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Sample records for subcritical reactor systems

  1. Activity report of working party on reactor physics of subcritical system. October 2001 to March 2003

    International Nuclear Information System (INIS)

    2004-03-01

    Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Subcritical System (ADS-WP) was set in July 2001 to research reactor physics of subcritical system such as Accelerator-Driven System (ADS). The WP, at the first meeting, discussed a guideline of its activity for two years and decided to perform theoretical research for the following subjects: (1) study of reactor physics for a subcritical core, (2) benchmark problems for a subcritical core and their calculations, (3) study of physical parameters affecting to set subcriticality of ADS, and (4) study of measurement and surveillance methods of subcriticality of a subcritical core. The activity of ADS-WP continued up to March 2003. In this duration, the members of the WP met together eight times, including four meetings jointly held with the Workshop on Accelerator-Driven Subcritical Reactor at Kyoto University Research Reactor Institute. This report summarizes the result obtained by the above WP activity and research. (author)

  2. Subcriticality determination of nuclear reactor

    International Nuclear Information System (INIS)

    Borisenko, V.I.; Goranchuk, V.V.; Sidoruk, N.M.; Volokh, A.F.

    2014-01-01

    In this article the subcriticality determination of nuclear reactor is considered. Emphasized that, despite the requirements of regulatory documents on the subcriticality determination of WWER from the beginning of their operation, so far, this problem has not been solved. The results of subcriticality determination of Rossi-α method of the WWER-M is presented. The possibility of subcriticality determination of WWER is considered. The possibility of subcriticality determination of Rossi-α method with time resolution is of about 100 microseconds is also considered. The possible reasons for the error in subcriticality determination of the reactor are indicated

  3. Applicability of Avery's coupled reactor theory to estimate subcriticality of test region in two region system

    International Nuclear Information System (INIS)

    Kugo, Teruhiko

    1992-01-01

    The author examined the validity to estimate the subcriticality of a test region in a coupled reactor system using only measurable quantities on the basis of Avery's coupled reactor theory. For the purpose, we analyzed coupled reactor experiments performed at the Tank-type Critical Assembly in Japan Atomic Energy Research Institute by using two region systems and evaluated the subcriticality of the test region through a numerical study. Coupling coefficients were redefined at the quasi-static state because their definitions by Avery were not clear. With the coupling coefficients obtained by the numerical calculation, the multiplication factor of the test region was evaluated by two formulas; one for the evaluation using only the measurable quantities and the other for the accurate evaluation which contains the terms dropped in the former formula by assuming the unchangeableness for the perturbation induced in a driver region. From the comparison between the results of the evaluations, it was found that the estimation using only the measurable quantities is valid only for the coupled reactor system where the subcriticality of the test region was very small within a few dollars in reactivity. Consequently, it is concluded that the estimation using only the measurable quantities is not applicable to a general coupled reactor system. (author)

  4. Subcriticality monitoring method for reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto.

    1991-01-01

    The present invention accurately monitors the reactor subcriticality and ensures the critical safety, irrespective of the presence or absence of artificial neutron sources. That is, when the subcriticality is monitored upon reactivity changing operation which causes reactivity change to the reactor during shutdown, neutron monitors are disposed at a plurality of monitoring positions. Then, neutron counting ratio before and after conducting the reactivity changing operation is determined. The subcriticality of the reactor is monitored by the ratio and the state of scattering of the ratio of neutron counting rate between each of the neutron monitors. With such procedures, signals of the neutron monitors are used, the characteristic that the change of the signals depend on the change of the neutron multiplication of the reactor core can be utilized whether artificial neutron sources (external neutron sources) are disposed or not. Accordingly, the subcriticality can be monitored more reliably. (I.S.)

  5. Cost optimization of ADS design: Comparative study of externally driven heterogeneous and homogeneous two-zone subcritical reactor systems

    International Nuclear Information System (INIS)

    Gulik, Volodymyr; Tkaczyk, Alan H.

    2014-01-01

    Highlights: • The optimization of two-zone homogeneous subcritical systems has been performed. • A Serpent model for two-zone heterogeneous subcritical systems has been developed. • The optimization of two-zone heterogeneous subcritical systems has been carried out. • Economically optimal core composition of two-zone subcritical system was found. • The neutron spectra of the heterogeneous subcritical systems have been obtained. - Abstract: Subcritical systems driven by external neutron sources, commonly known as Accelerator-Driven System (ADS), are one type of advanced nuclear reactor exhibiting attractive characteristics, distinguished from the traditional critical systems by their intrinsic safety features. In addition, an ADS can be used for the transmutation of the nuclear waste, accumulated during the operation of existing reactors. The optimization of a subcritical nuclear reactor in terms of materials (fuel content, coolant, etc.), geometrical, and economical parameters is a crucial step in the process of their design and construction. This article describes the optimization modeling performed for homogeneous and heterogeneous two-zone subcritical systems in terms of geometry of the fuel zones. Economical assessment was also carried out for the costs of the fuel in the core of the system. Optimization modeling was performed with the Serpent-1.1.18 Monte Carlo code. The model of a two-zone subcritical system with a fast inner and a thermal gas-cooled graphite-moderated outer zone was developed, simulated, and analyzed. The optimal value for the pitch of fuel elements in the thermal outer zone was investigated from the viewpoint of the cost of subcritical system. As the main goal of ADS development is nuclear waste transmutation, neutron spectra for both fast and thermal zones were obtained for different system configurations. The results of optimization modeling of homogeneous and heterogeneous two-zone subcritical systems show that an optimal

  6. Accelerator driven subcritical reactors

    International Nuclear Information System (INIS)

    Salvatores, M.

    2001-01-01

    : 241 Am, 242m Am, 243 Am and Cm isotopes, since the typical target accuracies required are not yet achieved; review the consistency of data and eliminate non-physical data (e.g. 242 Cm σ f and 243 Cm σ in in ENDF/B-VI); perform extensive sensitivity studies for significant parameters of the subcritical core for systems with different ratios of Pu and MA in the fuel and for the initial stage and at end of cycle; review uncertainties and, more generally, the status of data for the calculation of β eff , decay heat, γ-heating for the systems mentioned; review the status of data for materials which are potential new candidates as core support materials (e.g., MgO, Zr, Ti), for Pb, Bi, N, 15 N (σ t , σ nγ , σ el , σ inel , σ np below 20 MeV) and review the 242 Pu and 238 Pu data accuracy, (in particular σ nγ ), and assess uncertainty on MA and Pu isotopes σ f in the region 1-20 MeV

  7. Activities of working party on 'Subcritical core of accelerator-driven system' under the research committee on reactor physics of AESJ and JAERI

    International Nuclear Information System (INIS)

    Iwasaki, T.; Tsujimoto, K.; Nishihara, K.; Kitamura, Y.

    2004-01-01

    The Research Committee on Reactor Physics under the Atomic Energy Society of Japan and the Japan Atomic Energy Research Inst. organized the working party (ADS-WP) on S ubcritical Core of Accelerator-Driven System . The ADS-WP investigated reactor physics of subcriticality from the viewpoint of the accelerator driven system (ADS) since subcriticality has been almost studied from the viewpoint of critical safety. The working party was set in July 2001 and it worked for two years. The activities of the ADS-WP are (Work-I) theory of subcriticality, (Work-II) benchmark of subcritical core, (Work-III) setting of subcriticality level of ADS and (Work-JAO monitoring of subcriticality. These activities clarified about the important issues related to the subcriticality or the subcritical core from the wide ranges of theory, analysis, calculation, design and monitoring for ADS. The activities were already summarized and the report will be published in March 2004. (authors)

  8. Implementation of the neutron noise technique for subcritical reactors using a new data acquisition system

    International Nuclear Information System (INIS)

    Bellino, Pablo A.; Gomez, Angel

    2009-01-01

    A new data acquisition system was designed and programmed for nuclear kinetics parameter estimations in subcritical reactors. The system allows using any of the neutron noise techniques, since it could store the whole information available in the neutron detection system. The α Rossi, α Feynman and spectral analysis methods were performed in order to estimate the prompt neutron decay constant (and hence the reactivity). The measurements were done in the nuclear research reactor RA-1, where introducing the control rods, different reactivity levels where reached (until -7 dollars). With the three methods used, agreement was found between the estimations and the reference reactivities in each level, even when the detector efficiency was low. All the measurements were performed with a high gamma flux, although the results were found to be satisfactory. (author)

  9. Journey from discovery of nuclear fission to accelerator-driven sub-critical reactor systems (ADS)

    International Nuclear Information System (INIS)

    Kapoor, S.S.

    2005-01-01

    The epoch making discovery of nuclear fission in 1939, which resulted purely from the curiosity driven basic research to understand the atomic and nuclear structure has changed the world forever with the onset of a new era in the history of human civilization. The basic nuclear physics research pursued after the discovery of fission has also been of much relevance in the harnessing of nuclear energy. In the recent years, there is considerable interest towards developing accelerator driven sub-critical reactor systems (ADS) for the incineration of the long-lived spent fuel radioactive waste and for the utilization of thorium fuel for nuclear power generation. In this talk, we discuss important milestones in the journey from discovery of nuclear fission to ADS. (author)

  10. Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Klain, Kimberly L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-21

    The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set of multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the

  11. Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    2001-01-01

    The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k eff ∼0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k eff ∼0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early

  12. Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T.

    1997-01-01

    The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

  13. Electron versus proton accelerator driven sub-critical system performance using TRIGA reactors at power

    International Nuclear Information System (INIS)

    Carta, M.; Burgio, N.; D'Angelo, A.; Santagata, A.; Petrovich, C.; Schikorr, M.; Beller, D.; Felice, L. S.; Imel, G.; Salvatores, M.

    2006-01-01

    This paper provides a comparison of the performance of an electron accelerator-driven experiment, under discussion within the Reactor Accelerator Coupling Experiments (RACE) Project, being conducted within the U.S. Dept. of Energy's Advanced Fuel Cycle Initiative (AFCI), and of the proton-driven experiment TRADE (TRIGA Accelerator Driven Experiment) originally planned at ENEA-Casaccia in Italy. Both experiments foresee the coupling to sub-critical TRIGA core configurations, and are aimed to investigate the relevant kinetic and dynamic accelerator-driven systems (ADS) core behavior characteristics in the presence of thermal reactivity feedback effects. TRADE was based on the coupling of an upgraded proton cyclotron, producing neutrons via spallation reactions on a tantalum (Ta) target, with the core driven at a maximum power around 200 kW. RACE is based on the coupling of an Electron Linac accelerator, producing neutrons via photoneutron reactions on a tungsten-copper (W-Cu) or uranium (U) target, with the core driven at a maximum power around 50 kW. The paper is focused on analysis of expected dynamic power response of the RACE core following reactivity and/or source transients. TRADE and RACE target-core power coupling coefficients are compared and discussed. (authors)

  14. Reactor noise in critical and accelerator driven sub-critical systems

    International Nuclear Information System (INIS)

    Degweker, S.B.; Rana, Y.S.

    2007-01-01

    Noise methods have long been used for reactor kinetics parameters measurement and as diagnostic tools for monitoring the health of a nuclear power plant. It is conceivable that noise techniques would find similar applications in ADS. Measurement/monitoring the degree of sub-criticality of an ADS is one such application for which noise based methods are being considered, among others such as the pulsed source method. For this reason, theoretical studies on ADS noise have appeared since the late nineties. The principal difference between critical reactor noise and ADS noise is due to the statistical properties of the source. Unlike the source due to radioactive decay present in ordinary reactors, the machine produced ADS source cannot be assumed to be a Poisson process. In addition the source is pulsed. All this requires a new theoretical approach to the subject. In a number of papers (beginning in 2000) such a theoretical approach has been developed in BARC. Over the years, our approach has received general acceptance. The paper gives a description of the subject of reactor noise and its applications in critical reactors. The theory of noise in ADS is then outlined, highlighting the differences in approach and results from that of critical reactors. (author)

  15. A novel concept for CRIEC-driven subcritical research reactors

    International Nuclear Information System (INIS)

    Nieto, M.; Miley, G.H.

    2001-01-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

  16. Omega-mode perturbation theory and reactor kinetics for analyzing accelerator-driven subcritical systems

    International Nuclear Information System (INIS)

    Ren-Tai, Chiang

    2003-01-01

    An ω-mode first-order perturbation theory is developed for analyzing the time- and space-dependent neutron behavior in Accelerator-Driven Subcritical Systems (ADSS). The generalized point-kinetics equations are systematically derived using the ω-mode first-order perturbation theory and Fredholm Alternative Theorem. Seven sets of the ω-mode eigenvalues exist with using six groups of delayed neutrons and all ω eigenvalues are negative in ADSS. Seven ω-mode adjoint and forward eigenfunctions are employed to form the point-kinetic parameters. The neutron flux is expressed as a linear combination of the products of seven ω-eigenvalue-mode shape functions and their corresponding time functions up to the first order terms, and the lowest negative ω-eigenvalue mode is the dominant mode. (author)

  17. Continuous reactivity calculation for subcritical system

    International Nuclear Information System (INIS)

    Silva, Cristiano; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da

    2011-01-01

    With the rise of a new generation of nuclear reactors as for existence the ADS (Accelerator-Driven System), it is important to have a fast and accurate prediction of the variation in reactivity during a possible variation in the intensity of external sources. This paper presents a formulation for the calculation of reactivity in subcritical systems using the inverse method related only to nuclear power derivatives. One of the applications of the proposed method is the possibility of developing reactimeters that allow the continuous monitoring of subcritical systems. (author)

  18. Continuous reactivity calculation for subcritical system

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Cristiano; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da, E-mail: cristiano@herzeleid.net, E-mail: aquilino@lmp.ufrj.br, E-mail: fernando@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, 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)

    2011-07-01

    With the rise of a new generation of nuclear reactors as for existence the ADS (Accelerator-Driven System), it is important to have a fast and accurate prediction of the variation in reactivity during a possible variation in the intensity of external sources. This paper presents a formulation for the calculation of reactivity in subcritical systems using the inverse method related only to nuclear power derivatives. One of the applications of the proposed method is the possibility of developing reactimeters that allow the continuous monitoring of subcritical systems. (author)

  19. Comparative analysis of sub-critical transmutation reactor concepts

    International Nuclear Information System (INIS)

    Chang, S. H.

