulation were used for scaling reduced models of prototype passive residual heat removal system of a 600 M We PWR. The physical scales of the thermohydraulic parameters obtained presented a reasonable agreement when compared with simplified analytic models of the systems. (author)

  4. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR

    International Nuclear Information System (INIS)

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G.

    2008-01-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO 2 . (Author)

  5. High conversion burner type reactor

    International Nuclear Information System (INIS)

    Higuchi, Shin-ichi; Kawashima, Masatoshi

    1987-01-01

    Purpose: To simply and easily dismantle and reassemble densified fuel assemblies taken out of a high conversion ratio area thereby improve the neutron and fuel economy. Constitution: The burner portion for the purpose of fuel combustion is divided into a first burner region in adjacent with the high conversion ratio area at the center of the reactor core, and a second burner region formed to the outer circumference thereof and two types of fuels are charged therein. Densified fuel assemblies charged in the high conversion ratio area are separatably formed as fuel assemblies for use in the two types of burners. In this way, dense fuel assembly is separated into two types of fuel assemblies for use in burner of different number and arranging density of fuel elements which can be directly charged to the burner portion and facilitate the dismantling and reassembling of the fuel assemblies. Further, since the two types of fuel assemblies are charged in the burner portion, utilization factor for the neutron fuels can be improved. (Kamimura, M.)

  6. Estimation of damage by inmates of a PWR Reactor neutron irradiation. Project ZIRP; Estimacion del Dano por Irradiacion Neutronica en los Internos de un Reactor PWR. Proyecto ZIRP

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.

    2013-07-01

    The study presented here focuses on the analysis of neutron and gamma irradiation damage suffered by the inmates of the JC NPP reactor metallic materials throughout its operational life. Such analysis of radiation are part of a project of great international impact, led by EPRI (Electric Power Research Institute) from the MRP (Materials Reliability Program), which aims to relate the degradation of the properties of metallic materials of the inmates of the reactor, with the conditions of operation and irradiation to which have been subjected during the operational life of the plant.

  7. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  8. Livermore pool-type reactor

    International Nuclear Information System (INIS)

    Mann, L.G.

    1977-01-01

    The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

  9. Characterization and modeling of the thermal hydraulic and chemical environment of fuel claddings of PWR reactors during boiling

    International Nuclear Information System (INIS)

    March, Ph.

    1999-01-01

    In pressurised water reactors (PWR), nucleate boiling can strongly influence the oxidation rate of the fuel cladding. To improve our understanding of the effect of the boiling phenomenon on corrosion kinetics, information about the chemical and thermal hydraulic boundary conditions at the heating rod surface is needed. Moreover, very few data are available in the range of thermal hydraulic parameters of PWR cores (15,5 MPa and 340 deg C) concerning the two-phase flow pattern close to the fuel cladding. A visualization device has been adapted on an out-of-pile loop Reggae to obtain both qualitative and quantitative data. These observations provide a direct access to the geometrical properties of the vapor inclusions, the onset of nucleate boiling and the gas velocity and trajectory. An image processing method has been validated to measure both void fraction and interfacial area concentration in a bubbly two-phase flow. Thus, the visualization device proves to be a suitable and accurate instrumentation to characterize nucleate boiling in PWR conditions. The experimental results analysis indicates that a local approach is needed for the modelling of the fuel rod chemical environment. To simulate the chemical additives enrichment, a new model is proposed where the vapor bubbles are now considered as physical obstacles for the liquid access to the rod surface. The influence of the two-phase flow pattern appears to be of major importance for the enrichment phenomenon. This study clearly demonstrates the existence of strong interactions between the two-phase flow pattern, the rod surface condition, the corrosion process and the water chemistry. (author)

  10. Development of an engineering simulator for integral type PWR for nuclear ship

    International Nuclear Information System (INIS)

    Takahashi, Teruo; Shimazaki, Junya; Nakazawa, Toshio

    2000-01-01

    JAERI has developed a real-time engineering simulator for the integral type reactor MRX (Marine Reactor X) of power 100 MWt to evaluate the design and operational performance and to study highly automatic operations of a reactor plant. Marine reactor is operated under the conditions of pitching and rolling and load change, in comparison with a reactor for a land-based generating plant. And the MRX has systems with structural features, such as water-filled containment vessel, once-through type steam generator and emergency decay heat removal system. Considerations are paid to take these operational conditions and structural features into the simulation model. It is shown that the simulated results are consistent with the planned design and operational performance, and on the other hand present us some technical issues to be investigated in the design specifications. (author)

  11. Methodology to evaluate the crack growth rate by stress corrosion cracking in dissimilar metals weld in simulated environment of PWR nuclear reactor

    International Nuclear Information System (INIS)

    Paula, Raphael G.; Figueiredo, Celia A.; Rabelo, Emerson G.

    2013-01-01

    Inconel alloys weld metal is widely used to join dissimilar metals in nuclear reactors applications. It was recently observed failures of weld components in plants, which have triggered an international effort to determine reliable data on the stress corrosion cracking behavior of this material in reactor environment. The objective of this work is to develop a methodology to determine the crack growth rate caused by stress corrosion in Inconel alloy 182, using the specimen (Compact Tensile) in simulated PWR environment. (author)

  12. Hydroelastic model of PWR reactor internals SAFRAN 1 - Validation of a vibration calculation method

    International Nuclear Information System (INIS)

    Epstein, A.; Gibert, R.J.; Jeanpierre, F.; Livolant, M.

    1978-01-01

    The SAFRAN 1 test loop consists of an hydroelastic similitude of a 1/8 scale model of a 3 loop P.W.R. Vibrations of the main internals (thermal shield and core barrel) and pressure fluctuations in water thin sections between vessel and internals, and in inlet and outlet pipes, have been measured. The calculation method consists of: an evaluation of the main vibration and acoustic sources owing to the flow (unsteady jet impingement on the core barrel, turbulent flow in a water thin section). A calculation of the internal modal parameters taking into account the inertial effects of fluid (the computer codes AQUAMODE and TRISTANA have been used). A calculation of the acoustic response of the circuit (the computer code VIBRAPHONE has been used). The good agreement between the calculation and the experimental results allows using this method with better security for the prediction of the vibration levels of full scale P.W.R. internals

  13. Light water reactors development in Japan. (1) Introduction of LWR technology (PWR)

    International Nuclear Information System (INIS)

    Yamada, Ichita; Suzuki, Shigemitsu

    2008-01-01

    Evolutionary progress of the LWR plants in the last half-century was reviewed in series. Introduction of LWR technology (PWR) in Japan was reviewed in this article. Kansai Electric Power imported the Mihama-1 - a 340 MWe PWR built by Westinghouse Corp. It began operating in 1970 to supply power to the World Exposition (EXPO70). There followed a period in which designs was purchased from US vendors and they were constructed with the co-operation of Mitsubishi Heavy Industry, who would then receive a license to build similar plants in Japan and develop the capacity to design and construct PWRs by itself. Progress of designs, fabrications, project management and construction of PWRs were reviewed from technology transfer to its autonomy age. (T. Tanaka)

  14. Computer codes for the study of the loss of coolant accident of PWR reactors

    International Nuclear Information System (INIS)

    Gomolinski, M.; Menessier, D.; Tellier, N.

    1975-01-01

    The CEA has undertaken a large programme to study the consequence on the core of the LOCA of a PWR. In the programme, simultaneously carried out experiments and the development of the calculations means are described. Several experiments such as OMEGA, ERSEC and PHEBUS tests, which provide data to check the computer codes are outlined briefly in the paper. For analysis of the LOCA of a PWR, a series of computer codes, which are at present in use or under development, are linked with each other. The codes are DANAIDES for blowdown, CERES for refill and reflood, THETA-1B and FLIRA for heat up calculation during the blow-down and the reflooding period respectively. FLIRA-PASTEL, a combination of FLIRA and PASTEL which calculate the stress and deformations of material using the finite element method, will be used in place of FLIRA. The basic models and flowcharts of the above codes are described in the paper

  15. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  16. CRISTE - a subcomputer code for axial distribution, transient, of temperatures in a reactor channel of PWR

    International Nuclear Information System (INIS)

    Silva Neto, A.J. da; Roberty, N.C.; Carmo, E.G.D. do.

    1983-12-01

    The subroutine CRISTE was developed to calculate the temperature distribution for transients in a PWR coolant. The Crank-Nicholson approximation was used for the temporal discretization and a semi-analytical spatial solution was obtained. The temperature in the cladding was simulated by a routine adapted from the permanent distribution, and was used in on iterative method, following CRISTE subroutine. (E.G.) [pt

  17. Studies of a small PWR for onsite industrial power

    International Nuclear Information System (INIS)

    Klepper, O.H.; Smith, W.R.

    1977-01-01

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application

  18. Comparison of performance indicators of different types of reactors based on ISOE database

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2005-01-01

    The optimisation of the operation of a nuclear power plant (NPP) is a challenging issue due to the fact that besides general management issues, a risk associated to nuclear facilities should be included. In order to optimise the radiation protection programmes in around 440 reactors in operation with more than 500 000 monitored workers each year, the international exchange of performance indicators (PI) related to radiation protection issues seems to be essential. Those indicators are a function of a type of a reactor as well as the age and the quality of the management of the reactor. in general three main types of radiation protection PI could be recognised. These are: occupational exposure of workers, public exposure and management of PI related to radioactive waste. The occupational exposure could be efficiently studied using ISOC database. The dependence of occupational exposure on different types of reactors, e.g. PWR, BWR, are given, analysed and compared. (authors)

  19. Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

    International Nuclear Information System (INIS)

    Vallet, V.