    1997-01-01

    The long-lived nuclear wastes have been substantially generated from the light water reactor for a few decades. The toxicity of these spent fuels will be higher than that of the uranium ore, even if those will be stored in the repository more than ten thousands. Hence the means of transmuting the key long-lived nuclear wastes, primarily the minor actinides, using a hybrid proton accelerator and subcritical transmutation reactor, are proposed. Until now, the representative concepts for a subcritical transmutation reactor are the Energy Amplifier, the OMEGA project, the ATW and the MSBR. The detailed concepts and the specifications are illustrated in Table 1. The design requirements for the subcritical transmutation reactor are the high transmutation rate of long-lived nuclear wastes, safety and economics. And to propose the subcritical transmutation reactor concepts, the coolant, the target material and fuel type are carefully considered. In these aspects, the representative concepts for a subcritical transmutation reactor in Table 1 have been surveyed. The requirements for a target and a coolant are the reliable, low maintenance operation and safe operation to minimize the wastes. The reliable, low maintenance operation and safe operation to minimize the wastes. The reliable coolant must have the low melting point, high heat capacity and excellent physical properties. And the target material must have high neutron yield for a given proton condition and easy heat removal capability. Therefore in respect with the above requirements, Pb-Bi is proposed as the coolant and the target material for the subcritical reactor. Because the neutron yield for a given proton energy increases linearly with mass number up to bismuth but in heavier elements spallation events sharply increase both the neutron and heat outputs, Pb-Bi meets not only such the requirements as the above for the coolant but also those for the coolant and target, the simplification of system can be achieved

  20. Design project of fast subcritical system 'Mala Lasta'

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Popovic, D.; Pesic, M.; Zavaljevski, N.; Nikolic, D.; Arsenovic, M.

    1988-10-01

    This report contains two parts. Part one covers the objective and fundamental elements for the choice of fast subcritical system 'Mala Lasta', review of the existing fast subcritical assemblies, and a description of the available domestic computer codes applied for calculating neutron reactor parameters. Comparison of results obtained by these codes for a number of existing subcritical assemblies was used for the choice of the design project described in part two of this report. It contains detailed description of the operating parameters of the chosen subcritical system based on the obtained calculated parameters

  1. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  2. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  3. Conceptual research on reactor core physics for accelerator driven sub-critical reactor

    International Nuclear Information System (INIS)

    Zhao Zhixiang; Ding Dazhao; Liu Guisheng; Fan Sheng; Shen Qingbiao; Zhang Baocheng; Tian Ye

    2000-01-01

    The main properties of reactor core physics are analysed for accelerator driven sub-critical reactor. These properties include the breeding of fission nuclides, the condition of equilibrium, the accumulation of long-lived radioactive wastes, the effect from poison of fission products, as well as the thermal power output and the energy gain for sub-critical reactor. The comparison between thermal and fast system for main properties are carried out. The properties for a thermal-fast coupled system are also analysed

  4. Choosing the optimal parameters of subcritical reactors driven by accelerators

    International Nuclear Information System (INIS)

    Khudaverdyan, A.G.; Zhamkochyan, V.M.

    1998-03-01

    Physical aspects of a subcritical Nuclear Power Plants (NPP) driven by proton accelerators are considered. Estimated theoretical calculations are made for subcritical regimes of various types of reactors. It was shown that the creation of the quite effective explosion-safe NPP is real at an existing level of the accelerator technique by using available reactor units (including the serial ones). (author)

  5. Comparative analysis of operation and safety of subcritical nuclear systems and innovative critical reactors; Analyse comparative du fonctionnement et de la surete de systemes sous-critiques et de reacteurs critiques innovants

    Energy Technology Data Exchange (ETDEWEB)

    Bokov, P.M

    2005-05-01

    The main goal of this thesis work is to investigate the role of core subcriticality for safety enhancement of advanced nuclear systems, in particular, molten salt reactors, devoted to both energy production and waste incineration/transmutation. The inherent safety is considered as ultimate goal of this safety improvement. An attempt to apply a systematic approach for the analysis of the subcriticality contribution to inherent properties of hybrid system was performed. The results of this research prove that in many cases the subcriticality may improve radically the safety characteristics of nuclear reactors, and in some configurations it helps to reach the 'absolute' intrinsic safety. In any case, a proper choice of subcriticality level makes all analyzed transients considerably slower and monotonic. It was shown that the weakest point of the independent-source systems with respect to the intrinsic safety is thermohydraulic unprotected transients, while in the case of the coupled-source systems the excess reactivity/current insertion events remain a matter of concern. To overcome these inherent drawbacks a new principle of realization of a coupled sub-critical system (DENNY concept) is proposed. In addition, the ways to remedy some particular safety-related problems with the help of the core sub-criticality are demonstrated. A preliminary safety analysis of the fast-spectrum molten salt reactor (REBUS concept) is also carried out in this thesis work. Finally, the potential of the alternative (to spallation) neutron sources for application in hybrid systems is examined. (author)

  6. Subcritical enhanced safety molten-salt reactor concept

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Ignatiev, V.V.; Men'shikov, L.I.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.; Krasnykh, A.K.; Rudenko, V.T.; Somov, L.N.

    1995-01-01

    The nuclear power and its fuel cycle safety requirements can be met in the main by providing nuclear power with subcritical molten salt reactors (SMSR) - 'burner' with an external neutron source. The utilized molten salt fuel is the decisive advantage of the SMSR over other burners. Fissile and fertile nuclides in the burner are solved in a liquid salt in the form of fluorides. This composition acts simultaneously as: a) fuel, b) coolant, c) medium for chemical partitioning and reprocessing. The effective way of reducing the external source power consists in the cascade neutron multiplication in the system of coupled reactors with suppressed feedback between them. (author)

  7. Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor; Diseno y construccion de un sistema electronico automatico de medicion y graficado del flujo neutronico para el reactor subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

  8. Analysis of neutronics and dynamic characteristics with reactivity injection in LBE cooled sub-critical reactor

    International Nuclear Information System (INIS)

    Chen Sen; Wu Yican; Jin Ming; Chen Zhibin; Bai Yunqing; Zhao Zhumin

    2014-01-01

    Accelerator Driven Sub-critical System (ADS) has particular neutronics behaviors compared with the critical system. Prompt jump approximation point reactor kinetics equations taken external source into account have been deduced using an approach of prompt jump approximation. And the relationship between injection reactivity and power ampliation has been achieved. In addition, based on the RELAP5 code the prolong development of point reactor kinetics code used into assessing sub-critical system have been promoted. Different sub-criticality (k eff = 0.90, 0.95, 0.97, 0.98 and 0.99) have been assessed in preliminary design of a type of natural circulation cooling sub-critical reactor under conditions of reactivity injection +1 β in one second. It shows that the external source prompt transient approximation method has an accurate solution after injecting reactivity around short time and has a capacity to solve the dynamic equation, and the sub-critical system has an inner stability while the deeper sub-criticality the less impact on the sub-critical system. (authors)

  9. Designing a mini subcritical nuclear reactor

    International Nuclear Information System (INIS)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M.

    2015-10-01

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of 239 PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of 239 PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k e -f f , the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k eff , the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons 239 PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k eff was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k eff of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five cylinders not met. (Author)

  10. Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

    2009-01-01

    The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

  11. The spatial kinetic analysis of accelerator-driven subcritical reactor

    International Nuclear Information System (INIS)

    Takahashi, H.; An, Y.; Chen, X.

    1998-02-01

    The operation of the accelerator driven reactor with subcritical condition provides a more flexible choice of the reactor materials and of design parameters. A deep subcriticality is chosen sometime from the analysis of point kinetics. When a large reactor is operated in deep subcritical condition by using a localized spallation source, the power distribution has strong spatial dependence, and point kinetics does not provide proper analysis for reactor safety. In order to analyze the spatial and energy dependent kinetic behavior in the subcritical reactor, the authors developed a computation code which is composed of two parts, the first one is for creating the group cross section and the second part solves the multi-group kinetic diffusion equations. The reactor parameters such as the cross section of fission, scattering, and energy transfer among the several energy groups and regions are calculated by using a code modified from the Monte Carlo codes MCNPA and LAHET instead of the usual analytical method of ANISN, TWOTRAN codes. Thus the complicated geometry of the accelerator driven reactor core can be precisely taken into account. The authors analyzed the subcritical minor actinide transmutor studied by Japan Atomic Energy Research Institute (JAERI) using the code

  12. Advisory group meeting on design and performance of reactor and subcritical blanket systems with lead and lead-bismuth as coolant and/or target material. Working material

    International Nuclear Information System (INIS)

    2001-01-01

    The purpose of the IAEA Advisory Group Meeting (AGM) on Design and Performance of Reactor and Sub-critical Blanket Systems with Lead and Lead-Bismuth as Coolant and/or Target Material was to provide a forum for international information exchange on all the topics relevant to Pb and Pb/Bi cooled critical and sub-critical reactors. In addition, the AGM aimed at: (1) finding ways and means to improve international co-ordination efforts in this area; (2) obtaining advice from the Member States with regard to the activities to be implemented in this area by the IAEA, in order to best meet their needs; and (3) laying out the plans for an effective co-ordination and support of the R and D activities in this area. The AGM stressed that nuclear energy is a realistic solution to satisfy the energy demand, considering the limited resources of fossil fuel, its uneven distribution in the world and the impact of its use on the planet, and taking into account the expected doubling of the world population in the 21st century and tripling of the electricity demand (especially in the developing countries). However, the AGM concluded that the development of an innovative nuclear technology meeting the following requirements must be pursued: (a) deterministic exclusion of any severe accident; (b) proliferation resistance; (c) cost competitiveness with alternative energy sources; (d) sustainable fuel supply; and (e) solution of the radioactive waste management problem

  13. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1995-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  14. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1994-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved in very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  15. Safety features of subcritical fluid fueled systems

    Energy Technology Data Exchange (ETDEWEB)

    Bell, C.R. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

  16. Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods

    International Nuclear Information System (INIS)

    Rothrock, R.B.

    1991-01-01

    Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs

  17. Monte Carlo Modeling Electronuclear Processes in Cascade Subcritical Reactor

    CERN Document Server

    Bznuni, S A; Zhamkochyan, V M; Polyanskii, A A; Sosnin, A N; Khudaverdian, A G

    2000-01-01

    Accelerator driven subcritical cascade reactor composed of the main thermal neutron reactor constructed analogous to the core of the VVER-1000 reactor and a booster-reactor, which is constructed similar to the core of the BN-350 fast breeder reactor, is taken as a model example. It is shown by means of Monte Carlo calculations that such system is a safe energy source (k_{eff}=0.94-0.98) and it is capable of transmuting produced radioactive wastes (neutron flux density in the thermal zone is PHI^{max} (r,z)=10^{14} n/(cm^{-2} s^{-1}), neutron flux in the fast zone is respectively equal PHI^{max} (r,z)=2.25 cdot 10^{15} n/(cm^{-2} s^{-1}) if the beam current of the proton accelerator is k_{eff}=0.98 and I=5.3 mA). Suggested configuration of the "cascade" reactor system essentially reduces the requirements on the proton accelerator current.

  18. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

    International Nuclear Information System (INIS)

    Carluccio, Thiago

    2011-01-01

    This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k eff and k src , and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

  19. Introduction of fusion driven subcritical system plasma design

    International Nuclear Information System (INIS)

    Bin Wu

    2003-01-01

    Fusion driven subcritical nuclear system (FDS) is a multifunctional hybrid reactor, which could breed nuclear fuel, transmute long-lived wastes, producing tritium and so on. This paper presents an introduction of FDS plasma design. Several different advance equilibrium configurations have been proposed and a 1.5-D discharge simulation of FDS was also present

  20. Study on uranium-water multiplicative means of the (RESUCO-Subcritical experimental reactor of uranium with oxygen) subcritical assembly by pulsed neutron technique

    International Nuclear Information System (INIS)

    Jesus Barbosa, S. de.

    1987-01-01

    The effective multiplication factor and the nuclear parameters associated with the variation of (RESUCO- Subcritical Experimental Reactor of Uranium with Oxygen) Subcritical Assembly Configuration, using pulsed neutron technique are analysed. BF3 detectors were used to detect the variation of thermal neutrons in the system, positioned parallelly to fuel elements, and a proton recoil detector was used for monitoring the neutron generation. (M.C.K.) [pt

  1. Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

    International Nuclear Information System (INIS)

    Romanello, V.; Salvatores, M.; Schwenk-Ferrero, A.; Gabrielli, F.; Maschek, W.; Vezzoni, B.

    2011-01-01

    Research highlights: → Double-strata fuel cycle has a potential to minimize transuranics mass in Europe. → European Minor Actinides legacy can be reduced down to 0 before the end of century. → 40% higher capacity needed to burn MA for fast critical reactor then for EFIT fleet. → Na cooled fast reactor cores with high content of MA and low CR have been assessed. → Fast critical and ADS-EFIT reactors show comparable MA transmutation performance. - Abstract: In the frame of Partitioning and Transmutation (P and T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a

  2. Reactor Dynamics Experiments with a Sub-Critical Assembly

    International Nuclear Information System (INIS)

    Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

    2004-01-01

    A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory

  3. Application of the Modified Source Multiplication (MSM) Technique to Subcritical Reactivity Worth Measurements in Thermal and Fast Reactor Systems

    International Nuclear Information System (INIS)

    Blaise, P.; Fougeras, Ph.; Mellier, F.