    2012-01-01

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

  20. On reactor type comparisons for the next generation of reactors

    International Nuclear Information System (INIS)

    Alesso, H.P.; Majumdar, K.C.

    1991-01-01

    In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

  1. Fatigue crack growth characteristics of nitrogen-alloyed type 347 stainless under operating conditions of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Hong, Seok Min; Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Nuclear Materials Safety Research Division, Daejeon (Korea, Republic of); Kim, Seon Jin [Hanyang University, Division of materials science and engineering, Seoul (Korea, Republic of)

    2017-06-15

    The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

  2. A new formulation of the pseudocontinuous synthesis algorithm applied to the calculation of neutronic flux in PWR reactors

    International Nuclear Information System (INIS)

    Silva, C.F. da.

    1979-09-01

    A new formulation of the pseudocontinuous synthesis algorithm is applied to solve the static three dimensional two-group diffusion equations. The new method avoids ambiguities regarding interface conditions, which are inherent to the differential formulation, by resorting to the finite difference version of the differential equations involved. A considerable number of input/output options, possible core configurations and control rod positioning are implemented resulting in a very flexible as well as economical code to compute 3D fluxes, power density and reactivities of PWR reactors with partial inserted control rods. The performance of this new code is checked against the IAEA 3D Benchmark problem and results show that SINT3D yields comparable accuracy with much less computing time and memory required than in conventional 3D finite differerence codes. (Author) [pt

  3. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  4. Design of the control room of the N4-type PWR: main features and feedback operating experience

    International Nuclear Information System (INIS)

    Peyrouton, J.M.; Guillas, J.; Nougaret, Ch.

    2004-01-01

    This article presents the design, specificities and innovating features of the control room of the N4-type PWR. A brief description of control rooms of previous 900 MW and 1300 MW -type PWR allows us to assess the change. The design of the first control room dates back to 1972, at that time 2 considerations were taken into account: first the design has to be similar to that of control rooms for thermal plants because plant operators were satisfied with it and secondly the normal operating situation has to be privileged to the prejudice of accidental situations just as it was in a thermal plant. The turning point was the TMI accident that showed the weight of human factor in accidental situations in terms of pilot team, training, procedures and the ergonomics of the work station. The impact of TMI can be seen in the design of 1300 MW-type PWR. In the beginning of the eighties EDF decided to launch a study for a complete overhaul of the control room concept, the aim was to continue reducing the human factor risk and to provide a better quality of piloting the plant in any situation. The result is the control room of the N4-type PWR. Today the cumulated feedback experience of N4 control rooms represents more than 20 years over a wide range of situations from normal to incidental, a survey shows that the N4 design has fulfilled its aims. (A.C.)

  5. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  6. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data

  7. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    International Nuclear Information System (INIS)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia

    2017-01-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k ef f . The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi L ayer P erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  8. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2015-01-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  9. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  10. Proposal for a advanced PWR core with adequate characteristics for passive safety concept

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

    This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

  11. Improved identification to prevent transposition during operation of 900 MWe PWR reactors

    International Nuclear Information System (INIS)

    Leckner, J.M.; Dien, Y.; Cernes, A.

    1986-04-01

    Detailed human factors analysis of 900 MWe PWR control room identification systems was carried out by the Nuclear and Fossil Generation Division of Electricite de France (EDF) consequent to a series of incidents where personnel confused one plant unit, room or piece of equipment for another. Preliminary analysis uncovered coding inadequacies and suggested possible remedies. This data was used to prepare specifications for identification redesign at a pilot plant on which detailed investigations could be carried out. Recommended solutions were submitted to pilot plant operators and their opinion sollicited. Operator recommendations will be tried out on the pilot plant and adopted on a grid-wide basis if trials prove satisfactory

  12. Reactor types for the future

    International Nuclear Information System (INIS)

    Hall, A.C.

    1990-01-01

    The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

  13. Reactor types for the future

    Energy Technology Data Exchange (ETDEWEB)

    Hall, A C [PWR Power Projects Ltd., Knutsford, Cheshire (United Kingdom)

    1990-06-01

    The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

  14. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  15. Application of a PID controller based on fuzzy logic to reduce variations in the control parameters in PWR reactors

    International Nuclear Information System (INIS)

    Vasconcelos, Wagner Eustaquio de; Lira, Carlos Alberto Brayner de Oliveira; Brito, Thiago Souza Pereira de; Afonso, Antonio Claudio Marques; Cruz Filho, Antonio Jose da; Marques, Jose Antonio; Teixeira, Marcello Goulart

    2013-01-01

    Nuclear reactors are in nature nonlinear systems and their parameters vary with time as a function of power level. These characteristics must be considered if large power variations occur in power plant operational regimes, such as in load-following conditions. A PWR reactor has a component called pressurizer, whose function is to supply the necessary high pressure for its operation and to contain pressure variations in the primary cooling system. The use of control systems capable of reducing fast variations of the operation variables and to maintain the stability of this system is of fundamental importance. The best-known controllers used in industrial control processes are proportional-integral-derivative (PID) controllers due to their simple structure and robust performance in a wide range of operating conditions. However, designing a fuzzy controller is seen to be a much less difficult task. Once a Fuzzy Logic controller is designed for a particular set of parameters of the nonlinear element, it yields satisfactory performance for a range of these parameters. The objective of this work is to develop fuzzy proportional-integral-derivative (fuzzy-PID) control strategies to control the level of water in the reactor. In the study of the pressurizer, several computer codes are used to simulate its dynamic behavior. At the fuzzy-PID control strategy, the fuzzy logic controller is exploited to extend the finite sets of PID gains to the possible combinations of PID gains in stable region. Thus the fuzzy logic controller tunes the gain of PID controller to adapt the model with changes in the water level of reactor. The simulation results showed a favorable performance with the use to fuzzy-PID controllers. (author)

  16. Identification of mechanical vibrations in a PWR reactor using neutron noise signal analysis of the standard instrumentation; Identifikacija mehanichkih varijacija analizom signala shuma standardne neutronske instrumentacije PWR reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Kostic, Lj [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia); Runkel, J [Institut fuer Kerntechnik und Zerstoerungsfreie Pruefverfahren, Hannover (Germany)

    1988-07-01

    The neutron noise signals in a PWR power plant were analysed in terms of auto- and cross-power spectral densities, phases and coherences. Core barrel motion, fuel element vibrations and reactivity noise effect due to pressure variations have been monitored and analysed. (author)

  17. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  18. 'Kazmer' a complex noise diagnostic system for 1000 MWe PWR WWER type nuclear power units

    International Nuclear Information System (INIS)

    Por, G.

    1992-06-01

    Noise diagnostic systems have previously been developed and installed for the WWER-440 type reactors at the Paks Nuclear Power Plant, Hungary. Based on the experiences, the system has been extended and modified for use in 1000 MWe, WWER-1000 type units. KAZMER consists of three subsystem, the KARD reactor noise diagnostic system, ARGUS vibration monitoring system for rotation machinery, and ALMOS acoustic monitoring system. The installation of the KAZMER system at the Kalinin Nuclear Power Station, Russia, and the first operational experiences are outlined. (R.P.) 15 refs.; 9 figs

  19. Pressure loss coefficient evaluation based on CFD analysis for simple geometries and PWR reactor vessel without geometry simplification

    International Nuclear Information System (INIS)

    Ko II, B.; Park, J. P.; Jeong, J. H.

    2008-01-01

    Nuclear vendors and utilities perform lots of simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes and most of them were developed based on 1-dimensional lumped parameter models. These thermal-hydraulic system analysis codes require user input for pressure loss coefficient, k-factor; since they numerically solve Euler-equation. In spite of its high impact on the safety analysis results, there has not been good validation method for the selection of loss coefficient. During the past decade, however; computers, parallel computation methods, and 3-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. It is believed to be beneficial to take advantage of advanced commercial CFD codes in safety analysis and design of NPP5. The present work aims to validate pressure loss coefficient evaluation for simple geometries and k-factor calculation for PWR based on CFD. The performances of standard k-ε model, RNG k-ε model, Reynolds stress model (RSM) on the simulation of pressure drop for simple geometry such as, or sudden-expansion, and sudden-contraction are evaluated. The calculated value was compared with pressure loss coefficient in handbook of hydraulic resistance. Then the present work carried out analysis for flow distribution in downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of lower plenum is used. The results give a clear figure about flow fields in the reactor vessel, which is one of major safety concerns. The calculated pressure drop across downcomer and lower plenum appears to be in good agreement

  20. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  1. Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

    International Nuclear Information System (INIS)

    Huang Xueqing; Zhang Senru

    1999-01-01

    Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

  2. Refitting of the 'Celimene' hot cell for following up the fuel assembly of 900 MWe PWR power reactors

    International Nuclear Information System (INIS)

    Lhermenier, Andre; Van Craeynest, J.-C.

    1980-05-01

    The 'Celimene' cell adjoining the EL3 reactor provides for the acceptance, handling and the examination of irradiated fuel assemblies from power reactors (length approximately 4m, weight approximately 700 kg). Within the framework of the PWR fuel behavior follow-up or reprocessing, it is possible to extract an assembly representative of the normal fuel cycle, carry out non destructive tests on this assembly, extract pencils from it and re-insert this assembly, after refitting the head, into the normal fuel cycle for handling in a reprocessing plant or storage pond. Given suitable refitting techniques, the re-irradiation of the assembly can be considered after examination. Significant changes have been made to the buildings and the hoist facilities for handling very heavy flasks. It was necessary to rearrange the handling, machining and in-cell storage facilities. The development of an inspection rig will make it possible, some time in 1980, to carry out non destructive tests of assemblies, optical and metrological examination of assemblies prior to dismantling or of the structure after dismantling [fr

  3. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Programa de Pós-Graduação em Ciências e Técnicas Nucleares

    2017-07-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k{sub ef{sub f}}. The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi{sub L}ayer{sub P}erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  4. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    Abdelrazek, I.D.

    2008-01-01

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  5. Hot isostatically pressed (HIPed) thick-walled component for a pressurised water reactor (PWR) application

    International Nuclear Information System (INIS)

    Hookham, I.; Burdett, B.; Bridger, K.; Sulley, J.L.

    2009-01-01

    This paper presents the work conducted to justify and provide a quality assured HIPed thick-walled component for a PWR application; the component being designed and manufactured by Rolls-Royce. Rolls-Royce has previously published (ICAPP 08) its overall, staged approach to the introduction of powder HIPed components; starting with thin-walled, leak limited pressure boundaries, and culminating in the use of the powder HIPed process for thick walled components. This paper presents details specific to a thick walled pressure vessel component. Results are presented of non-destructive and destructive examinations of one of a batch of components. Mechanical testing and metallurgical examination results of sample material taken from different sections of the component are presented. A full range of test results is provided covering, as examples: tensile, Charpy impact and sensitization susceptibility. Differences in weldability between the HIPed and the previous forged form are also documented. (author)

  6. Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

    International Nuclear Information System (INIS)

    Pruitt, D.W.

    1992-01-01

    This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

  7. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    International Nuclear Information System (INIS)

    Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

    2004-01-01

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

  8. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V [ed.; Feinberg, O; Morozov, A [Russian Research Centre ` Kurchatov Institute` , Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  9. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    Ignatiev, V.; Devell, L.

    1995-01-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  10. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  11. Process and apparatus for adjusting a new upper reactor internals in a reactor vessel of a PWR

    International Nuclear Information System (INIS)

    Frizot, A.; Cadaureille, G.; Lalere, C.; Machuron, J.Y.

    1987-01-01

    On the new upper reactor internals is mounted devices for alignment and clearances, before introducing in the reactor vessel. After introducing alignment and clearances are measured. Adjustment pieces are provided for optimum clearances and alignment and fixed after removal from vessel. Decontamination is made by using water jets prior to fitting recess parts [fr

  12. Water-hammer experimental set-up and water-hammer experimental study for new types of check valve applied to PWR

    International Nuclear Information System (INIS)

    Liu Hanxun.

    1985-01-01

    This paper describes a self-designed constant temperature water-hammer shock test rig with stainless steel loop in which deionized water is used as working medium. To conduct water-hammer shock simulation tests for the countercurrent phenomenon occurred in the process of shutting, stopping, parallelling and switching the coolant loops of nuclear reactor, a specially designed four-way switching valve and its pneumatic mechanism are used. Water-hammer experimental study is performed for two types of PWR's nonshock check valve with diameter of 150 mm and 200 mm simultaneously. Transient performance of the shock waves, magnitude of their peaks and durations of their fluctuation, is obtained. Some analyses for existing calculational method on water-hammer are made

  13. PWR type overpower tests at 1620 GJ/KGU (18,800 MWD/MTU)

    International Nuclear Information System (INIS)

    Knudsen, P.; Bagger, C.; Carlsen, H.

    1979-01-01

    Three PWR type test fuel pins accumulated a burnup of 1620 GJ/kgU at heat loads decreasing from 45 to 28 kW/m (avg. test levels). One pin was ramped to 43 kW/m at 31 W/m/s; after 15 ks the power was increased to 45 kW/m and held constant for 1.9 Ms without failure indication. The other two pins were ramped to 44 kW/m at 23 W/m/s and then to 49 kW/m in a further 1.2 ks; both failed after max. 360 s. The post-irradiation examination revealed large stress-corrosion (SCC) type cladding cracks. Other cracks, down to a few μm deep, were probably early stages of large SCC cracks. Fission gas release in the intact pin was as high as 42% and estimated to be much lower for the two failed pins

  14. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

    International Nuclear Information System (INIS)

    Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

    2007-01-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  15. Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX

    International Nuclear Information System (INIS)

    Martinez, M.; Barrachina, T.; Miro, R.; Verdu Martin, G.; Chiva, S.

    2012-01-01

    In this work is proposed to evaluate the potential of the OpenFOAM code for the simulation of typical fluid flows in reactors PWR, in particular for the study of pressurized thermal stress. Test T1-1 has been simulated , within the OECD ROSA project, with the objective of evaluating the performance of the code OpenFOAM and models of turbulence that has implemented to capture the effect of the thrust forces in the case study.

  16. HTGR type reactors for the heat market

    International Nuclear Information System (INIS)

    Oesterwind, D.

    1981-01-01

    Information about the standard of development of the HTGR type reactor are followed by an assessment of its utilization on the heat market. The utilization of HTGR type reactors is considered suitable for the production of synthesis gas, district heat, and industrial process heat. A comparison with a pit coal power station shows the economy of the HTGR. Finally, some aspects of introducing new technologies into the market, i.e. small plants in particular are investigated. (UA) [de

  17. On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Pan Xiren; Zhang Chen.

    1988-01-01

    The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

  18. Fuzzy algorithm for an automatic reactor power control in a PWR

    International Nuclear Information System (INIS)

    Hah, Yung Joon; Song, In Ho; Yu, Sung Sik; Choi, Jung In; Lee, Byong Whi

    1994-01-01

    A fuzzy algorithm is presented for automatic reactor power control in a pressurized water reactor. Automatic power shape control is complicated by the use of control rods because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability for the load - follow operation including frequency control. In an attempt to achieve automatic power shape control without any design modification of the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multi - input multi - output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to the Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of the pressurized water reactor during the load - follow operation

  19. CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

    2002-10-01

    In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

  20. Study of passive residual heat removal system of a modular small PWR reactor

    International Nuclear Information System (INIS)

    Araujo, Nathália N.; Su, Jian

    2017-01-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

  1. Loop-type FBR reactor

    International Nuclear Information System (INIS)

    Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

    1996-01-01

    The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

  2. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

    Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

  3. Aging management of reactor internals and license renewal of US PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Tang, H. T. [Electric Power Research Institute, EPRI, 3420 Hillview Avenue, Palo Alto, California 94304 (United States)

    2006-09-15

    Age-related degradation mechanisms of key components are subject to aging management review by utilities considering plant license renewal. The management of aging effects in PWR internals must be demonstrated as specified in the US NRC Standard Review. The US NRC staff has also issued a Generic Aging Lessons Learned (GALL) report that documents the staff's basis for determining when existing generic programs are adequate to manage aging without change and when existing generic programs should be augmented for license renewal. The EPRI Materials Reliability Program (MRP) has been conducting studies to develop technical bases and guidelines to support aging management of PWR internals, with a particular attention to utility License Renewal commitments. The strategic approach taken by the MRP includes: developing an overall aging management framework, defining degradation mechanism screening values, categorizing and ranking internals components based on screening, performing functionality analyses and safety evaluation, and developing inspection and evaluation guidelines associated with each category of components. Screening criteria are developed for the following potential internals degradation mechanisms: - Stress Corrosion Cracking [Excluding Irradiation Effects]; - Irradiation-Assisted Stress Corrosion Cracking; - Thermal Aging Embrittlement; - Irradiation Embrittlement; - Void Swelling; - Stress Relaxation and Creep [Irradiation-enhanced]; - Wear; - Fatigue. The ranking and categorization calls to bin internals components into four categories: - Category A: component items for which aging degradation significance is minimal and aging effects are below the screening criteria; - Category C: 'lead' component items for which aging degradation significance is high or moderate and aging effects are above screening levels; - Category B: component items above screening levels but are not 'lead' component items and aging degradation significance

  4. Vibration characteristics of a PWR fuel rod supported by optimized H type spacer grids

    International Nuclear Information System (INIS)

    Choi, M. H.; Kang, H. S.; Yoon, K. H.; Kim, H. K.; Song, K. N.