    2011-01-01

    The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. The ASM methodology uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. While this method works quite well for small reactivity variations, the raw results need to be corrected to take into account the flux perturbation at the fission chamber location. This is performed by applying to the measurement a correction factor called MSM. This paper describes in detail both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw ASM data. (authors)

  4. Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

  5. Inverse kinetics for subcritical systems with external neutron source

    International Nuclear Information System (INIS)

    Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

    2017-01-01

    Highlights: • It was developed formalism for reactivity calculation. • The importance function is related to the system subcriticality. • The importance function is also related with the value of the external source. • The equations were analyzed for seven different levels of sub criticality. • The results are physically consistent with others formalism discussed in the paper. - Abstract: Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication. Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of k eff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

  6. Design project of fast subcritical system 'Mala Lasta'; Idejno resenje brzog podkriticnog sistema Mala LASTA

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Stefanovic, D; Popovic, D; Pesic, M; Zavaljevski, N; Nikolic, D; Arsenovic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1988-10-15

    This report contains two parts. Part one covers the objective and fundamental elements for the choice of fast subcritical system 'Mala Lasta', review of the existing fast subcritical assemblies, and a description of the available domestic computer codes applied for calculating neutron reactor parameters. Comparison of results obtained by these codes for a number of existing subcritical assemblies was used for the choice of the design project described in part two of this report. It contains detailed description of the operating parameters of the chosen subcritical system based on the obtained calculated parameters.

  7. Application of the Modified Source Multiplication (MSM) technique to subcritical reactivity worth measurements in thermal and fast reactor systems

    International Nuclear Information System (INIS)

    Blaise, P.; Fougeras, P.; Mellier, F.

    2009-01-01

    The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. They have been successfully applied to absorber (single or clusters) worth measurement in both thermal and fast spectra, or for (sodium or water) void reactivity worths. The ASM methodology, which is the basic technique to estimate a reactivity worth, uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. If this method works quite well for small reactivity variation (a few effective delayed neutron fraction), its raw results needs to be corrected to take into account the flux perturbation in the fission chamber. This is performed by applying to the measurement a correction factor called MSM. Its characteristics is to take into account the local space and energy variation of the spectrum in the fission chamber, through standard perturbation theory applied to neutron transport calculation in the perturbed configuration. The proposed paper describes in details both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw data. (authors)

  8. Neutron chain length distributions in subcritical systems

    International Nuclear Information System (INIS)

    Nolen, S.D.; Spriggs, G.

    1999-01-01

    In this paper, the authors present the results of the chain-length distribution as a function of k in subcritical systems. These results were obtained from a point Monte Carlo code and a three-dimensional Monte Carlo code, MC++. Based on these results, they then attempt to explain why several of the common neutron noise techniques, such as the Rossi-α and Feynman's variance-to-mean techniques, are difficult to perform in highly subcritical systems using low-efficiency detectors

  9. 3D CAD model of the subcritical nuclear reactor of IPN

    International Nuclear Information System (INIS)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

    2016-09-01

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  10. Reactivity estimation for subcritical and critical reactors

    International Nuclear Information System (INIS)

    Benhaim A; Bellino P; Gomez A

    2012-01-01

    We developed a digital reactimeter that works in both current and pulse mode. This reactimeter will allow to estimate the reactivity of the reactor at any state. We st obtained for the measurements taken in the experimental reactor RA-1 the reactivity around the critical state without a neutron source. Measurements were made using simultaneously a compensated ionization chamber and a 3He proportional counter. The results were compared with the ones obtained from the digital reactimeter of reference with matching results within the experimental errors (author)

  11. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  12. Subcriticality calculation in nuclear reactors with external neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

    2007-07-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  13. Subcriticality calculation in nuclear reactors with external neutron sources

    International Nuclear Information System (INIS)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2007-01-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  14. A new formulation for the importance function in the kinetics of subcritical reactors

    International Nuclear Information System (INIS)

    Silva, Cristiano da; Senra Martinez, Aquilino; Carvalho da Silva, Fernando

    2012-01-01

    Highlights: ► In this paper we propose a new formulation for the importance function in the kinetics of subcritical systems. ► We analyze the relevance of an external neutron source for the subcritical interval 0.95 eff eff is the multiplication factor according to the physical properties of the nuclear reactor. For the purposes of validation of the proposed method we will use, as a reference method, the expansion in modes of the time-dependent neutron flux for the solution of the onedimensional diffusion equation. It will be presented results that demonstrate the precision of the proposed method when compared to the conventional point kinetic equations. The results show that the new point kinetic equations are rather precise in the subcriticality range considered.

  15. Sub-Critical Nuclear Reactor Based on FFAG-Accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee Seok; Kang, Hung Sik; Lee, Tae Yeon [Pohang Accelerator Laboratory, Pohang (Korea, Republic of)

    2011-10-15

    After the East-Japan earthquake and the subsequent nuclear disaster, the anti-nuclear mood has been wide spread. It is very unfortunate both for nuclear science community and for the future of mankind, which is threatened by two serious challenges, the global warming caused by the greenhouse effect and the shortage of energy cause by the petroleum exhaustion. While the nuclear energy seemed to be the only solution to these problems, it is clear that it has its own problems, one of which broke out so strikingly in Japan. There are also other problems such as the radiotoxic nuclear wastes that survive up to even tens of thousands years and the limited reserves of Uranium. To solve these problems of nuclear fission energy, accelerator-based sub-critical nuclear reactor was once proposed. (Its details will be explained below.) First of all, it is safe in a disaster such as an earthquake, because the deriving accelerator stops immediately by the earthquake. It also minimizes the nuclear waste problem by reducing the amount of the toxic waste and shortening their half lifetime to only a few hundred years. Finally, it solves the Uranium reserve problem because it can use Thorium as its fuel. The Thorium reserve is much larger than that of Uranium. Although the idea of the accelerator-driven nuclear reactor was proposed long time ago, it has not been utilized yet first by technical difficulty and economical reasons. The accelerator-based system needs 1 GeV, 10 MW power proton accelerator. A conventional linear accelerator would need several hundred m length, which is highly costly particularly in Korea because of the high land cost. However, recent technologies make it possible to realize that scale accelerator by a reasonable size. That is the fixed-field alternating gradient (FFAG) accelerator that is described in this article

  16. Sub-Critical Nuclear Reactor Based on FFAG-Accelerator

    International Nuclear Information System (INIS)

    Lee, Hee Seok; Kang, Hung Sik; Lee, Tae Yeon

    2011-01-01

    After the East-Japan earthquake and the subsequent nuclear disaster, the anti-nuclear mood has been wide spread. It is very unfortunate both for nuclear science community and for the future of mankind, which is threatened by two serious challenges, the global warming caused by the greenhouse effect and the shortage of energy cause by the petroleum exhaustion. While the nuclear energy seemed to be the only solution to these problems, it is clear that it has its own problems, one of which broke out so strikingly in Japan. There are also other problems such as the radiotoxic nuclear wastes that survive up to even tens of thousands years and the limited reserves of Uranium. To solve these problems of nuclear fission energy, accelerator-based sub-critical nuclear reactor was once proposed. (Its details will be explained below.) First of all, it is safe in a disaster such as an earthquake, because the deriving accelerator stops immediately by the earthquake. It also minimizes the nuclear waste problem by reducing the amount of the toxic waste and shortening their half lifetime to only a few hundred years. Finally, it solves the Uranium reserve problem because it can use Thorium as its fuel. The Thorium reserve is much larger than that of Uranium. Although the idea of the accelerator-driven nuclear reactor was proposed long time ago, it has not been utilized yet first by technical difficulty and economical reasons. The accelerator-based system needs 1 GeV, 10 MW power proton accelerator. A conventional linear accelerator would need several hundred m length, which is highly costly particularly in Korea because of the high land cost. However, recent technologies make it possible to realize that scale accelerator by a reasonable size. That is the fixed-field alternating gradient (FFAG) accelerator that is described in this article

  17. Submersion-Subcritical Safe Space (S4) reactor

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  18. Selection of initial events of accelerator driven subcritical system

    International Nuclear Information System (INIS)

    Wang Qianglong; Hu Liqin; Wang Jiaqun; Li Yazhou; Yang Zhiyi

    2013-01-01

    The Probabilistic Safety Assessment (PSA) is an important tool in reactor safety analysis and a significant reference to the design and operation of reactor. It is the origin and foundation of the PSA for a reactor to select the initial events. Accelerator Driven Subcritical System (ADS) has advanced design characteristics, complicated subsystems and little engineering and operating experience, which makes it much more difficult to identify the initial events of ADS. Based on the current design project of ADS, the system's safety characteristics and special issues were analyzed in this article. After a series of deductions with Master Logic Diagram (MLD) and considering the relating experience of other advanced research reactors, a preliminary initial events was listed finally, which provided the foundation for the next safety assessment. (authors)

  19. Simulation an Accelerator driven Subcritical Reactor core with thorium fuel

    International Nuclear Information System (INIS)

    Shirmohammadi, L.; Pazirandeh, A.

    2011-01-01

    The main purpose of this work is simulation An Accelerator driven Subcritical core with Thorium as a new generation nuclear fuel. In this design core , A subcritical core coupled to an accelerator with proton beam (E p =1 GeV) is simulated by MCNPX code .Although the main purpose of ADS systems are transmutation and use MA (Minor Actinides) as a nuclear fuel but another use of these systems are use thorium fuel. This simulated core has two fuel assembly type : (Th-U) and (U-Pu) . Consequence , Neutronic parameters related to ADS core are calculated. It has shown that Thorium fuel is use able in this core and less nuclear waste ,Although Iran has not Thorium reserves but study on Thorium fuel cycle can open a new horizontal in use nuclear energy as a clean energy and without nuclear waste

  20. Reactor parameters and constants determination by using measurements in subcritical and exponential assembly

    International Nuclear Information System (INIS)

    Voi, Dante Luiz; Santos Bastos, Wilma dos

    1995-01-01

    Subcritical and exponential experiments are important for Reactor Physics integral parameter determinations both to validate and confirm theoretical models for reactor calculations. An exponential and subcritical facility has been constructed to be used on the internal thermal column of the Argonauta reactor at IEN-CNEN- Rio de Janeiro. An experimental research program has been developed for the determination of fundamental reactor constants as buckling, migration areas, resonance escape probabilities, thermal utilization, fast fission and fuel eta factors. (author) 23 refs

  1. The safe, economical operation of a slightly subcritical reactor and transmutor with a small proton accelerator

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi

    1994-01-01

    I suggest that an accelerator can be used to increase the safety and neutron economy of a power reactor and a transmutor of long-lived radioactive wastes, such as minor actinides and fission products, by providing neutrons for its subcritical operation. Instead of the large subcriticality k=0.9-0.95 which we originally proposed for such transmutor, we propose to use a slightly subcritical reactor, such as k=0.99, which will avoid many of the technical difficulties that are associated with large subcriticality, such as localized power peaking, radiation damage due to injection of medium-energy protons, the high current accelerator, and the requirement for a long beam-expansion section. We analyze the power drop that occurred in Phenix reactor, and show that the operating this reactor in subcritical conditions improves safety. (author). 13 refs., 5 figs

  2. The safe, economical operation of a slightly subcritical reactor and transmutor with a small proton accelerator

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi.

    1994-01-01

    This report describes methods in which an accelerator can be used to increase the safety and neutron economy of a power reactor and transmutor of long-lived radioactive wastes, such as minor actinides and fission products, by providing neutrons for its subcritical operation. Instead of the rather large subcriticality of k=0.9--0.95 which we originally proposed for such a transmutor, we propose to use a slightly subcritical reactor, such as k=0.99, which will avoid many of the technical difficulties that are associated with large subcriticality, such as localized power peaking, radiation damage due to the injection of medium-energy protons, the high current accelerator, and the requirement for a long beam-expansion section. We analyzed the power drop that occurred in Phoenix reactor, and show that the operating this reactor in subcritical condition improves its safety

  3. Feasibility of waste transmutation using accelerator-driven IRIS subcritical system

    International Nuclear Information System (INIS)

    Petroviae, B.; Carelli, M.; Paramonov, D.

    2001-01-01

    Waste transmutation is considered for reducing radio-toxicity of nuclear waste generated in power reactors. Accelerator driven subcritical systems (ADS) offer certain advantages over the use of nuclear reactors. Transmutation of fission products (e.g. 99 Tc) generally requires thermal neutron spectrum, while for actinides fast spectrum provides better performance. Proposed solutions to this problem include a multi-strata approach as well as a multi-zone (thermal/fast-spectrum) single systems. In this paper we examine the feasibility of employing a dual-spectrum two-zone accelerator-driven IRIS subcritical for waste transmutation. (author)

  4. High power ring methods and accelerator driven subcritical reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Tahar, Malek Haj [Univ. of Grenoble (France)

    2016-08-07

    High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g., PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the

  5. Safety and control of accelerator-driven subcritical systems

    Energy Technology Data Exchange (ETDEWEB)

    Rief, H. [Ispra Establishment (Italy); Takahashi, H. [Brookhaven National Laboratory, Long Island, NY (United States)

    1995-10-01

    To study control and safety of accelertor driven nuclear systems, a one point kinetic model was developed and programed. It deals with fast transients as a function of reactivity insertion. Doppler feedback, and the intensity of an external neutron source. The model allows for a simultaneous calculation of an equivalent critical reactor. It was validated by a comparison with a benchmark specified by the Nuclear Energy Agency Committee of Reactor Physics. Additional features are the possibility of inserting a linear or quadratic time dependent reactivity ramp which may account for gravity induced accidents like earthquakes, the possibility to shut down the external neutron source by an exponential decay law of the form exp({minus}t/{tau}), and a graphical display of the power and reactivity changes. The calculations revealed that such boosters behave quite benignly even if they are only slightly subcritical.

  6. Measurement of subcriticality by a pulsing α-method

    International Nuclear Information System (INIS)

    Jitarev, V.E.; Kachanov, V.M.; Kuzmin, A.N.

    1999-01-01

    The report presents results of a pulsing α-method [1] for determination of the WWER system subcriticality. The pulsing α-method permits to conduct measurements of system subcriticality in conditions of subcritical state and large neutron background. Therefore this method can be used for the control of a subcriticality of storehouses of a burn up nuclear fuel and stopped reactor. (Authors)

  7. Subcritical molten salt reactor with fast/intermediate spectrum for minor actinides transmutation

    International Nuclear Information System (INIS)

    Degtyarev, Alexey M.; Feinberg, Olga S.; Kolyaskin, Oleg E.; Myasnikov, Andrey A.; Karmanov, Fedor I.; Kuznetsov, Andrey Yu.; Ponomarev, Leonid I.; Seregin, Mikhail B.; Sidorkin, Stanislav F.

    2011-01-01

    The subcritical molten-salt reactor for transmutation of Am and Cm with the fast-intermediate neutron spectrum is suggested. It is shown that ∼10 such reactor-burners is enough to support the future nuclear power based on the fast reactors as well as for the transmutation of Am and Cm accumulated in the spent fuel storages. (author)

  8. Subcriticality of accelerator driven system by AESJ/JAERI working party

    International Nuclear Information System (INIS)

    Iwasaki, Tomohiko

    2002-01-01

    Under Atomic Energy Society of Japan (AESJ) and Japan Atomic Energy Research Institute (JAERI), a Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) has been set since March 1999 to review and investigate special subjects related to reactor physics research of Accelerator-Driven System (ADS). In the ADS-WP, the extensive and aggressive activity is being made by 25 professional members in the field of reactor physics in Japan. The ADS is now studying three subjects related to subcriticality of ADS; (1) calculation accuracy of subcriticality on ADS, (2) critical safety issues of ADS, and (3) theoretical review of subcriticality and its measurement methods. This paper describes two topics related to the subjects (1) and (2); one is an analysis of maximum reactivity potentially inserted to a subcritical core and the other is a benchmark proposal for checking calculation accuracy of subcriticality on ADS. The full specification of the calculation benchmark will be supplied by June 2002. Researchers from overseas, especially from Korea, are welcome to join this benchmark

  9. Conceptual design based on scale laws and algorithms for sub-critical transmutation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    In order to conduct the effective integration of computer-aided conceptual design for integrated nuclear power reactor, not only is a smooth information flow required, but also decision making for both conceptual design and construction process design must be synthesized. In addition to the aboves, the relations between the one step and another step and the methodologies to optimize the decision variables are verified, in this paper especially, that is, scaling laws and scaling criteria. In the respect with the running of the system, the integrated optimization process is proposed in which decisions concerning both conceptual design are simultaneously made. According to the proposed reactor types and power levels, an integrated optimization problems are formulated. This optimization is expressed as a multi-objective optimization problem. The algorithm for solving the problem is also presented. The proposed method is applied to designing a integrated sub-critical reactors. 6 refs., 5 figs., 1 tab. (Author)

  10. Conceptual design based on scale laws and algorithms for sub-critical transmutation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    In order to conduct the effective integration of computer-aided conceptual design for integrated nuclear power reactor, not only is a smooth information flow required, but also decision making for both conceptual design and construction process design must be synthesized. In addition to the aboves, the relations between the one step and another step and the methodologies to optimize the decision variables are verified, in this paper especially, that is, scaling laws and scaling criteria. In the respect with the running of the system, the integrated optimization process is proposed in which decisions concerning both conceptual design are simultaneously made. According to the proposed reactor types and power levels, an integrated optimization problems are formulated. This optimization is expressed as a multi-objective optimization problem. The algorithm for solving the problem is also presented. The proposed method is applied to designing a integrated sub-critical reactors. 6 refs., 5 figs., 1 tab. (Author)

  11. 233U breeding in accelerator-driven sub-critical fast reactor

    International Nuclear Information System (INIS)

    Yang Yongwei; An Yu

    1999-01-01

    Accelerator-driven Sub-critical Fast Reactor (ADFR) is chosen as fissile-material-breeding reactor. (U-Pu)O x is chosen as fuel in the core and ThO 2 as fertile material in the blanket zone to breed 233 U. Molten lead is chosen as coolant because of its better neutronic and chemical characteristics over sodium. The program system used for neutronics study consists of: LAHET, for the simulation of the interaction between the proton with medium energy and the nuclei of the target; MCNP4A, for the simulation of neutron transport with energy below 20 MeV in the sub-critical reactor; CONNECT1, for the processing of some tallies provided by the output of MCNP4A in order to prepare micro-cross sections for elements used for burnup calculation; ORIGEN2, used for multi-region burnup calculation; CONNECT2, for the processing of atom densities of some elements provided in the output of ORIGEN2 in order to prepare input to LAHET calculation for next time step. The calculated results show that the proposed case is feasible for breeding fissile material considering the criticality safety, power density, burnup, etc

  12. Accelerator Driven Sub-Critical System for the Radioactive Waste Transmutation

    International Nuclear Information System (INIS)

    Avramovic, I.; Pesic, M.

    2008-01-01

    Spent nuclear fuel discharged from nuclear power plants is the main problem during design of radioactive waste disposal. Most of the hazard stems from only a few chemical elements. The radiotoxicity of these elements can be efficiently reduced using partitioning and transmutation in fast reactors and accelerator driven subcritical systems. (author)

  13. Comparison of the transient behavior of lead-based advanced critical and sub-critical reactors

    International Nuclear Information System (INIS)

    Wang Gang; Gu Zhixing; Wang Zhen; Jin Ming; Bai Yunqing

    2014-01-01

    A lead-based reactor developed by FDS Team is proposed in 2011 and designed to be 10 MW. It is a pool type reactor and the primary coolant is driven by natural circulation. The reactor has two operation modes, which are a lead-based critical fast reactor mode and a lead-based sub-critical reactor mode. The conceptual designs of the two modes are both completed by 2013. In this paper, four transient accidents were simulated for both the critical and sub-critical reactors above by NTC-2D code, which is developed by FDS Team for advanced reactor safety analysis. The four accidents were protected and unprotected loss of heat sink accidents (PLOHS and ULOHS), protected and unprotected transient overpower accidents (PTOP and UTOP). The simulation results of the two reactors were compared and analyzed. The results showed that during PLOHS and PTOP accidents for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the sub-critical reactor increased bigger than those of the critical one. For UTOP, the parameters above of the critical fast reactor were much bigger than those of the sub-critical one. The analysis results showed different safety advantages of the lead-based critical fast and sub-critical reactors during different transient accidents. (author)

  14. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    Accelerator driven systems (ADS) are attracting worldwide attention .... The region of interest (or the entire reactor core) is divided into a suitable number ..... have also presented the status of the theoretical and experimental activities being.

  15. Nuclear data requirements for accelerator driven sub-critical systems

    Indian Academy of Sciences (India)

    The development of accelerator driven sub-critical systems (ADSS) require significant amount of new nuclear data in extended energy regions as well as for a variety of new materials. This paper reviews these perspectives in the Indian context.

  16. Power reactors and sub-critical blanket systems with lead and lead-bismuth as coolant and/or target material. Utilization and transmutation of actinides and long lived fission products

    International Nuclear Information System (INIS)

    2003-05-01

    High level radioactive waste disposal is an issue of great importance in the discussion of the sustainability of nuclear power generation. The main contributors to the high radioactivity are the fission products and the minor actinides. The long lived fission products and minor actinides set severe demands on the arrangements for safe waste disposal. Fast reactors and accelerator driven systems (ADS) are under development in Member States to reduce the long term hazard of spent fuel and radioactive waste, taking advantage of their incineration and transmutation capability. Important R and D programmes are being undertaken in many Member States to substantiate this option and advance the basic knowledge in this innovative area of nuclear energy development. The conceptual design of the lead cooled fast reactor concept BREST-OD-300, as well as various other conceptual designs of lead/lead-bismuth cooled fast reactors have been developed to meet enhanced safety and non-proliferation requirements, aiming at both energy production and transmutation of nuclear waste. Some R and D studies indicate that the use of lead and lead-bismuth coolant has some advantages in comparison with existing sodium cooled fast reactor systems, e.g.: simplified design of fast reactor core and BOP, enhanced inherent safety, and easier radwaste management in related fuel cycles. Moreover, various ADS conceptual designs with lead and lead-bismuth as target material and coolant also have been pursued. The results to date are encouraging, indicating that the ADS has the potential to offer an option for meeting the challenges of the back end fuel cycle. During the last decade, there have been substantial advances in several countries with their own R and D programme in the fields of lead/lead-bismuth cooled critical and sub-critical concepts. coolant technology, and experimental validation. In this context, international exchange of information and experience, as well as international

  17. Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

    International Nuclear Information System (INIS)

    Orso, J; Marenzana, A

    2012-01-01

    Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

  18. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  19. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-01-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

  20. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  1. Operation and reactivity measurements of an accelerator driven subcritical TRIGA reactor

    Science.gov (United States)

    O'Kelly, David Sean

    Experiments were performed at the Nuclear Engineering Teaching Laboratory (NETL) in 2005 and 2006 in which a 20 MeV linear electron accelerator operating as a photoneutron source was coupled to the TRIGA (Training, Research, Isotope production, General Atomics) Mark II research reactor at the University of Texas at Austin (UT) to simulate the operation and characteristics of a full-scale accelerator driven subcritical system (ADSS). The experimental program provided a relatively low-cost substitute for the higher power and complexity of internationally proposed systems utilizing proton accelerators and spallation neutron sources for an advanced ADSS that may be used for the burning of high-level radioactive waste. Various instrumentation methods that permitted ADSS neutron flux monitoring in high gamma radiation fields were successfully explored and the data was used to evaluate the Stochastic Pulsed Feynman method for reactivity monitoring.

  2. Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, Pavel N.; Balanin, Andrey L.; Dudnikov, Anatoly A.; Fomichenko, Petr A.; Nevinitsa, Vladimir A.; Frolov, Aleksey A.; Lubina, Anna S.; Sedov, Aleksey A.; Subbotin, Aleksey S.; Blandinsky, Viktor Yu. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for {sup 233}U conversion from {sup 232}Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

  3. A new approach to make collapsed cross section for burnup calculation of subcritical system

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Kondo, Keitaro; Miyamaru, Hiroyuki; Murata, Isao

    2008-01-01

    A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)

  4. Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation

    International Nuclear Information System (INIS)

    Pesic, Milan; Neskovic, Nebojsa

    2006-01-01

    Study of a small accelerator-driven subcritical research reactor in the Vinca Institute of Nuclear Sciences was initiated in 1999. The idea was to extract a beam of medium-energy protons or deuterons from the TESLA accelerator installation, and to transport and inject it into the reactor. The reactor core was to be composed of the highly enriched uranium fuel elements. The reactor was designated as ADSRR-H. Since the use of this type of fuel elements was not recommended any more, the study of a small accelerator-driven subcritical research reactor employing the low-enriched uranium fuel elements began in 2004. The reactor was designated as ADSRR-L. We compare here the results of the initial computer simulations of ADSRR-H and ADSRR-L. The results have confirmed that our concept could be the basis for designing and construction of a low neutron flux model of the proposed accelerator-driven subcritical power reactor to be moderated and cooled by lead. Our objective is to study the physics and technologies necessary to design and construct ADSRR-L. The reactor would be used for development of nuclear techniques and technologies, and for basic and applied research in neutron physics, metrology, radiation protection and radiobiology

  5. Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

    International Nuclear Information System (INIS)

    Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

    2012-01-01

    The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

  6. Turn-key SRF accelerators to drive subcritical reactors

    International Nuclear Information System (INIS)

    Johnson, Rolland P.

    2011-01-01

    Large particle accelerator projects, both accomplished and proposed, have been used to engage US industry through contracts and grants to develop efficient capabilities to design, develop, produce, and deliver entire accelerator systems or any needed subsystems. Staffed in many cases by experienced scientists and engineers from National Laboratories and Universities, existing companies could extend their portfolios to offer turn-key accelerators with parameters to match the needs of ADS. If the reactors were based on molten salt fuel such that trip rate requirements were relaxed, the developments needed for a multi-MW proton accelerator for ADS would be minimal. Turn-key SRF proton linacs for ADS operation can be ordered now to enable GW-level power generation from natural thorium, natural uranium, or nuclear waste from conventional reactors. (author)

  7. Neutronic Design of an Accelerator Driven Sub-Critical Research Reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2002-01-01

    Conceptual design of an accelerator driven sub-critical research reactor (ADSRR), as a new project in the Vinca Institute of Nuclear Sciences, is suggested for support to the Ministry of science, technologies and development of Republic Serbia, Yugoslavia. This paper show initial results of neutronic analyses of the proposed ADSRR carried out by Monte Carlo based MCNP and SHIELD codes. According to the proposal, the ADSRR would be constructed, in a later phase, at high-energy channel H5B of the VINCY cyclotron of the TESLA Accelerator Installation, that is under completion in the Vinca Institute. The fuel elements of 80%-enriched uranium dioxide dispersed in aluminium matrix, available in the Vinca Institute, are proposed for the ADSRR core design. The HEU fuel elements are placed in aluminium tubes filled by the 'primary moderator' - light water. These 'fuel tubes' are placed in a square lattice within lead matrix in a stainless steel tank. The lead is used as a 'secondary moderator' in the core and as the axial and radial reflector. Such design of the ADSRR shows that this small low neutron flux system can be used as an experimental 'demonstration' ADS with some neutron characteristics similar to proposed well-known lead moderated and cooled power sub-critical ADS with intermediate or fast neutron spectrum. The proposed experimental ADSRR, beside usage as a valuable research machine in reactor and neutron physics, will contribute to following and developing new nuclear technologies in the country, useful for eventual nuclear power option and nuclear waste incineration in future. (author)

  8. Conceptual design for accelerator-driven sodium-cooled sub-critical transmutation reactors using scale laws

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates the reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scaled and verified through the methodology in this paper, which is referred to Advanced Liquid Metal Reactor (ALMR). A Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the integrated code system. Integrated Code System (ICS) consists of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related regions are analyzed once-through Results of conceptual design are attached in this paper. 5 refs., 4 figs., 1 tab. (Author)

  9. Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

    International Nuclear Information System (INIS)

    Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

    1997-01-01

    Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

  10. Analysis of an accelerator-driven subcritical light water reactor

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

    1997-01-01

    An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

  11. Study on the method of determining the sub-criticality of a reactor via the measurement of core neutron flux spatial distribution

    International Nuclear Information System (INIS)

    Ma Aifeng; Jiang Xiaofeng; Zhang Shaohong

    2007-01-01

    A new methodology based on rigorous reactor physics theory astead of the point reactor assumption was proposed to determine or monitor the sub-criticality ora reactor, especially the sub-critical reactor of ADS, via the measurement of in-core flux spatial distribution. Preliminary numerical studies on the 1st ADS sub-critical experimental facilities-Venus No.1 in China have demonstrated the feasibility of this new method. Related discussions pointed out the potential applications of the method. (authors)

  12. Comparison between two gas-cooled TRU burner subcritical reactors: fusion-fission and ADS

    International Nuclear Information System (INIS)

    Carluccio, T.; Rossi, P.C.R.; Angelo, G.; Maiorino, J.R.