    2002-01-01

    The spacer grids are one of the main structural components in the fuel assembly, which supports and protects the fuel rods from the external loads by seismic and coolant flow. In this study, a modal test and a FE vibration analysis using ABAQUS are performed on a PWR dummy fuel rod of 3.847 m which is continuously supported by eight Optimized H type spacer grids. The experimental results agree with previous works that the natural frequencies decrease, while the amplitudes increase, with the increase of the excitation force. The force levels showing the maximum displacement of 0.2 mm are in the range from 0.2 N to 0.3 N, and at the same force range the fundamental frequencies are measured around 42.0 Hz, at which the relatively big displacements are observed at the 7th span. The results from the modal tests and the FE analyses are compared by both Modal Assurance Criteria (MAC) values and mode shapes. The MAC values at 2nd, 4th, and 7th mode are below 50%. It is believed that the reason of the low MACs at those modes is that the vibration amplitudes of the modes are more distorted by the excitation force than those of the other modes

  5. Reliability analysis of 2 types of auxiliary feedwater system for PWR

    International Nuclear Information System (INIS)

    Ekariansyah, Andi Sofrany

    2002-01-01

    This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 - 2 compared with design A of 1,09 x 10 - 3 . The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

  6. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  7. Boron mixing transients in a 900 MW PWR vessel for a reactor start-up operation

    International Nuclear Information System (INIS)

    Alvarez, D.; Martin, A.; Schneider, J.P.

    1995-01-01

    In 1991 a R and D action, based on numerical simulations and experiments on PWRs'S primary coolant temperature or boron mixing capabilities, was initiated. This paper presents the test facility BORA-BORA (a 1/5th scaled mock-up of a 900 MW PWR vessel) and the Thermalhydraulic Finite Element Code N3S used for 3D calculations performed on the accurate geometry of the plant. As a validation test case of these experimental and numerical tools, we present the results obtained on the primary coolant mixing capabilities in the vessel with the three loops balanced in mass flow rate. The second part of this report deals with the mixing of a clear water plug in the vessel when a primary coolant pump start-up. The results are obtained in the mock-up in terms of boron concentration at the core inlet for several clear water plug volumes. The numerical results give the complete fluid flow and boron concentration patterns but comparisons were made at the core inlet. (author). 15 refs., 9 figs., 1 tab

  8. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  9. Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

    International Nuclear Information System (INIS)

    Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

    2010-01-01

    In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

  10. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  11. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

  12. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  13. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  14. Analysis of French (Paluel) pressurized water reactor design differences compared to current US PWR designs

    International Nuclear Information System (INIS)

    1986-05-01

    To understand better the regulatory approaches to reactor safety in foreign countries, the staff of the Nuclear Regulatory Commisssion has reviewed design information on the Paluel nuclear power plant, one of the current standard 1300-MWe plant operating in France. This report provides the staff's evaluation of major design differences between this standardized French plant and current US pressurized water reactor plants, as well as insights concerning French regulatory practices. The staff identified approximately 25 design differences, and an analysis of the safety significance of each of these design features is presented, along with an assessment comparing the relative safety benefit of each

  15. A study of core melting phenomena in reactor severe accident of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Gyoo Dong; Park, Shane; Kim, Jong Sun; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Jin Man [Korea Maritime Univ., Busan (Korea, Republic of)

    2001-03-15

    In the 4th year, SCDAP/RELAP5 best estimate input data obtained from the TMI-2 accident analysis were applied to the analysis of domestic nuclear power plant. Ulchin nuclear power plant unit 3, 4 were selected as reference plant and steam generator tube rupture, station blackout SCDAP/RELAP5 calculation were performed to verify the adequacy of the best estimate input parameters and the adequacy of related models. Also, System 80+ EVSE simulation was executed to study steam explosion phenomena in the reactor cavity and EVSE load test was performed on the simplified reactor cavity geometry using TRACER-II code.

  16. Development of an environment-insensitive PWR radial reflector model applicable to modern nodal reactor analysis method

    International Nuclear Information System (INIS)

    Mueller, E.M.

    1989-05-01

    This research is concerned with the development and analysis of methods for generating equivalent nodal diffusion parameters for the radial reflector of a PWR. The requirement that the equivalent reflector data be insensitive to changing core conditions is set as a principle objective. Hence, the environment dependence of the currently most reputable nodal reflector models, almost all of which are based on the nodal equivalence theory homgenization methods of Koebke and Smith, is investigated in detail. For this purpose, a special 1-D nodal equivalence theory reflector model, called the NGET model, is developed and used in 1-D and 2-D numerical experiments. The results demonstrate that these modern radial reflector models exhibit sufficient sensitivity to core conditions to warrant the development of alternative models. A new 1-D nodal reflector model, which is based on a novel combination of the nodal equivalence theory and the response matrix homogenization methods, is developed. Numerical results varify that this homogenized baffle/reflector model, which is called the NGET-RM model, is highly insensitive to changing core conditions. It is also shown that the NGET-RM model is not inferior to any of the existing 1-D nodal reflector models and that it has features which makes it an attractive alternative model for multi-dimensional reactor analysis. 61 refs., 40 figs., 36 tabs

  17. Criteria for safety-related nuclear-power-plant operator actions: 1982 pressurized-water-reactor (PWR) simulator exercises

    International Nuclear Information System (INIS)

    Crowe, D.S.; Beare, A.N.; Kozinsky, E.J.; Haas, P.M.

    1983-06-01

    The primary objective of the Safety-Related Operator Action (SROA) Program at Oak Ridge National Laboratory is to provide a data base to support development of criteria for safety-related actions by nuclear power plant operators. When compared to field data collected on similar events, a base of operator performance data developed from the simulator experiments can then be used to establish safety-related operator action design evaluation criteria, evaluate the effects of performance shaping factors, and support safety/risk assessment analyses. This report presents data obtained from refresher training exercises conducted in a pressurized water reactor (PWR) power plant control room simulator. The 14 exercises were performed by 24 teams of licensed operators from one utility, and operator performance was recorded by an automatic Performance Measurement System. Data tapes were analyzed to extract operator response times (RTs) and error rate information. Demographic and subjective data were collected by means of brief questionnaires and analyzed in an attempt to evaluate the effects of selected performance shaping factors on operator performance

  18. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system

    International Nuclear Information System (INIS)

    Hervouet, C.; Ranval, W.; Parozzi, F.; Eusebi, M.

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO 2 and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs

  19. Modelisation of soluble aerosols behaviour in the atmosphere of a PWR nuclear reactor in case of accident

    International Nuclear Information System (INIS)

    Abbas, A.F.

    1984-07-01

    After a short description of soluble aerosols accidental production in a PWR, a calculation model is given for physical properties of a gaz and steam mixture in a given atmosphere. Then the equilibrium of a saline drop with steam is studied. From the MASON equation, a calculation model is given for kinetic of volume variation of a saline drop and also a sensitivity study showing the little influence of the boundary layer on the drop surface, of the drop settling and of the thermodynamic conditions of the containment. As a numerical application, this condensation/evaporation model, and a simplified one with faster numerical resolution, is introduced in the AEROSOLS codes of the CEA-DEMT. The AEROSOLS/A2 suppose a log-normal distribution of the suspended particles in the containment. This application shows the very large sensitivity of the condensation depending on the moisture ratio inside the reactor building, and its primary importance on the behaviour of the aerosols. It is also shown that the simplified model gives a very little difference compared with the detailed model, and that the computation time is much more lower [fr

  20. Effect of the flexibility of the base mat on seismic response of a PWR-reactor building

    International Nuclear Information System (INIS)

    Waas, G.; Riggs, H.R.

    1983-01-01

    The flexibility of the base mat influences the stiffness and the radiation damping of foundations, In this paper its effect on the seismic response of an axisymmetric PWR-reactor building is investigated. The base mat of the building is stiffened by cylindrical concrete walls and by a rigid block in the center. Soft and stiff soil conditions are considered. The structure and its foundation are modelled by axisymmetric shell and volume elements with Fourier expansions in the circumferential direction. The soil is treated as a horizontally layered viscoelastic medium. Soil and structure are coupled along nodal rings. The stiffness matrix of the soil is computed using an explicit semi-analytic solution for displacements caused by ring loads acting on the surface or within a layered medium. The analysis is performed in the frequency domain, and the response in the time domain is computed by the fast Fourier transformation. The earthquake response is computed with and without including the flexibility of the relatively stiff base mat. The comparison shows that including the flexibility of the mat has hardly any effect on the resonant frequencies and the damping of the fundamental rocking and vertical modes. This is the case for soft and stiff soil conditions. However, the flexibility of the mat strongly affects the first structural deformation mode, in which the external and internal structures deflect in opposite directions. (orig./HP)

  1. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  2. Probabilistic analysis of fuel pin behaviour during an eventual loss of coolant in PWR reactors

    International Nuclear Information System (INIS)

    1981-02-01

    Brief description of the development of the coolant loss incident in a pressurized water reactor and analysis of its significance for the behaviour of the fuel rods. Description of a probalistic method for estimating the effects of the accident on the fuel rods and results obtained [fr

  3. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  4. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  5. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  6. Different types of power reactors and provenness

    International Nuclear Information System (INIS)

    Goodman, E.I.

    1977-01-01

    The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

  7. Some aspects of the utilization of zicaloy and austenitic steel as cladding material for PWR reactor fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Perrotta, J.A.

    1985-01-01

    The behaviour under irradiation of fuel rods for light water reactors was simulated by using fuel performance codes. Two types of cladding were analyzed: zircaloy and austenitic stainless steel. The fuel performance codes, originally made for zircaloy cladding, were adapted for austenitic stainless steel. The simulation results for the two types of cladding are presented, compared and discussed. (F.E.) [pt

  8. Development of a built-in type Control Rod Drive Mechanism (CRDM) for Advanced Marine Reactor X (MRX)

    International Nuclear Information System (INIS)

    Ishizaka, Y.; Iida, H.; Yamaji, A.