    2011-01-01

    This work shows a preliminary comparative study between two gas cooled subcritical fast reactor as dedicated transuranics (TRU) transmuters: using a spallation neutron source or a D-T fusion neutron source based on ITER. The two concepts are compared in terms of a minor actinides burning performance. Further investigations are required to choose the best partition and transmutation strategy. Mainly due to geometric factors, the ADS shows better neutron multiplication. Other designs, like SABR and lead cooled ADS may show better performances than a Gas Coolead Subcritical Fast Reactors and should be investigated. We noticed that both designs can be utilized to transmutation. Besides the diverse source neutron spectra, we may notice that the geometric design and cycle parameters play a more important role. (author)

  13. Evaluation of subcritical hybrid systems loaded with reprocessed fuel

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

    2015-01-01

    Highlights: • Accelerator driven systems (ADS) and fusion–fission systems are investigated for transmutation and fuel regeneration. • The calculations were performed using Monteburns code. • The results indicate the most suitable system for achieve transmutation. - Abstract: Two subcritical hybrid systems containing spent fuel reprocessed by Ganex technique and spiked with thorium were submitted to neutron irradiation of two different sources: ADS (Accelerator-driven subcritical) and Fusion. The aim is to investigate the nuclear fuel evolution using reprocessed fuel and the neutronic parameters under neutron irradiation. The source multiplication factor and fuel depletion for both systems were analysed during 10 years. The simulations were performed using MONTEBURNS code (MCNP/ORIGEN). The results indicate the main differences when irradiating the fuel with different neutron sources as well as the most suitable system for achieving transmutation

  14. Optimization by simulation of the coupling between a sub-critical reactor and its spallation source. Towards a pilot reactor

    International Nuclear Information System (INIS)

    Kerdraon, D.

    2001-10-01

    Accelerator Driven Systems (ADS), based on a proton accelerator and a sub-critical core coupled with a spallation target, offer advantages in order to reduce the nuclear waste radiotoxicity before repository closure. Many studies carried out on the ADS should lead to the definition of an experimental plan which would federate the different works in progress. This thesis deals with the neutronic Monte Carlo simulations with the MCNPX code to optimize such a system in view of a pilot reactor building. First, we have recalled the main neutronic properties of an hybrid reactor. The concept of gas-cooled eXperimental Accelerator Driven System (XADS) chosen for our investigations comes from the preliminary studies done by the Framatome company. In order to transmute minor actinides, we have considered the time evolution of the main fuels which could be reasonably used for the demonstration phases. The neutronic parameters of the reactor, concerning minor actinide transmutation, are reported. Also, we have calculated the characteristic times and the transmutation rates in the case of 99 Tc and 129 I isotopes. We have identified some neutronic differences between an experimental and a power ADS according to the infinite multiplication coefficient, the shape factor and the level of flux to extend the demonstrator concept. We have proposed geometric solutions to keep the radial shape factor of a power ADS acceptable. In the last part, beyond the experimental XADS scope, we have examined the possible transition towards an uranium/thorium cycle based on Molten Salt Reactors using a power ADS in order to generate the required 233 U proportion. (author)

  15. Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

    2015-10-15

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

  16. The measurement of subcritical reactivity in nuclear reactors by use of a high frequency sine-wave modulated neutron source

    International Nuclear Information System (INIS)

    Guppy, C.B.

    1964-11-01

    In this report the frequency response characteristics for phase and gain of the fundamental reactor mode of the zero power kinetics are given for various subcritical reactivities in a fast reactor and in a thermal reactor. Results, of a study on harmonic effects based on a small zero energy thermal reactor are presented which demonstrate the importance of spatial harmonic effects. A harmonic theory for thermal reactors is developed. A new method of measuring, subcritical reactivity at moderately high frequencies is suggested which circumvents the harmonic problem. It is shown that at high frequencies there is more sensitivity than at low frequencies and that this could lead to an increased range over which subcritical reactivity can be measured. (author)

  17. Fast accelerator driven subcritical system for energy production: nuclear fuel evolution

    International Nuclear Information System (INIS)

    Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A.F.; Costa, Antonella L.

    2011-01-01

    Accelerators Driven Systems (ADS) are an innovative type of nuclear system, which is useful for long-lived fission product transmutation and fuel regeneration. The ADS consist of a coupling of a sub-critical nuclear core reactor and a proton beam produced by a particle accelerator. These particles are injected into a target for the neutrons production by spallation reactions. The neutrons are then used to maintain the fission chain in the sub-critical core. The aim of this study is to investigate the nuclear fuel evolution of a lead cooled accelerator driven system used for energy production. The fuel studied is a mixture based upon "2"3"2Th and "2"3"3U. Since thorium is an abundant fertile material, there is hope for the thorium-cycle fuels for an accelerator driven sub-critical system. The target is a lead spallation target and the core is filled with a hexagonal lattice. High energy neutrons are used to reduce the negative reactivity caused by the presence of protoactinium, since this effect is most pronounced in the thermal range of the neutron spectrum. For that reason, such material is not added moderator to the system. In this work is used the Monte Carlo code MCNPX 2.6.0, that presents the the depletion/ burnup capability. The k_e_f_f evolution, the neutron energy spectrum in the core and the nuclear fuel evolution using ADS source (SDEF) and kcode-mode are evaluated during the burnup. (author)

  18. Sensitivity analysis of source driven subcritical systems by the HGPT methodology

    International Nuclear Information System (INIS)

    Gandini, A.

    1997-01-01

    The heuristically based generalized perturbation theory (HGPT) methodology has been extensively used in the last decades for analysis studies in the nuclear reactor field. Its use leads to fundamental reciprocity relationships from which perturbation, or sensitivity expressions can be derived, to first and higher order, in terms of simple integration operation of quantities calculated at unperturbed system conditions. Its application to subcritical, source-driven systems, now considered with increasing interest in many laboratories for their potential use as nuclear waste burners and/or safer energy producers, is here commented, with particular emphasis to problems implying an intensive system control variable. (author)

  19. Efficiency of an LBE spallation target in an accelerator-driven molten salt subcritical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Sang-In [Sungkyunkwan University, Suwon (Korea, Republic of); Hong, Seung-Woo [Sungkyunkwan University, Suwon (Korea, Republic of); Kadi, Yacine [CERN, Geneva (Switzerland)

    2016-10-15

    An Accelerator-Driven System (ADS) combined with a subcritical Molten Salt Reactor (MSR) is a type of hybrid reactor originally designed to breed uranium from thorium or to incinerate long-lived minor actinides in nuclear wastes. In an MSR, the salt material is used not only as a nuclear fuel but also as a primary coolant. In addition, this material is used as a target for inducing spallation neutrons in most AD-MSR concepts. A high energy proton beam impinges on a heavy metal target to induce spallation reactions and produces neutrons. Accordingly, a reliable proton accelerator is needed to feed the source neutrons. As ADSs have been criticized for requiring high power accelerators, minimization of beam power is an important aspect of ADS design. A primary concern associated with ADS development is stable high-power accelerators. We therefore studied the neutron source efficiencies of an AD-MSR involving chloride fuels by including a Pb-Bi eutectic (LBE) spallation target. The proton source efficiency and the accelerator beam power required have been studied for an AD-MSR. Adoption of an LBE spallation target induces an increase in proton source efficiencies in comparison to the case without a spallation target. Thus the presence of an efficient spallation target is useful in the reduction of the beam power of an accelerator. Almost 33 % of the beam power can be reduced in comparison to the case without the target for NaCl-Th/{sup 233}U fuel, and about 16 % for NaCl-U/TRU fuel. The beam power amplifications increase by 1.5 times for NaCl-Th/{sup 233}U and 1.2 times for NaCl-U/TRU in comparison with the no target AD-MSR.

  20. Neutronics of a sub-critical system burning non-recycled LWR waste

    International Nuclear Information System (INIS)

    Wallenius, J.; Tucek, K.; Gudowski, W.; Sanders, C.

    1999-01-01

    We have investigated neutronic properties of a subcritical system designed for transmutation of non-recycled TRU discharges from commercial light water reactors. Burnable absorbers (BA) and depleted uranium in variable concentrations are introduced in order to maximize fission to absorption ratios, and to minimize power peaking as well as reactivity losses. The use of nitride fuel raises linear power ratings to 60-110 kW/m while keeping fuel center line temperatures below 1400 K after gap closure. A comparatively small power peaking of 1.5 at a subcriticality level of 0.97 allows for a total core power of 1200 MWth with a corresponding proton beam power of 20 MW at BOL. Core averaged fission to absorption ratios for Np and Am as high as 0.5 are achieved using 10 B enriched B 4 C as BA. Hence, both Pu and minor actinide inventories are reduced during burnup in the here proposed system, mitigating swelling problems arising due to high-activity in MA-based fuels. Disadvantages following BA introduction, such as increase of void coefficients and decrease of negative doppler feedback in conjunction with small values of β eff , are addressed by setting the BOL subcriticality level to 0.97. (author)

  1. Consultancy on the potential of fusion/fission sub-critical neutron systems for energy production and transmutation. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    The Workshop on Sub-critical Neutron Production held at the University of Maryland and the Eisenhower Institute on 11-13 October 2004 brought together members of fusion, fission and accelerator technical communities to discuss issues of spent fuel, nonproliferation, reactor safety and the use of neutrons for sub-critical operation of nuclear reactors. The Workshop strongly recommended that the fusion community work closely with other technical communities to ensure that a wider range of technical solutions is available to solve the spent fuel problem and to utilize the current actinide inventories. Participants of the Workshop recommended that a follow-on Workshop, possibly under the aegis of the IAEA, should be held in the first half of the year 2005. The Consultancy Meeting is the response to this recommendation. The objectives of the Consultancy meeting were to hold discussions on the role of fusion/fission systems in sub-critical operations of nuclear reactors. The participants agreed that development of innovative (fourth generation) fission reactors, advanced fuel cycle options, and disposition of existing spent nuclear fuel inventories in various Member Sates can significantly benefit from including sub-critical systems, which are driven by external neutron sources. Spallation neutrons produced by accelerators have been accepted in the past as the means of driving sub-critical reactors. The accelerator community deserves credit in pioneering this novel approach to reactor design. Progress in the design and operation of fusion devices now offers additional innovative means, broadening the range of sub-critical operations of fission reactors. Participants felt that fusion should participate with accelerators in providing a range of technical options in reactor design. Participants discussed concrete steps to set up a small fusion/fission system to demonstrate actinide burning in the laboratory and what advice should be given to the Agency on its role in

  2. 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)

  3. Spatial and spectral effects in subcritical system pulsed experiments

    International Nuclear Information System (INIS)

    Dulla, S.; Nervo, M.; Ravetto, P.; Carta, M.

    2013-01-01

    Accurate neutronic models are needed for the interpretation of pulsed experiments in subcritical systems. In this work, the extent of spatial and spectral effects in the pulse propagation phenomena is investigated and the analysis is applied to the GUINEVERE experiment. The multigroup cross section data is generated by the Monte Carlo SERPENT code and the neutronic evolution following the source pulse is simulated by a kinetic diffusion code. The results presented show that important spatial and spectral aspects need to be properly accounted for and that a detailed energy approach may be needed to adequately capture the physical features of the system to the pulse injection. (authors)

  4. Neutronic calculations for a subcritical system with external source

    International Nuclear Information System (INIS)

    Cintas, A; Lopasso, E.M; Marquez Damian, J. I

    2006-01-01

    We present a neutronic study on an A D S, systems capable of transmute minor actinides and fission products in order to reduce their radiotoxicity and mean-life.We compare neutronic parameters obtained with Scale/Tort and M C N P modelling a sub-critical system with source from a N E A Benchmark.Due to lack of nuclear data at the temperature of the system, we perform calculations at available temperature of libraries (300 K); to compensate the reactivity insertion due to the temperature change we reduce the size of the fuel zone in order to get a sub-critical system that allow u s to evaluate neutronic parameters of the system with source.We have found that the numerical results (neutron spectrum, neutron flux distributions and other neutronic parameters) are in agreement with the M C N P and with those of the benchmark participants even though the geometric models used are not exactly the same. We conclude that with the real temperature cross sections, the calculation scheme developed (Scale/Tort and M C N P) will give reliable results in A D S evaluations [es

  5. Obtainment of the subcritical reactivity by mean of measurement of APSD and CPSD employing pulse mode detectors in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Lee, Seung Min

    2014-01-01

    This work presents a new experimental approach to determine the reactivity levels of subcritical systems. The method employs the subcritical kinetic model developed by Gandini and Salvatores and it is based only on measured quantities such as counting rates of the detectors employed in the experiments and the parameters arising from the least squares fitting of the APSD (Auto Power Spectral Density) and CPSD (Cross Power Spectral Density). Detector efficiencies, quantity required in other procedures such as Neutron Source Multiplication (NSM) method, are not needed in the proposed method. The only hypothesis made in the method was the independence of the effective delayed neutron fraction and the prompt neutron generation time to the subcriticality level of the system. The proposed method was applied to measure the reactivity of several subcritical configurations of the IPEN/MB-01 reactor. Measurements of APSD and CPSD were performed in several degrees of subcriticality (up to around -7000 pcm). The spectral densities data were least squares fitted to get the prompt decay mode (α) and other quantities. Beside the startup source of the facility, an external neutron source of Am-Be was installed near the core in order to improve neutron counting statistics. The final experimental results are of good quality. The proposed experimental method shows clearly that the classical point kinetic theory cannot describe the measured reactivity. Instead, the reactivity inferred from this model follows closely the subcriticality index (ζ) for the source arrangements in the experiment. The agreement of the MCNP5 and GPT-TORT results, both with ENDF/B-VII.0 as the basic nuclear data library, when compared to the corresponding experimental ones was also good. (author)

  6. Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

    Energy Technology Data Exchange (ETDEWEB)

    He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)

    2015-12-15

    Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

  7. The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N. [National Research Centre Kurchatov Institute (Russian Federation); Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu., E-mail: yuri.titarenko@itep.ru [Institute for Theoretical and Experimental Physics (Russian Federation)

    2015-12-15

    A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

  8. Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yavuz, U. [Turkish Atomic Energy Authority, Ankara (Turkey). Nuclear Safety Dept.; Zabunoolu, O.H. [Hacettepe Univ., Ankara (Turkey). Dept. of Nuclear Engineering

    2006-08-15

    Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33 000, 45 000, and 55 000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards. (orig.)

  9. Source term determination from subcritical multiplication measurements at Koral-1 reactor

    International Nuclear Information System (INIS)

    Blazquez, J.B.; Barrado, J.M.

    1978-01-01

    By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

  10. Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium

    International Nuclear Information System (INIS)

    Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R.; Sajo B, L.

    2015-10-01

    This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of 252 Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of 252 Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ( 7 LiF - BeF 2 - ThF 4 - UF 4 ) as fuel. For design the k eff , neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of 233 U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k eff of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)

  11. The trade experiment: Shielding calculations for the building hosting the subcritical system

    International Nuclear Information System (INIS)

    Burn, K. W.; Carta, M.; Casalini, L.; Kadi, Y.; Monti, S.; Nava, E.; Palomba, M.; Petrovich, C.; Picardi, L.; Rubbia, C.; Troiani, F.

    2005-01-01

    The TRADE project (TRIGA Accelerator Driven Experiment), to be performed at the existing TRIGA reactor at ENEA Casaccia, has been proposed as a validation of the accelerator-driven system (ADS) concept. TRADE will be the first experiment in which the three main components of an ADS-the accelerator, spallation target and sub-critical blanket-are coupled at a power level sufficient to encounter reactivity feedback effects. As such, TRADE represents the necessary intermediate step in the development of hybrid transmutation systems, its expected outcomes being considered crucial-in terms of proof of stability of operation, dynamic behaviour and licensing issues-for the subsequent realisation of an ADS Transmutation Demonstrator. An essential role in the feasibility study of the experiment is played by radioprotection calculations. Such a system exhibits new characteristics with respect to a traditional reactor, owing to the presence of the proton accelerator. As beam losses always occur under normal operating conditions of an accelerator, shielding studies need to be performed not only around the reactor but also along the beam line from the accelerator to the spallation target. This paper illustrates a preliminary evaluation, using Monte Carlo methods, of the additional shielding to be located around the reactor structures, the beam transport line and the existing reactor building to allow access into the reactor hall and to restrict the doses outside to their legal limits. (authors)

  12. Neutronics analysis of minor actinides transmutation in a fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Yang, Chao; Cao, Liangzhi; Wu, Hongchun; Zheng, Youqi; Zu, Tiejun

    2013-01-01

    Highlights: • A fusion fission hybrid system for MA transmutation is proposed. • The analysis of neutronics effects on the transmutation is performed. • The transmutation rate of MA reaches 86.5% by 25 times of recycling. -- Abstract: The minor actinides (MAs) transmutation in a fusion-driven subcritical system is analyzed in this paper. The subcritical reactor is driven by a tokamak D-T fusion device with relatively easily achieved plasma parameters and tokamak technologies. The MAs discharged from the light water reactor (LWR) are loaded in transmutation zone. Sodium is used as the coolant. The mass percentage of the reprocessed plutonium (Pu) in the fuel is raised from 0 to 48% and stepped by 12% to determine its effect on the MAs transmutation. The lesser the Pu is loaded, the larger the MAs transmutation rate is, but the smaller the energy multiplication factor is. The neutronics analysis of two loading patterns is performed and compared. The loading pattern where the mass percentage of Pu in two regions is 15% and 32.9% respectively is conducive to the improvement of the transmutation fraction within the limits of burn-up. The final transmutation fraction of MAs can reach 17.8% after five years of irradiation. The multiple recycling is investigated. The transmutation fraction of MAs can reach about 61.8% after six times of recycling, and goes up to about 86.5% after 25

  13. A study of calculation methodology and experimental measurements of the kinetic parameters for source driven subcritical systems

    International Nuclear Information System (INIS)

    Lee, Seung Min

    2009-01-01

    This work presents a theoretical study of reactor kinetics focusing on the methodology of calculation and the experimental measurements of the so-called kinetic parameters. A comparison between the methodology based on the Dulla's formalism and the classical method is made. The objective is to exhibit the dependence of the parameters on subcriticality level and perturbation. Two different slab type systems were considered: thermal one and fast one, both with homogeneous media. One group diffusion model was used for the fast reactor, and for the thermal system, two groups diffusion model, considering, in both case, only one precursor's family. The solutions were obtained using the expansion method. Also, descriptions of the main experimental methods of measurements of the kinetic parameters are presented in order to put a question about the compatibility of these methods in subcritical region. (author)

  14. The consequences of a sharp temperature change in the fuel pins of an accelerator-driven subcritical system

    International Nuclear Information System (INIS)

    Dagan, R.; Jianu, A.; Weisenburger, A.; Schikorr, M.; Rimpault, G.

    2013-01-01

    The effect of temperature changes and in particular those that are accompanied by strong gradients was extensively investigated for fast reactors. Subcritical systems designed for their transmutation ability are to some extent similar to critical power reactors in their subassembly structure. However, they differ in two main aspects. First, the coolant in a subcritical system is lead or lead-bismuth eutectic (LBE) and not sodium, and second, the main cause for steep temperature gradients in a fast power reactor is sudden control rod insertion, or scram, whereas in subcritical systems shutdown of the accelerator and its proton beam is the main cause for temperature gradients. Furthermore, the increased probability of operational interruptions in an accelerator driven system is largely due to the instability of the accelerator generating the proton beam. This study uses the knowledge gained from fast reactors as a preliminary reference and concentrates further on the unique features of the proposed subcritical systems. In particular, the effect of beam trips on the fuel pin integrity is evaluated as a function of the temperature gradients and the duration of the beam trips. It seems, however, that the largest hazard to the fuel pin integrity is due to the lead (or LBE) coolant. In particular, the stability of the protective oxide layer built on the clad surface with the lead coolant appears quite sensitive to sudden temperature changes. In the second part of this study, several available experimental results show that even very moderate temperature changes are sufficient to cause crack formation in the oxide layer thereby exposing the clad surface to enhanced LBE corrosion. In the worst case, complete exfoliation of the magnetite outer layer is observed. As a consequence, clad failure probability due to corrosion is considerably increased. (authors)

  15. Nuclear fission sustainability with subcritical reactors driven by external neutron sources

    International Nuclear Information System (INIS)

    Lafuente, A.; Piera, M.

    2011-01-01

    Although nuclear breeder reactors are a promising way to enhance the potential energy currently retrievable from the Uranium reserves, they still have disadvantages because of their safety features (i.e. poor stabilizing mechanisms) and the security of their fuel cycle (diversion of Pu for non-civilian purposes). Loading natural nuclear fuels to a reactor and completely burning them without reprocessing would be ideal, however, this is not possible in critical reactors due to the limitations imposed by the maximum achievable burn-up. An alternative option to attain very high percentages of nuclear natural materials exploitation, while meeting other objectives of Nuclear Sustainability, could consist of using externally-driven subcritical reactors to reach the desired high burn-ups (of the order of 30% and more) without reprocessing. Such scheme would lead to an efficient exploitation of the available raw material, without any risk of proliferation. Exploring this type of reactor concept, this paper analyzes the different ways to accomplish this goal while identifying potential setbacks.

  16. Catalytic effect of different reactor materials under subcritical water conditions: decarboxylation of cysteic acid into taurine

    Science.gov (United States)

    Faisal, M.

    2018-03-01

    In order to understand the influence of reactor materials on the catalytic effect for a particular reaction, the decomposition of cysteic acid from Ni/Fe-based alloy reactors under subcritical water conditions was examined. Experiments were carried out in three batch reactors made of Inconel 625, Hastelloy C-22 and SUS 316 over temperatures of 200 to 300 °C. The highest amount of eluted metals was found for SUS 316. The results demonstrated that reactor materials contribute to the resulting product. Under the tested conditions, cysteic acid decomposes readily with SUS 316. However, the Ni-based materials (Inconel 625 and Hastelloy C-22) show better resistance to metal elution. It was found that among the materials used in this work, SUS 316 gave the highest reaction rate constant of 0.1934 s‑1. The same results were obtained at temperatures of 260 and 300 °C. Investigation of the Arrhenius activation energy revealed that the highest activation energy was for Hastelloy C-22 (109 kJ/mol), followed by Inconel 625 (90 kJ/mol) and SUS 316 (70 kJ/mol). The decomposition rate of cysteic acid was found to follow the results for the trend of the eluted metals. Therefore, it can be concluded that the decomposition of cysteic acid was catalyzed by the elution of heavy metals from the surface of the reactor. The highest amount of taurine from the decarboxylation of cysteic acid was obtained from SUS 316.

  17. International conference on sub-critical accelerator driven systems. Proceedings

    International Nuclear Information System (INIS)

    Litovkina, L.P.; Titarenko, Yu.E.

    1999-01-01

    The International Meeting on Sub-Critical Accelerator Driven Systems was organized by the State Scientific Center - Institute for Theoretical and Experimental Physics with participation of Atomic Ministry of RF. The Meeting objective was to analyze the recent achievements and tendencies of the accelerator-driven systems development. The Meeting program covers a broad range of problems including the accelerator-driven systems (ADS) conceptual design; analyzing the ADS role in nuclear fuel cycle; accuracy of modeling the main parameters of ADS; conceptual design of high-current accelerators. Moreover, the results of recent experimental and theoretical studies on nuclear data accumulation to support the ADS technologies are presented. About 70 scientists from the main scientific centers of Russia, as well as scientists from USA, France, Belgium, India, and Yugoslavia, attended the meeting and presented 44 works [ru

  18. Application of the modified neutron source multiplication method for a measurement of sub-criticality in AGN-201K reactor

    International Nuclear Information System (INIS)

    Myung-Hyun Kim

    2010-01-01

    Measurement of sub-criticality is a challenging and required task in nuclear industry both for nuclear criticality safety and physics test in nuclear power plant. A relatively new method named as Modified Neutron Source Multiplication Method (MNSM) was proposed in Japan. This method is an improvement of traditional Neutron Source Multiplication (NSM) Method, in which three correction factors are applied additionally. In this study, MNSM was tested in calculation of rod worth using an educational reactor in Kyung Hee University, AGN-201K. For this study, a revised nuclear data library and a neutron transport code system TRANSX-PARTISN were used for the calculation of correction factors for various control rod positions and source locations. Experiments were designed and performed to enhance errors in NSM from the location effects of source and detectors. MNSM can correct these effects but current results showed not much correction effects. (author)

  19. Recirculator "SALO" - A basis for creation of a model of sub-critical reactor controlled by the accelerator

    NARCIS (Netherlands)

    Gann, V.V.; Guk, I.S.; Dovbnya, A.N.; Kononenko, S.G.; Kostromin, A.S.; Peev, F.A.; Prochorets, I.M.; Soldatov, C.A.; Tarasenko, A.S.; Wiel, van der M.J.; Botman, J.I.M.

    2006-01-01

    Recently opportunities of creation test facility with sub-crit. reactors controlled by accelerators of particles are actively discussed. At an initial stage of these researches it is the most expedient to use electron accelerators as cost of such facility will be much less, than at use of proton

  20. Economic analysis of the fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Huang Desuo; Wu Yican; Chu Delin; Hu Liqin

    2004-01-01

    The economic performance of the Fusion-Driven Subcritical system (FDS) is discussed. At first, as an example, the impacts of parameters, such as plasma aspect-ratio, elongation, normalized beta, on-axis toroidal field and the blanket energy-gain are analyzed on the costs of the typical case (moderate aspect-ratio) of FDS. Then, the economic characteristics of the 3 possible scenarios of FDS are estimated with respect to the neutronics parameters. The results calculated with the SYSCODE developed by the FDS team show that the cost of electricity of Scenario-1 (low aspect-ratio) and Scenario-2 (moderate aspect-ratio) of FDS is cheaper than that of pure fusion power plant at the same plane size (1 GW e ). The cost of electricity of the FDS power plant depends heavily on the functions of blanket and the blanket energy-gain. (authors)

  1. Design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium

    International Nuclear Information System (INIS)

    Medina C, D.; Hernandez A, P.; Letechipia de L, C.; Vega C, H. R.; Sajo B, L.