    1992-01-01

    For realization of the next generation Advanced Marine Reactor X(MRX) with higher safety, design studies and basic experiments have been done on the built-in type Control Rod Drive Mechanism (CRDM). The concept has been made clear of the CRDM that can be placed inside the reactor vessel and fits best to the MRX - an integrated-type PWR. In particular, the design has almost been completed for the driving motor and the latch magnet, which are the core of this CRDM. It is expected that the required performance can be assured even if there are losses due to the high temperature effect. (author)

  9. Analysis of ASTEC code adaptability to severe accident simulation for CANDU type reactors

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2008-01-01

    In order to prepare the adaptation of the ASTEC code to CANDU NPP severe accident analysis two kinds of activities were performed: - analyses of the ASTEC modules from the point of view of models and options, followed by CANDU exploratory calculation for the appropriate modules/models; - preparing the specifications for ASTEC adaptation for CANDU NPP. The paper is structured in three parts: - a comparison of PWR and CANDU concepts (from the point of view of severe accident phenomena); - exploratory calculations with some ASTEC modules- SOPHAEROS, CPA, IODE, CESAR, DIVA - for CANDU type reactors specific problems; - development needs analysis - algorithms, methods, modules. (authors)

  10. Vibration monitoring of the mechanical behavior of the internal structures of PWR reactors

    International Nuclear Information System (INIS)

    Assedo, R.; Carre, J.C.; Sol, J.C.

    1979-01-01

    The internal structures of pressurized water reactors are the seat of vibrations induced by fluctuations in primary fluid flow. A knowledge of these phenomena is indispensable in order to ensure that the structures are in proper mechanical order. It can also be used for operational monitoring. This paper describes all the methods developed and the results already achieved in this domain. The first part deals with tests on mockup associated with the calculation models which afforded a good knowledge of the vibrational characteristics of the internal structures, as well as the measurements made during hot tests of certain reactors which made it possible to qualify these models on real structures. The second part describes the means of detection (neutron noise, external accelerometers) as well as the processing methods used in the follow-up. A few typical results obtained on site are then presented. Finally, the general principles of operational monitoring of the mechanical behavior of the internal structures are described [fr

  11. Eddy current testing of PWR fuel pencils in the pool of the Osiris reactor

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1983-12-01

    A nondestructive testing bench is described. It is devoted to examination of high residual power fuel pencils without stress on the cladding nor interference with cooling. Guiding by fluid bearings decrease the background noise. Scanning speed is limited only by safety criteria and data acquisition configuration. Simultaneous control of various parameters is possible. Associated to an irradiation loop, loaded and unloaded in a reactor swinning pool, this bench can follow fuel pencil degradation after each irradiation cycle [fr

  12. Emotional learning based intelligent controller for a PWR nuclear reactor core during load following operation

    International Nuclear Information System (INIS)

    Khorramabadi, Sima Seidi; Boroushaki, Mehrdad; Lucas, Caro

    2008-01-01

    The design and evaluation of a novel approach to reactor core power control based on emotional learning is described. The controller includes a neuro-fuzzy system with power error and its derivative as inputs. A fuzzy critic evaluates the present situation, and provides the emotional signal (stress). The controller modifies its characteristics so that the critic's stress is reduced. Simulation results show that the controller has good convergence and performance robustness characteristics over a wide range of operational parameters

  13. Radioactive waste management at WWER type reactors

    International Nuclear Information System (INIS)

    1993-05-01

    This report was prepared within the framework of the Technical Assistance Regional Project on Advice on Waste Management at WWER Type Reactors, which was initiated by the IAEA in 1991. The Regional Project is an integral part of the IAEA's activities directed towards improvement of the safety and reliability of nuclear power plants with WWER type reactors (Soviet designed PWRs). Forty-five WWER type units are currently in operation and twenty-five are under construction in Bulgaria, Czechoslovakia, Finland, Hungary and the former USSR. The idea of regional collaboration between eastern European countries under the auspices of the IAEA was discussed for the first time during the last meeting of the Council for Mutual Economic Assistance (CMEA) on spent fuel and radioactive waste management, held in Rez, Czechoslovakia, in October 1990. Since then, the CMEA and some of its former Member States have ceased to exist. However, there are many reasons for eastern European countries to continue their regional collaboration at a higher level. The USSR, the designer and supplier of WWER type reactors in eastern European countries, participated in the first phase of the project. The majority of WWER type reactors are situated in States of the former USSR (Russia and Ukraine). The main results of the first phase of the Regional Project are: (i) Re-establishment of communication channels among eastern European countries operating WWER type reactors by incorporating the IAEA's technical assistance; (ii) Identification of common waste management problems (administrative and technical) requiring resolution; (iii) Familiarization with radioactive waste management systems at nuclear power plants with WWER type reactors - Paks (Hungary), Loviisa (Finland), Jaslovske Bohunice (Czechoslovakia) and Novovoronezh (Russian Federation). Tabs

  14. Mathematical modelling of plant transients in the PWR for simulator purposes

    International Nuclear Information System (INIS)

    Hartel, K.

    1984-01-01

    This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

  15. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  16. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  17. The advanced main control console for next japanese PWR plants

    International Nuclear Information System (INIS)

    Tsuchiya, A.; Ito, K.; Yokoyama, M.

    2001-01-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  18. Monitoring of core barrel vibrations in WWER type reactor using out-of-reactor ionization chambers

    International Nuclear Information System (INIS)

    Dach, K.

    1982-01-01

    Vibration of the core barrel is least desirable for safe operation of the PWR reactor. These mechanical vibrations are in correlation with the fluctuations of neutron flux density whose time and frequency analysis serves failure diagnosis. The mathematical model is described of the transfer of mechanical vibrations of the core barrel to neutron noise. Other steps are indicated indispensable for the application of the method of neutron noise analysis for in-service diagnostics of nuclear power plants. (Z.M.)

  19. Effects of the reactor coolant pumps following a small break in a Westinghouse PWR

    International Nuclear Information System (INIS)

    Koenig, J.E.

    1983-10-01

    Numerical simulations of the thermal-hydraulic events following a small cold-leg break in a Westinghouse pressurized water reactor were performed to address the pumps-on/off issue. The mode of pump operation was varied in each calculation to ascertain the optimum mode. It was found that pump operation was not critical for this break size and location because the fuel rods remained cool in all accidents analyzed. In terms of system mass, however, it was preferable to leave the pumps in operation

  20. Operation results of the secondary circuits of the French PWR type power plant park

    International Nuclear Information System (INIS)

    Mercier, J.P.

    1984-01-01

    Global results of performances realized since 1981 by the French PWR 900 MW power plants (installed power, availability, casual or planned shutdowns); analysis of the behaviour (casual unavailability) comparing together the performances of the different components in the secondary circuit; behaviour of the principal materials of the secondary circuit and their weight in the unavailabilities of the whole French nuclear park [fr

  1. Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-08-15

    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.

  2. Behaviour of a reactor PWR containment submitted to an external explosion

    International Nuclear Information System (INIS)

    Barbe, B.; Avet-Flancard, R.; Perrot, J.; Berriaud, C.; Dulac, J.

    1981-01-01

    The aims of this study are to obtain experimental data and theoretical evaluation of the transient field pressure existing on importants buildings of the plant. The knowledge of the pressure loading permits then to predict the structure mechanical behaviour. For this purpose the cylindrical reactor building and the parallelepipedic fuel building have been modelized to a 1/40 scale. These models were realized as carefully as possible with prestressing in the thickness of microconcrete walls and were submitted to incident shock waves obtained by T.N.T. explosions. Several characteristics explosion directions have been tested. Experimental data were recorded with pressure and displacement transducers and also by accelerometers. The results show that: 1) the geometrical dihedral between reactor and fuel building induces local overpressures five times the incident pressures; 2) no apparent damage occurred on the structure, for the range of field pressure tested so far; this may related to only small effects of resonances. Simultaneously a tridimensional, acoustic code has been developed an conveniently correlates experimental data. (orig./HP)

  3. The management routes for materials produced by the dismantling of the BR3-PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Demeulemeester, Y.; Ponnet, M.; Emond, M.; Emond, O.; Dadoumont, J.; Massaut, V.

    2000-01-01

    The dismantling of the BR3 reactor produces quite large masses of contaminated materials, mainly metals or concrete. The main management routes are: conditioning of the radioactive wastes and disposal, recycling of radioactive materials in the nuclear sector and the recycling of free released materials in the industrial sector or their evacuation as industrial waste. The conditioning of the radioactive wastes is essentially performed in the installations of Belgoprocess and must follow the specifications imposed by the national radwaste management agency ONDRAF/NIRAS. The conditioning of the pieces produced during the cutting of the reactor pressure vessel is given as example. The recycling of radioactive materials in the nuclear sector is possible for metals and for concrete. For metals, SCK.CEN has an agreement with a nuclear foundry which reuses these materials for the fabrication of shieldings. For concrete, an R and D programme is going on with the objective to demonstrate the possible reuse of baryte concrete as raw materials for the production of mortar used in the conditioning of radioactive wastes. The free release of radioactive materials and their reuse or evacuation as radioactive wastes requires the strict respect of procedures and the use of low level measurement techniques. Various decontamination techniques are used at SCK.CEN to reach this objective. For the metals, we use mainly simple washing, abrasive decontamination and hard chemical decontamination. For concrete, we use mainly scabbling or shaving techniques. (authors)

  4. Reactor internals production under ideal conditions at Pensacola. [PWR type reactors

    Energy Technology Data Exchange (ETDEWEB)

    1975-12-01

    The Westinghouse factory at Pensacola, Florida, which specialises in the production of pressure vessel internal components for PWRs, is described. Its excellent manufacturing and inspection facilities, supported by careful attention to staff training and motivation, are responsible for the extremely high level of quality and continual improvement in productivity.