    2015-09-01

    This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a 252 Cf source, whose dose levels at the periphery allows its use in teaching and research activities. The design was realized by the Monte Carlo method, where the geometry, dimensions and the fuel was varied in order to obtain the best design. The result was a cubic reactor of 110 cm of side, with graphite moderator and reflector. In the central part having 9 ducts of 3 cm in diameter, eight of them are 110 cm long, which were placed on the Y axis; the separation between each duct is 10 cm. The central duct has 60 cm in length and this contains the 252 Cf source, also there are two irradiation channels and the other six contain a molten salt ( 7 LiF - BeF 2 - ThF 4 - UF 4 ) as fuel. For the design the k eff was calculated, neutron spectra and ambient dose equivalent. In the first instance the above was calculated for a virgin fuel, was called case 1; then a percentage of 233 U was used and the percentage of Th was decreased and was called case 2. This with the purpose of comparing two different fuels operating within the reactor. For the two irradiation ducts three positions are used: center, back and front, in each duct in order to have different flows. (Author)

  2. Studies of thermal hydraulics and heat transfer in cascade subcritical molten salt reactor

    International Nuclear Information System (INIS)

    Aysen, E.M.; Sedov, A.A.; Subbotin, A.S.

    2005-01-01

    Full text of publication follows: Cascade Subcritical Molten Salt Reactor (CSMSR) consists of three main parts: accelerator-driven proton-bombarded target, central and peripheral zones. External neutrons generated in the result of interaction of protons with the target nuclei are multiplied then in the central zone and leak farther into the peripheral reactor zone, where an efficient burning of Minor Actinides dissolved in a molten salt fluoride composition is produced. The bunch of target and two zones is designed so that preset subcriticality of reactor would not be less than 1% of k eff . A characteristic feature of the reactor is a high density of neutron flux (2.10 15 n/cm 2 s) in the central zone and target and very high volumetric power rate (2000 - 6000 W/cm 3 ) in all the parts of CSMSR. To provide a workability of the core structures under condition of so big level of power rate it is necessary to impose strict limitations on the temperatures and temperature gradients developed in the coolants and constructions. In this reason it has been arranged a calculational-designing study to reveal the problems of heat transfer in the coolant and core structures and to find more appropriate variant of the core and target design, which is a compromise of contradictory requirements: provision of high neutron flux and coolability of the core structures. In this paper the results of studies of thermal hydraulics and heat transfer in the core zones and proton-beam target are presented. Different variants of the target and central zone design as well as application of different kind of coolants in them are discussed and the main problems of heat removal in their structures are analyzed. Multidimensional fields of velocity and temperature got in thermal hydraulics calculations for free flow of fuelled molten salt in cylindrical-cave peripheral CSMSR zone without structures inside are demonstrated. The role of turbulent exchange of momentum and heat for free flow in the

  3. The Chain-Length Distribution in Subcritical Systems

    International Nuclear Information System (INIS)

    Nolen, Steven Douglas

    2000-01-01

    The individual fission chains that appear in any neutron multiplying system provide a means, via neutron noise analysis, to unlock a wealth of information regarding the nature of the system. This work begins by determining the probability density distributions for fission chain lengths in zero-dimensional systems over a range of prompt neutron multiplication constant (K) values. This section is followed by showing how the integral representation of the chain-length distribution can be used to obtain an estimate of the system's subcritical prompt multiplication (MP). The lifetime of the chains is then used to provide a basis for determining whether a neutron noise analysis will be successful in assessing the neutron multiplication constant, k, of the system in the presence of a strong intrinsic source. A Monte Carlo transport code, MC++, is used to model the evolution of the individual fission chains and to determine how they are influenced by spatial effects. The dissertation concludes by demonstrating how experimental validation of certain global system parameters by neutron noise analysis may be precluded in situations in which the system K is relatively low and in which realistic detector efficiencies are simulated

  4. The Chain-Length Distribution in Subcritical Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nolen, Steven Douglas [Texas A & M Univ., College Station, TX (United States)

    2000-06-01

    The individual fission chains that appear in any neutron multiplying system provide a means, via neutron noise analysis, to unlock a wealth of information regarding the nature of the system. This work begins by determining the probability density distributions for fission chain lengths in zero-dimensional systems over a range of prompt neutron multiplication constant (K) values. This section is followed by showing how the integral representation of the chain-length distribution can be used to obtain an estimate of the system's subcritical prompt multiplication (MP). The lifetime of the chains is then used to provide a basis for determining whether a neutron noise analysis will be successful in assessing the neutron multiplication constant, k, of the system in the presence of a strong intrinsic source. A Monte Carlo transport code, MC++, is used to model the evolution of the individual fission chains and to determine how they are influenced by spatial effects. The dissertation concludes by demonstrating how experimental validation of certain global system parameters by neutron noise analysis may be precluded in situations in which the system K is relatively low and in which realistic detector efficiencies are simulated.

  5. Methods for the reactivity evaluation in subcritical systems analysis: a review

    International Nuclear Information System (INIS)

    Dulla, S.; Picca, P.; Carta, M.

    2011-01-01

    The assessment of the subcritical source-driven system technology for waste incineration and power production requires the development of reliable and efficient techniques for the reactivity evaluation and monitoring. Starting from the standard methods developed for close-to-criticality systems, extensive research activities have been carried out to analyze the behavior of subcritical assembly in time-dependent condition and to infer the subcriticality level from local flux values. In the present work, a review of some key aspects in the method development for ADS analysis is proposed, with special attention to the techniques for reactivity evaluation. (author)

  6. Preliminary analysis of typical transients in fusion driven subcritical system (FDS-I)

    International Nuclear Information System (INIS)

    Bai Yunqing; Ke Yan; Wu Yican

    2007-01-01

    The potential safety characteristic is expected as one of the advantages of fusion-driven subcritical system (FDS-I) for the transmutation and incineration of nuclear waste compared with the critical reactor. Transients of the FDS-I may occur due to the perturbation of external neutron source, the failure of functional device, and the occurrence of the uncontrolled event. As typical transient scenarios, the following cases were analyzed: unprotected plasma overpower (UPOP), unprotected loss of flow (ULOF), unprotected transient overpower (UTOP). The transient analyses for the FDS-I were performed with a coupled two-dimensional thermal-hydraulics and neutronics transient analysis code NTC2D. The negative feedback of reactivity is the interesting safety feature of FDS-I as temperature increase, due to the fuel form of the circulating particle. The present simulation results showed that the current FDS-I design has a resistance against severe transient scenarios. (author)

  7. Neutron fluctuation analysis in a subcritical multiplying system with a stochastically pulsed poisson source

    International Nuclear Information System (INIS)

    Kostic, Lj.

    2003-01-01

    The influence of the stochastically pulsed Poisson source to the statistical properties of the subcritical multiplying system is analyzed in the paper. It is shown a strong dependence on the pulse period and pulse width of the source (author)

  8. Hydrolysis of polycarbonate in sub-critical water in fused silica capillary reactor with in situ Raman spectroscopy

    Science.gov (United States)

    Pan, Z.; Chou, I-Ming; Burruss, R.C.

    2009-01-01

    The advantages of using fused silica capillary reactor (FSCR) instead of conventional autoclave for studying chemical reactions at elevated pressure and temperature conditions were demonstrated in this study, including the allowance for visual observation under a microscope and in situ Raman spectroscopic characterization of polycarbonate and coexisting phases during hydrolysis in subcritical water.

  9. Sensitivity Analysis of Core Neutronic Parameters in Electron Accelerator-driven Subcritical Advanced Liquid Metal Reactor

    Directory of Open Access Journals (Sweden)

    Marziye Ebrahimkhani

    2016-02-01

    Full Text Available Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (Ee and source multiplication coefficient (ks, has been investigated. A Monte Carlo code (MCNPX_2.6 has been used to calculate neutronic parameters such as effective multiplication coefficient (keff, net neutron multiplication (M, neutron yield (Yn/e, energy constant gain (G0, energy gain (G, importance of neutron source (φ∗, axial and radial distributions of neutron flux, and power peaking factor (Pmax/Pave in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current (Ie have been decreased in the highest case of ks, but G and φ∗ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing Ee from 100 MeV up to 1 GeV, Yn/e and G improved by 91.09% and 10.21%, and Ie and Pacc decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np–Pu assemblies on the periphery allows for a consistent keff because the Np–Pu assemblies experience less burn-up.

  10. Monte-Carlo Modeling of Parameters of a Subcritical Cascade Reactor Based on MSBR and LMFBR Technologies

    CERN Document Server

    Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H

    2001-01-01

    Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.

  11. Monte-Carlo modeling of parameters of a subcritical cascade reactor based on MSBR and LMFBR technologies

    International Nuclear Information System (INIS)

    Bznuni, S.A.; Zhamkochyan, V.M.; Khudaverdyan, A.G.; Barashenkov, V.S.; Sosnin, A.N.; Polanski, A.

    2001-01-01

    Parameters are investigated of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k eff = 0.94 - 0.98), is capable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10 14 cm 12 · s -1 , in the fast booster zone is 5.12 · 10 15 cm 12 · s -1 at k eff = 0.98 and proton beam current I = 2.1 mA. (author)

  12. Discriminators for the Accelerator-Based Conversion (ABC) concept using a subcritical molten salt system

    International Nuclear Information System (INIS)

    Arthur, E.; Busksa, J.; Davidson, W.; Poston, D.

    1995-05-01

    Discriminators are described that quantify enhancements added to plutonium destruction and/or nuclear waste transmutation systems through use of an accelerator/fluid fuel combination. This combination produces a robust and flexible nuclear system capable of the destruction of all major long-lived actinides (including plutonium) and fission products. The discriminators discussed in this report are (1) impact of subcritical operation on safety, (2) impact of subcritical and fluid fuel operation on plutonium burnout scenarios, and (3) neutron economy enhancements brought about by subcritical operation. Neutron economy enhancements are quantified through assessment of long-term dose reduction resulting from transmutation of key fission products along with relaxation of processing frequencies afforded by subcritical operation

  13. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    International Nuclear Information System (INIS)

    Li Gang; Zhang Zhongshuai; Chi Qian; Liu Linmao

    2012-01-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 × 10 8 n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 μs. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  14. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    Energy Technology Data Exchange (ETDEWEB)

    Li Gang; Zhang Zhongshuai [Northeast Normal University, Changchun 130024 (China); Chi Qian [Guang Hua College of Chang Chun University, Changchun 130117 (China); Liu Linmao, E-mail: ll888@nenu.edu.cn [Northeast Normal University, Changchun 130024 (China)

    2012-11-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 Multiplication-Sign 10{sup 8} n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 {mu}s. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  15. Conception of electron beam-driven subcritical molten salt ultimate safety reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Alekseev, P.N.; Ignat`ev, V.V. [Kurchatov Institute, Moscow (Russian Federation)] [and others

    1995-10-01

    This paper is a preliminary sketch of a conception to develop the {open_quotes}ultimate safety reactor{close_quotes} using modern reactor and accelerator technologies. This approach would not require a long-range R&D program. The ultimate safety reactor could produce heat and electric energy, expand the production of fuel, or be used for the transmutation of long-lived wastes. The use of the combined double molten salt reactor system allows adequate neutron multiplication to permit using an electron accelerator for the initial neutron flux. The general parameters of such a system are discussed in this paper.

  16. Thermionic nuclear reactor systems

    International Nuclear Information System (INIS)

    Kennel, E.B.

    1986-01-01

    Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

  17. Research project on accelerator-driven subcritical system using FFAG accelerator and Kyoto University critical assembly

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Unesaki, Hironobu; Misawa, Tsuyoshi; Tanigaki, Minoru; Mori, Yoshiharu; Shiroya, Seiji; Inoue, Makoto; Ishi, Y.; Fukumoto, Shintaro

    2005-01-01

    The KART (Kumatori Accelerator-driven Reactor Test facility) project started in Research Reactor Institute, Kyoto University in fiscal year 2002 with the grant by the Japanese Ministry of Education, Culture, Sports, Science and Technology. The purpose of this research project is to demonstrate the basis feasibility of accelerator driven system (ADS), studying the effect of incident neutron energy on the effective multiplication factor in a subcritical nuclear fuel system. For this purpose, a variable-energy FFAG (Fixed Field Alternating Gradient) accelerator complex is being constructed to be coupled with the Kyoto University Critical Assembly (KUCA). The FFAG proton accelerator complex consists of ion-beta, booster and main rings. This system aims to attain 1 μA proton beam with energy range from 20 to 150 MeV with a repetition rate of 120 Hz. The first beam from the FFAG complex is expected to be available by the end of FY 2005, and the experiment on ADS with KUCA and the FFAG complex (FFAG-KUCA experiment) will start in FY 2006. Before the FFAG-KUCA experiment starts, preliminary experiments with 14 MeV neutrons are currently being performed using a Cockcroft-Walton type accelerator coupled with the KUCA. Experimental data are analyzed using continuous energy Monte-Carlo codes MVP, MCNP and MNCP-X. (author)

  18. Accelerator-driven subcritical systems - An analysis with a focus on non-proliferation and export control

    International Nuclear Information System (INIS)

    Andersson, Per; Nielsen, Fredrik; Sunhede, Daniel

    2013-01-01

    The Department of Nuclear Weapons Related Issues at The Swedish Defence Research Agency, FOI, as commissioned by the Swedish Radiation Safety Authority, SSM, conducted a study concerning Accelerator Driven Subcritical Systems, ADS, with emphasis on non-proliferation and export control. An ADS looks at first glance like a traditional nuclear reactor, but the nuclear core is designed to always remain subcritical, both during normal and off-normal conditions. Neutrons are instead supplied by an external source in the form of an proton accelerator and a spallation target. This report gives a short walk-through to the physical processes that governs the neutron flux and reactivity in the core and how they are affected by the design of the core including the accelerator and spallation target. Furthermore is the results from reactor core simulations presented, where the isotopic nuclear fuel inventory has been studied as a function of burn up and initial configuration. Finally the report contains an analysis of the potential risks involved from the perspective of nuclear proliferation and exports. This study shows that ADS in the future could constitute a proliferation concern. The subsystems and components in question share design and materials with the equivalent components in traditional reactors with the exception of the proton accelerator and spallation target, which is unique for accelerator driven systems

  19. Fuel research for subcritical and critical GEN-IV systems cooled by heavy liquid metal

    International Nuclear Information System (INIS)

    Sobolev, V.; Verwerft, M.