  5. Operating function tests of the PWR type RHR pump for engineering safety system under simulated strong ground excitation

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, Kazuhiro; Homma, Toshiaki; Inazuka, Hisashi; Nakajima, Norifumi.

    1979-08-01

    Results are described of operating function verification tests of a PWR RHR pump during an earthquake. Of the active reactor components, the PWR residual heat removal pump was chosen from view points of aseismic classification, safety function, structural complexity and past aseismic tests. Through survey of the service conditions and structure of this pump, seismic test conditions such as acceleration level, simulated seismic wave form and earthquake duration were decided for seismicity of the operating pump. Then, plans were prepared to evaluate vibration chracteristics of the pump and to estimate its aseismic design margins. Subsequently, test facility and instrumentation system were designed and constructed. Experimental results could thus be acquired on vibration characteristics of the pump and its dynamic behavior during different kinds and levels of simulated earthquake. In conclusion: (1) Stiffeners attached to the auxiliary system piping do improve aseismic performance of the pump. (2) The rotor-shaft-bearing system is secure unless it is subjected to transient disturbunces having high frequency content. (3) The motor and pump casing having resonance frequencies much higher than frequency content of the seismic wave show only small amplifications. (4) The RHR pump possesses an aseismic design margin more than 2.6 times the expected ultimate earthquake on design basis. (author)

  6. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

    2017-04-01

    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

  7. Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

    International Nuclear Information System (INIS)

    Bernard, M.

    1995-01-01

    Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

  8. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  9. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  10. Diapo, applying advanced AI methods to diagnosis of PWR reactor coolant pump

    International Nuclear Information System (INIS)

    Porcheron, M.; Ricard, B.

    1993-01-01

    Electricite de France has decided to increase the capabilities of its monitoring and diagnostic architecture with the development of an AI system for reactor coolant pump diagnostic support. This development is carried out with the cooperation of the equipment constructor Jeumont Schneider Industries. This diagnostic system will eventually be included in an integrated surveillance architecture. We present the architecture of the system and the basics of the knowledge model used. Main data for diagnosis are provided by sensor data issued by the pump monitoring system. Diagnostic reasoning is based on the cooperation of two main activities : a heuristic search among typical symptomatic situations that leads to the formulation of hypotheses and a ''deep'' causal analysis that consists in backtracking from identified situations up to initial faults or causes. This approach is well fitted to field expert reasoning, and provides powerful diagnostic capabilities that help to overcome conventional limitations of expert systems entirely based on heuristic knowledge. (authors). 9 figs., 11 refs

  11. The results of studies of the thermohydraulics of the primary pumps in PWR reactors

    International Nuclear Information System (INIS)

    Grison, P.; Laura, J.F.

    1982-01-01

    In the context of its nuclear-safety programme for pressurized-water reactors, E.D.F. has engaged in theoretical and experimental studies in order to gain better knowledge of the behaviour of a pump under accident conditions, with passage to diphasic regime. The results of these studies are presented here, both from the experimental and theoretical points of view. They show in particular that the behaviour of the pump is essentially dictated by the interfacial friction for small flows, and by the appearance of a critical flow rate for large flows. The outline of the theoretical model describing the operation of the pump in the first three quadrants (positive and negative flow, positive and negative rotation) is described, as are certain special applications of the model, such as the determination of racing velocities in diphasic conditions [fr

  12. Results of studies of the thermohydraulics of the primary pumps in PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grison, P; Laura, J F [E.D.F., Chatou (France)

    1982-01-01

    In the context of its nuclear-safety program for pressurized-water reactors, E.D.F. has engaged in theoretical and experimental studies in order to gain better knowledge of the behaviour of a pump under accident conditions, with passage to diphasic regime. The results of these studies are presented here, both from the experimental and theoretical points of view. They show in particular that the behaviour of the pump is essentially dictated by the interfacial friction for small flows, and by the appearance of a critical flow rate for large flows. The outline of the theoretical model describing the operation of the pump in the first three quadrants (positive and negative flow, positive and negative rotation) is described, as are certain special applications of the model, such as the determination of racing velocities in diphasic conditions.

  13. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

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

    International Nuclear Information System (INIS)

    2011-08-01

    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.

  15. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method

    International Nuclear Information System (INIS)

    Silva, Alonso F.O.

    2007-01-01

    From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

  16. Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX; Evaluacion del codigo de Dinamica de Fluidos Computacional (CFD) Open FOAM en el estudio del estres termico presurizado de los reactores PWR. Comparacion con el codigo comercial Ansys-CFX

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, M.; Barrachina, T.; Miro, R.; Verdu Martin, G.; Chiva, S.

    2012-07-01

    In this work is proposed to evaluate the potential of the OpenFOAM code for the simulation of typical fluid flows in reactors PWR, in particular for the study of pressurized thermal stress. Test T1-1 has been simulated , within the OECD ROSA project, with the objective of evaluating the performance of the code OpenFOAM and models of turbulence that has implemented to capture the effect of the thrust forces in the case study.

  17. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  18. Thermal reactor strategy

    International Nuclear Information System (INIS)

    1981-01-01

    This statement sets down briefly the CEGB's views on the requirement for nuclear power and outlines current progress in the implementation of the CEGB's thermal reactor strategy. The programme is traced historically, together with statements of Government policy. The place of Magnox, AGR, SGHWR, PWR and fast breeder reactors is discussed. Advantages and problems associated with the various types are outlined. (U.K.)

  19. Dosimetry and fluence calculations on french PWR vessels comparisons between experiments and calculations

    International Nuclear Information System (INIS)

    Nimal, J.C.; Bourdet, L.; Guilleret, J.C.; Hedin, F.

    1988-01-01

    Fluence and damage calculations on PWR pressure vessels and irradiation test specimens are presented for two types of reactor: the franco-belgian (reactor CHOOZ) and the french reactors (CPY program). Comparisons with measurements are given for activation foils and fission detectors; most of them are about irradiation test specimen locations; comparisons are made for the Chooz plant on vessel stainless steel samplings and in the reactor pit

  20. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor; Ensaios de lixiviacao em produtos cimentados contendo rejeito simulado de concentrado do evaporador de reator PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes, E-mail: hauczmj@cdtn.b, E-mail: jaalmeida@cdtn.b, E-mail: tellocc@cdtn.b, E-mail: fdc@cdtn.b, E-mail: seless@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-10-26

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  1. Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

    International Nuclear Information System (INIS)

    Snaider, R.P.; Hodge, J.M.; Levin, H.A.; Zudans, J.J.

    1979-10-01

    This report summarizes work performed by the Nuclear Regulatory Commission staff and its contractor, Sandia Laboratories, in the resolution of Generic Technical Activity A-12, ''Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports.'' The report describes the technical issues, the technical studies performed by Sandia describes the technical issues, the technical studies performed by Sandia Laboratories, the NRC staff's technical positions based on these studies, and the staff's plan for implementing its technical positions. It also provides recommendations for further work. The complete technical input from Sandia Laboratories is appended to the report

  2. Effect of operating conditions and environment on properties of materials of PWR type nuclear power plant components

    International Nuclear Information System (INIS)

    Vacek, M.

    1987-01-01

    Operating reliability and service life of PWR type nuclear power plants are discussed with respect to the material properties of the plant components. The effects of the operating environment on the material properties and the methods of their determination are characterized. Discussed are core materials, such as fuel, its cladding and regulating rod materials, and the materials of pipes, steam generators and condensers. The advances in the production of pressure vessel materials and their degradation during operation are treated in great detail. (Z.M.)

  3. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  4. Conceptual study of advanced PWR systems. A study of passive and inherent safety design concepts for advanced light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; No, Hee Cheon; Baek, Won Pil; Jae, Shim Young; Lee, Goung Jin; Na, Man Gyun; Lee, Jae Young; Kim, Han Gon; Kang, Ki Sig; Moon, Sang Ki; Kim, Yun Il; Park, Jae Wook; Yang, Soo Hyung; Kim, Soo Hyung; Lee, Seong Wook; Kim, Hong Che; Park, Hyun Sik; Jeong, Ji Hwan; Lee, Sang Il; Jung, Hae Yong; Kim, Hyong Tae; Chae, Kyung Sun; Moon, Ki Hoon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-08-01

    The five thermal-hydraulic concepts chosen for advanced PWR have been studied as follows: (1) Critical Heat Flux: Review of previous works, analysis of parametric trends, analysis of transient CHF characteristics, extension of the CHF date bank, survey and assessment of correlations, design of a intermediate-pressure CHF test loop have been performed. (2) Passive Cooling Concepts for Concrete Containment system: Review of condensation phenomena with noncondensable gases, selection of a promising concept (i.e., use of external condensers), design of test loop according to scaling laws have been accomplished. and computer programs based on the control-volume approach, and the conceptual design of test loop have been accomplished. (4) Fluidic Diode Concepts: Review of previous applications of the concept, analysis major parameters affecting the performance, development of a computational code, and conceptual investigation of the verification test loop have been performed. (5) Wet Thermal Insulator: Review of previous works, selection of promising methods ( i.e. ceramic fiber in a steel case and mirror-type insulator), and conceptual design of the experimental loop have been performed. (author). 9 refs.

  5. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  6. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

    Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

    1990-01-01

    Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

  7. Method for controlling FBR type reactor

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

    1991-01-01

    The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

  8. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  9. The European pressurized water reactor. The French-German advanced PWR project

    International Nuclear Information System (INIS)

    Watteau, M.P.; Seidelberger, H.; Broecker, B.; Serviere, G.

    1995-01-01

    In order to derive full benefit from the Franco-German experience and to maintain a continuous development process, the EPR is of an evolutionary design. It is also an innovative product, intended to combine competitiveness, improved operability and enhanced safety. With a large electrical output, in the range of 1400-1500 MW, the EPR has an excellent cost-size ratio and is fully adapted to the scarcity of sites. It is also suitable for the development of scale-down products based on the same technology. The progress in safety that can be achieved with the EPR can be used to further enhance the already very high level of safety of the current French and German nuclear plants: (1) by an improvement of the preventive level of the defence-in-depth concept, and (2) by the implementation of additional features, mainly for the containment, to mitigate the consequences of severe accidents. Economically, the generation cost objective of the EPR will be at least as good as that of the French N4 reactor series. Improvements in safety may imply some additional investment costs, compared with those of the current plants, but the EPR is designed to achieve reductions of the other components of the generation costs, i.e. the fuel cycle costs and the operation and maintenance costs. The paper also describes some major design features of of the EPR: the safety systems, the containment and confinement functions, the arrangement of buildings, protection against external attacks, the man-machine interface, and instrumentation and control. 5 figs, 2 tabs

  10. Control rod for FBR type reactor

    International Nuclear Information System (INIS)

    Nakai, Koichi.

    1993-01-01

    In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

  11. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  12. Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code

    International Nuclear Information System (INIS)

    Perianez Alvarez, V.

    2013-01-01

    The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

  13. Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)

    2000-03-01

    JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

  14. A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

    International Nuclear Information System (INIS)

    Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

    2017-01-01

    This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

  15. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  16. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  17. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    2014-02-01

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  18. Development of a methodology of analysis of instabilities in BWR reactors; Desarrollo de una metodologia de analisis de inestabilidades en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fenoll, M.; Abarca, A.; Barrachina, T.; Miro, R.; Verdu, G.

    2012-07-01

    This paper presents a methodology of analysis of the reactors instabilities of BWR type. This methodology covers of modal analysis of the point operation techniques of signal analysis and simulation of transients, through 3D Coupled RELAP5/PARCSv2.7 code.

  19. Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

    International Nuclear Information System (INIS)

    Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.

    1997-01-01

    Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension

  20. Pre design processing of waste of ex-resin without materials matrix from nuclear power plant type PWR 1000 MW

    International Nuclear Information System (INIS)

    Cerdas Tarigan

    2010-01-01

    Have been done pre design processing of waste ex-resin without capacities matrix materials from nuclear power plant type PWR 1000 MW During the time radioactive waste of ex-resin processed to use process of immobilization use matrix materials like mixture cement and epoxy resin and then conditioning. This process is not effective and efficient because end result volume of end product bigger than volume early operation system and maintenance of its installation more difficult. To overcome this created a design of technology processing of waste of ex- resin without matrix materials through process of strainer, drying and conditioning represent technological innovation newly processing of radioactive waste of ex-resin. Besides this process more effective and efficient, volume of end product waste much more small from volume early and operation system and maintenance of its easier installation. Pre design is expected to be used as a basis to make conceptual of pre design installation of strainer, drying and conditioning for the processing of waste of ex-resin from nuclear power plant type PWR 1000 MW. (author)

  1. Design study on PWR-type reduced-moderation light water core. Investigation of core adopting seed-blanket fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    As a part of the design study on PWR-type Reduced-Moderation Water Reactors (RMWRs), a light water cooled core with the seed-blanket type fuel assemblies has been investigated. An assembly with seed of 13 layers and blanket of 5 layers was selected by optimization calculations. The core was composed with the 163 assemblies. The following results were obtained by burn-up calculations with the MVP-BURN code; The cycle length is 15 months by 3-batch refueling. The discharge burn-up including the inner blanket is about 25 GWd/t. The conversion ratio is about 1.0. The void reactivity coefficient is about-26.1 pcm/%void at BOC and -21.7pcm%void at EOC. About 10% of MA makes conversion ratio decrease about 0.05 to obtain the same burn-up. The void reactivity coefficient increased significantly and it is necessary to reduce it. FP amount corresponding to about 2 % of total plutonium weight makes reactivity decrease about 0.5 %{delta}k/k and void reactivity coefficient increase, however these changes are within the design margins. Capability of multi-recycling of plutonium was confirmed, using discharged plutonium for 4 cycles, if fissile plutonium of 15.5wt% is used. The conversion ratio increases by about 0.026 with recycling. However, void reactivity coefficient increases and some effort to obtain negative void reactivity coefficient is necessary. (author)

  2. Effects of aging in containment spray injection system of PWR reactor containment; Efeitos do envelhecimento no sistema de injecao de borrifo da contencao de reatores a agua pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L., E-mail: diogosb@outlook.com, E-mail: deise_dy@hotmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper presents a contribution to the study of the components aging process in commercial plants of Pressurized Water Reactors (PWR). The analysis is done by applying the method of Fault trees, Monte Carlo Method and Fussell-Vesely Importance Measurement. The study on the aging of nuclear plants, is related to economic factors involved directly with the extent of their operational life, and also provides important data on issues of safety. The most recent case involving the process of extending the life of a PWR plant can be seen in Angra 1 Nuclear Power Plant by investing $ 27 million in the installation of a new reactor cover. The corrective action generated an extension of the useful life of Angra 1 estimated in twenty years, and a great savings compared to the cost of building a new plant and the decommissioning of the first, if it had reached the operation time out 40 years. The extension of the lifetime of a nuclear power plant must be accompanied by special attention from the most sensitive components of the systems to the aging process. After the application of the methodology (aging analysis of Containment Spray Injection System (CSIS)) proposed in this paper, it can be seen that increasing the probability of failure of each component, due to the aging process, generate an increased general unavailability of the system that contains these basic components. The final results obtained were as expected and can contribute to the maintenance policy, preventing premature aging in nuclear power systems.

  3. Study of the influence of temperature and time on the electroplating nickel layer in Inconel 718 strips used in spacer grid of Pressurized Water Cooled nuclear reactors (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Renato; Abati, Amanda; Verne, Júlio; Panossian, Zehbour, E-mail: amanda.abati@marinha.mil.br, E-mail: jvernegropp@gmail.com, E-mail: renato.rezende@marinha.mil.br, E-mail: zep@ipt.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil). Laboratório de Desenvolvimento e Instrumentação de Combustível Nuclear; Instituto de Pesquisas Tecnológicas (IPT), São Paulo, SP (Brazil)

    2017-07-01

    The Inconel 718 (UNS N07718: Ni-{sup 19}Cr-{sup 18}Fe-{sup 5}Nb-3 Mo) is a precipitation hardenable nickel alloy that has good corrosion resistance and high mechanical strength. These strips are used for assembling the spacer grid of fuel element of pressurized water cooled nuclear reactors (PWR). The spacer grid is a structural component of fundamental importance in fuel elements of PWR reactors, maintaining the position and necessary spacing of the fuel rods within the arrangement of the fuel element. The spacer grid is formed by joining the points of intersection of the strips, by a joint process called brazing. For this process, these strips are stamped and plated with a thin layer of nickel by means of electroplating in order to protect against oxidation and allow a better flowability and wettability of the addition metal in the strips during brazing. Oxidation at the surface of the base material harms wettability and inhibits spreading of the liquid addition metal on the substrate surface during the brazing process. The use of coatings such as nickel plating is used to ensure such conditions. The results showed that there is a process of diffusion de some chemical elements such as chromium, iron, titanium and aluminum from the substrate to the nickel layer and nickel from the layer to the substrate. These chemical elements are responsible for the oxidation at the surface of the strip. (author)

  4. Water-immersion type ship reactor

    International Nuclear Information System (INIS)

    Okada, Hiroki; Yamamura, Toshio.

    1996-01-01

    In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

  5. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  6. Efforts onto nuclear research and development such as new reactor and so forth

    International Nuclear Information System (INIS)

    Onishi, Tuneji

    2000-01-01

    The Japan Atomic Power Co. which is one of specified business company on nuclear power generation, has carried out construction and operation of power plants with different types of reactor such as boiling light water reactor (BWR), pressurized light water rector (PWR), and so forth. And, by actively using technical powers and experiences accumulated before then, additional construction of a new power unit, and researches and developments on a simplified light water reactor, a future type rector, and a high breeder proof reactor have been made some efforts. Here were introduced some outlines on development of an improved type PWR, development of a new type reactor for example, deep embedded plant), future type reactor (for example, revolutionary middle and small type reactor, simplified PWR, and simplified BWR), a fast breeder reactor, and a reactor building suitable for a ship shell structure. (G.K.)

  7. Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

    Energy Technology Data Exchange (ETDEWEB)

    Bocanegra, R.; Jimenez, G.

    2013-07-01

    The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

  8. A study on the pressurized water reactor (PWR) containment response analysis methodologies for postulated severe accident

    International Nuclear Information System (INIS)

    Ahn, Kwang Il

    1992-02-01

    The present study contains two major parts: one is the treatment of uncertainties involved in the current APET and the other is the importance analysis of the APET uncertainty inputs. A clear disadvantage of the expert opinion polling process approach for uncertainty analysis of the current probabilistic risk assessment (PRA) is that the sufficient robustness in the final results may not be attained against the ambiguity of the information upon which the experts base their judgement or the judgmental uncertainty arising under various imprecise and incomplete information. For the treatment of such type of uncertainty, a new approach based on fuzzy set theory is proposed. Then its potential use to the uncertainty analysis of the current PRA is proved through an analysis of accident progression event tree (APET). As a product, a formal procedure with computational algorithms suitable for application of the fuzzy set theory to the APET analysis is provided. Comparing with the uncertainty analysis results obtained by the statistical approach currently used in PRA, the present approach has several major advantages: Firstly, it greatly enhances the robustness in the final results of APET uncertainty analysis by modeling the judgmental uncertainty that arises in the probabilistic quantification of APET top events. Secondly, the modeling of APET uncertainty analysis is far more convenient because of the nonprobabilistic features of fuzzy probabilities used for uncertainty quantification of the APET top events. Thirdly, the APET model can easily be operated by means of a well defined formal propagation logic of fuzzy set theory without going through a tedious sampling procedure. Finally, the fuzzy outcomes provide at least as much information as the existing methods based on the statistical approach. Thus, the present approach can be used as a valuable alternative approach to uncertainty analysis used in the current PRA. Two importance measures for the importance analysis of

  9. Method and Result of Experiment for Support of Technical Solutions in the Field of Perfection of a Nuclear Fuel Cycle for Future PWR Reactors

    International Nuclear Information System (INIS)

    Ostrovskiy, V.; Kudryavtsev, E.; Tutnov, I.

    2011-01-01

    The paper presents the basics of approach of planning and carrying out of experiments to validate safety PWR reactors of the future when accepting technical solutions concerning using of improved fuel rods in fuel assembly. Basic principles and criteria used for the validation of technical solutions and developments in improving of nuclear fuel cycle of PWR reactors of the future are presented from the point of safety of future operation of modified fuel rods. We explore the questions of safety operation of PWR reactors with fuel assemblies, containing fuel rods with different length of fuel. The paper discusses the ways of solving of important tasks of critical facility experiments conducting for verification of new technical solutions in the sphere of PWR nuclear fuel cycle improvement on the base of international standards ISO 2000:9000 and functional safety recommendations of IEC (International Electromechanical Commission). New Federal laws of Russian Federation define the main principle for demands to NPP and any supplier of nuclear techniques. The principle is 'quantity indicators of risk should not exceed comprehensible social size of the established indicators of safety for any moment of operation of NPP'. On the other hand the second principle should be applied to extraction of the greatest benefit from operation of the equipment, systems or the NPP as whole: 'The long operation and full commercial use of resource and service properties of the equipment, systems and the NPP as a whole'. Realization of this principle assumes development and introduction of new technical solutions for a validation of guarantees of safety of the future operation of NPP or it separate components. Solving the practical problems of a validation of safety use of fuel rods with the increased length of a fuel column in fuel assembly in nuclear reactors of the future, we should choose new strategies and programs of verification experiments on the base of the analysis of guarantees

  10. Rod consolidation of RG and E's [Rochester Gas and Electric Corporation] spent PWR [pressurized water reactor] fuel

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister

  11. Development of computational program for studying the reactor control system in PWR plants; Desenvolvimento de um programa computacional para estudo do sistema de controle do reator em plantas PWR

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Jose Ricardo de; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    In this work a computational program is presented which has been developed for specific application on the study of the reactor control system of a typical PWR plant. As to the basic function of simulating power transients the program has the following structure: a representative mathematical model of the dynamic and stationary behaviors of the primary circuit; a group of equations associated to the reactor power control and system pressure control; screens for the entry of reference data as well as of control blocks and control bar speed programming module parameters; main entering screens for the configuration of the excitement/transient function as well as of simulation time and control mood; and graphical output of all the process variables incorporated to the model. As premise it has been considered as sufficient the modeling of the primary circuit, a differential equation being used which associates the average temperature of the coolant within the steam generator with the potency transferred to the secondary circuit, denominated 'secondary potency', as an interface with the secondary circuit. Every transient - ramp or step - is established upon the 'turbine power' variable, which in turn is related to the 'secondary power' variable by means of a differential equation that represents a first - order delay, having adjustable parameters on the data - entry screen. In the neutronic model as defined for the reactor, the reactivity feedback effects due to primary circuit pressure variation, as well as fuel and coolant temperature variation, were taken into consideration. Thermo-hydraulics constants and project data taken from the available bibliography, adapted to a particular small PWR unit conception , were employed for loading the program. With the open-loop simulation results a positive qualitative evaluation of the program was obtained, in comparison to published results related to simulators bearing equal purposes, more

  12. Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

    International Nuclear Information System (INIS)

    Kheshtpaz, H.; Alison, C.

    2006-01-01

    As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

  13. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

  14. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

    The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

  15. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  16. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  17. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  18. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  19. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  20. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  1. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  2. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  3. Overview of recycling technologies for decommissioned materials. Lessons learned during the dismantling of a small PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Emond, O.; Ponnet, M.

    2001-01-01

    Full text: SCK CEN is dismantling its 11 MWe PWR reactor. The reactor was shutdown in 1987 after 25 years of operation and the dismantling started in 1990. For the management of the low radioactive materials, we apply a strategy promoting the minimisation of the production of radioactive waste and hence the maximisation of the production of recycled materials while keeping the costs as low as possible. The recycled materials are either reused in the non- nuclear industry as raw materials (metal scrap industry or building industry for the concrete) or recycled in the nuclear industry for specific applications (reuse of metals for fabrication of shielding, potential reuse of concrete for production of 'radioactive mortar'). The clearance of radioactive materials and their reuse require the strict respect of procedures and specifications. In our case, the Health Physics department under supervision of the Competent Authority establishes the procedures. This procedure is still a case by case practice but the legislation in Belgium is progressively put in place. For the recycling in the nuclear industry, we must respect the specifications of the end-user. Up to now, we have recycled low radioactive metals for the fabrication of shielding in the USA, so we had to respect the specifications of the melting facility and to obtain the authorisations for the transport abroad and for the transfer of property. Besides the radioactive waste route, we are using several evacuation routes for the dismantled materials: Evacuation of the cleared metals (iron, stainless steel, copper, electric motors...) to a local scrap dealer; Evacuation of metals to the Studsvik melting facility situated in Sweden: after clearance by the Swedish Authority, the non radioactive materials are sent to a local scrap dealer and the secondary radioactive waste is sent back to Belgium and conditioned by Belgoprocess. This technology further decontaminates the metals and allows performing an accurate

  4. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-07-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  5. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  6. The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

    International Nuclear Information System (INIS)

    Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

    2010-01-01

    LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

  7. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  8. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  9. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  10. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  11. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  12. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  13. A thermal hydraulic analysis in PWR reactors with UO{sub 2} or (U-Th)O{sub 2} fuel rods employing a simplified code

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Thiago A. dos; Maiorino, José R., E-mail: thiago.santos@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil); Stefanni, Giovanni L. de, E-mail: giovanni.stefanni@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O{sub 2}. For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O{sub 2}.The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO{sub 2} was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

  14. A thermal hydraulic analysis in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

    International Nuclear Information System (INIS)

    Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de

    2017-01-01

    In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

  15. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  16. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  17. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  18. Technical description of other types of reactors

    International Nuclear Information System (INIS)

    Vollmer, H.

    1977-01-01

    The paper reviews the development of reactor systems other than LWR, i. e. gas cooled reactors, heavy water reactors and fast breeders. The specific features of these reactors are discussed. Technical details on plant design of the various systems will be given as well as the present status-of-the-art. (orig.) [de

  19. Global calculation of PWR reactor core using the two group energy solution by the response matrix method

    International Nuclear Information System (INIS)

    Conti, C.F.S.; Watson, F.V.

    1991-01-01

    A computational code to solve a two energy group neutron diffusion problem has been developed base d on the Response Matrix Method. That method solves the global problem of PWR core, without using the cross sections homogenization process, thus it is equivalent to a pontwise core calculation. The present version of the code calculates the response matrices by the first order perturbative method and considers developments on arbitrary order Fourier series for the boundary fluxes and interior fluxes. (author)

  20. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC

    International Nuclear Information System (INIS)

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-01-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.