    2009-01-01

    The participation of the Belgian Nuclear Research Centre SCK-CEN in the worldwide GEN-IV research can be considered as an opportunity. Today's GEN-IV research at SCK-CEN is mainly driven by the interests of the project MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications). The main goal of this project is to build at SCK-CEN in Mol a new generation fast spectrum, subcritical, research and materials testing reactor MYRRHA driven by a high-energy proton accelerator. This GEN-IV MTR is cooled by heavy liquid metal (Pb-Bi) and will be used for the ADS concept demonstration, testing and qualification of new fuels, transmutation targets and innovative materials. On the European scale, MYRRHA is integrated in the Euratom FP6 Integrated Project (IP) EUROTRANS (EUROpean research programme for TRANSmutation of high level nuclear waste in an accelerator driven system), as the small-scale experimental machine for transmutation demonstration called XT-ADS. Last but not least, this experimental facility will also demonstrate the technological feasibility of the LFR (Lead-cooled Fast Reactor) GEN-IV concept; in EU the LFR design studies are performed in the framework of the Euratom FP6 ELSY (European Lead-cooled SYstem) project, where SCK-CEN is a partner. Among the research needed to ensure a safe and reliable operation of the MYRRHA/XT ADS reactor, the development and qualification of fuel and cladding materials have been recognized as one of the main key issues to be addressed

  20. A modernized and versatile startup reactivity measuring system installed at NPP Paks and its application for subcritical systems

    International Nuclear Information System (INIS)

    Czibok, T.; Dezso, Z.; Horvath, Cs.; Lipcsei, S.; Vegh, J.; Pos, I.

    2006-01-01

    In 2004 the Hungarian Paks NPP completed a project for upgrading the reactivity measuring system applied during reactor startup experiments. Almost all components of the previous system were replaced, only ex-core ionisation chambers remained unaltered. New hardware and software components were introduced for neutron flux signal handling, for data acquisition, as well as for measurement evaluation and data presentation. High-precision picoamper meters were installed at each reactor unit, current signals are handled by a portable signal processing unit. The system applies an accurate on-line reactivity calculation algorithm based on the point-kinetic model with six delayed neutron groups. Detailed off-line evaluation and analysis of startup measurements can be performed on the portable unit, as well. The paper describes the architecture, data acquisition modules, services and man-machine interface of the new system. Functions and results are illustrated with measured data recorded during a startup of Unit 3. In 2003 and 2004 the RMR was installed and tested at all Paks NPP units successfully and now it is in regular use during unit startups. The second part of the paper illustrates an extension of the new system to perform reactivity measurements using the well-known Rossi-α and Feynman-α statistical methods. The modified system was needed to estimate the reactivity of a subcritical system formed by damaged fuel assemblies stored at the fuel service pit of Paks Unit 2. Theoretical background of the applied algorithms is outlined, then results of validation tests and on site measurements are treated. The measurements have shown that the subcriticality of the damaged fuel was sufficiently deep if the high boron concentration in the fuel service pit was maintained

  1. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  2. Sub-criticality monitoring for ADTR trademark control

    International Nuclear Information System (INIS)

    Ashworth, Roger

    2014-01-01

    Following the debut of the Accelerator Driven Thorium Reactor (ADTR trademark) Power Station at ENC 2010 in Barcelona, thorium as a reactor fuel has gained increasing support. The ADTR trademark concept reactor introduced the combination of an accelerator driven system (ADS) with traditional control rod technology, to provide a very high gain novel sub-critical ADS reactor design. The high gain of the system, while significantly reducing the demands on the accelerator design, pushes up operational sub-criticality (k eff ) closer to unity. In this paper we review this design and the progress made since ENC 2010. We compare 2 different methods of measuring the sub-critical neutron multiplication factor as the fuel cycle develops. The paper discusses the most recent work on k eff measurement and the interesting relationship between neutron flux, accelerator current and fuel temperature when using beam pulse methods to determine operational sub-criticality, of which a European patent is being granted. (orig.)

  3. Analytical solution of point kinetic equations for sub-critical systems

    International Nuclear Information System (INIS)

    Henrice Junior, Edson; Goncalves, Alessandro C.

    2013-01-01

    This article presents an analytical solution for the set of point kinetic equations for sub-critical reactors. This solution stems from the ordinary, non-homogeneous differential equation that rules the neutron density and that presents the incomplete Gamma function in its functional form. The method used proved advantageous and allowed practical applications such as the linear insertion of reactivity, considering an external constant source or with both varying linearly. (author)

  4. Experimental subcritical reactivity determinations employing APSD measurements with pulse mode detectors in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Lee, Seung Min; Diniz, Ricardo; Jerez, Rogerio

    2011-01-01

    This work aims to determine experimentally the subcritical reactivity levels of several configurations of the IPEN/MB-01 reactor in an approach based on the subcritical kinetic model developed by Gandini and Salvatores. The procedure employs the measurements of the APSD (Auto Power Spectral Density) using pulse mode detectors. The proposed approach is based only on measured quantities such as counting rates and the parameters arising from the least square approach of the APSD. Other difficult quantity such as detector efficiencies is not needed in the method. Several measurements of APSD were performed in varying degrees of sub-criticality (up to around -7000 pcm). The APSD data were least-square fitted to get the prompt decay mode (α). Beside the startup source, an external neutron sources of Am-Be was installed near the core in order to improve neutron count statistics. The final experimental results are of very good quality. The experiment shows clearly that the classical one point kinetic theory cannot describe the measured reactivity. MCNP K eff results were compared to the corresponding experimental results. The agreement was fairly good. (author)

  5. Validation of neutronic methods applied to the analysis of fast subcritical systems. The MUSE-2 experiments

    International Nuclear Information System (INIS)

    Soule, R.; Salvatores, M.; Jacqmin, R.; Martini, M.; Lebrat, J.F.; Bertrand, P.; Broccoli, U.; Peluso, V.

    1997-01-01

    In the framework of the French SPIN program devoted to the separation and the transmutation of radioactive wastes, the CEA has launched the ISAAC program to investigate the potential of accelerator-driven systems and to provide an experimental validation of the physics characteristics of these systems. The neutronics of the subcritical core needs experimental validation. This can be done by decoupling the problem of the neutron source from the problem of the subcritical medium. Experiments with a well known external source placed in a subcritical medium have been performed in the MASURCA facility. The results confirm the high accuracy achievable with such experiments and the good quality of the ERANOS code system predictions. (author)

  6. Validation of neutronic methods applied to the analysis of fast subcritical systems. The MUSE-2 experiments

    Energy Technology Data Exchange (ETDEWEB)

    Soule, R; Salvatores, M; Jacqmin, R; Martini, M; Lebrat, J F; Bertrand, P [CEA Centre d` Etudes de Cadarache, Service de Physique des Reacteurs et du Cycle, 13 - Saint-Paul-lez-Durance (France); Broccoli, U; Peluso, V

    1998-12-31

    In the framework of the French SPIN program devoted to the separation and the transmutation of radioactive wastes, the CEA has launched the ISAAC program to investigate the potential of accelerator-driven systems and to provide an experimental validation of the physics characteristics of these systems. The neutronics of the subcritical core needs experimental validation. This can be done by decoupling the problem of the neutron source from the problem of the subcritical medium. Experiments with a well known external source placed in a subcritical medium have been performed in the MASURCA facility. The results confirm the high accuracy achievable with such experiments and the good quality of the ERANOS code system predictions. (author)

  7. Preliminary analysis of advanced equilibrium configuration for the fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Chu Delin; Wu Bin; Wu Yican

    2003-01-01

    The Fusion-Driven Subcritical System (FDS) is a subcritical nuclear energy system driven by fusion neutron source. In this paper, an advanced plasma configuration for FDS system has been proposed, which aims at high beta, high bootstrap current and good confinement. A fixed-boundary equilibrium code has been used to obtain ideal equilibrium configuration. In order to determine the feasibility of FDS operation, a two-dimensional time-dependent free boundary simulation code has been adopted to simulate time-scale evolution of plasma current profile and boundary position. By analyses, the Reversed Shear mode as the most attractive one has been recommended for the FDS equilibrium configuration design

  8. Computational investigation of 99Mo, 89Sr, and 131I production rates in a subcritical UO2(NO32 aqueous solution reactor driven by a 30-MeV proton accelerator

    Directory of Open Access Journals (Sweden)

    Z. Gholamzadeh

    2015-12-01

    Full Text Available The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing 99Mo. In this method, the medical isotope production system itself is used to extract 99Mo or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of 99Mo by irradiating targets. In this study, the neutronic performance and 99Mo, 89Sr, and 131I production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ∼1,500 Ci/wk (∼325 6-day Ci of 99Mo at the end of a cycle.

  9. Subcritical nuclear assembly

    International Nuclear Information System (INIS)

    Vega C, H. R.

    2014-08-01

    A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a 239 PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

  10. Calculation and analysis of burnup and optimum core design in accelerator driven sub-critical system

    International Nuclear Information System (INIS)

    Wang Yuwei; Yang Yongwei; Cui Pengfei

    2011-01-01

    The premise of the accelerator driven sub-critical system (ADS) in the accident is still subcritical, the biggest k eff change with burn time is less than 1.5% and the cladding material, HT9 steel, can withstand the maximum radiation damage, core fuel area is divided into fuel transmutation area and fuel multiplication area, and fuel transmutation area maintains the same fuel composition in the whole process. Through the analysis of the composition of the fuel, shape of core layout and the power distribution, etc., supposed outer and inner Pu enrichment ratio range of 1.0-1.5, then the fuel components of fuel multiplication area was adjusted. Time evolution of k eff was calculated by COUPLED2 which coupled with MCNP and ORIGEN. At the same time the power peaking factors, minoractinides transmutation rate desired to maximization and burnup were considered. A sub-critical system fitting for engineering practice was established. (authors)

  11. Conceptual design of the fusion-driven subcritical system FDS-I

    International Nuclear Information System (INIS)

    Wu, Y.; Zheng, S.; Zhu, X.; Wang, W.; Wang, H.; Liu, S.; Bai, Y.; Chen, H.; Hu, L.; Chen, M.; Huang, Q.; Huang, D.; Zhang, S.; Li, J.; Chu, D.; Jiang, J.; Song, Y.

    2006-01-01

    The fusion-driven subcritical system (named FDS-I) was previously proposed as an intermediate step toward the final application of fusion energy. A conceptual design of the FDS-I is presented, which consists of the fusion neutron driver with relatively easy-achieved plasma parameters, and the He-gas/liquid lithium-lead Dual-cooled subcritical Waste Transmutation (DWT) blanket used to transmute long-lived radioactive wastes and to generate energy on the basis of self-sustainable fission and fusion fuel cycle. An overview of the FDS-I is given and the specifications of the design analysis are summarized

  12. Benchmarks of subcriticality in accelerator-driven system at Kyoto University Critical Assembly

    Directory of Open Access Journals (Sweden)

    Cheol Ho Pyeon

    2017-09-01

    Full Text Available Basic research on the accelerator-driven system is conducted by combining 235U-fueled and 232Th-loaded cores in the Kyoto University Critical Assembly with the pulsed neutron generator (14 MeV neutrons and the proton beam accelerator (100 MeV protons with a heavy metal target. The results of experimental subcriticality are presented with a wide range of subcriticality level between near critical and 10,000 pcm, as obtained by the pulsed neutron source method, the Feynman-α method, and the neutron source multiplication method.

  13. Experimental study on accelerator driven subcritical reactor. JAERI's nuclear research promotion program, H12-031 (Contract research)

    International Nuclear Information System (INIS)

    Shiroya, Seiji; Misawa, Tsuyoshi; Unesaki, Hironobu

    2004-03-01

    In view of the future plan of Research Reactor Institute, Kyoto University (KURRI), the present study consisted of 1) the transmission experiments of high energy neutrons through materials, 2) experimental simulation of ADSR using the Kyoto University Critical Assembly (KUCA), and 3) conceptual neutronics design study on Kyoto University Reactor (KUR) type ADSR using the MCNPX code. The purpose of the present study was not only to obtain the knowledge usable for the realization of ADSR as a new neutron source for research but also to select technical issues in the field of reactor physics for the development of ADSR in general. Through the present study, valuable knowledge on the basic nuclear characteristics of ADSR was obtained both theoretically and experimentally. This kind of knowledge is indispensable to promote the study on ADSR further. If one dare say the main part of knowledge in short words, the basic nuclear characteristics of ADSR is overwhelmed by the characteristics of the subcritical reactor as expected. For the realization of ADSR in the future, it is considered to be necessary to accumulate results of research steadily. For this purpose, it is inevitable 1) to compile the more precise nuclear data for the wide energy range, 2) to establish experimental techniques for reactor physics study on ADSR including subcriticality measurement and absolute neutron flux measurement from the low energy region to the high energy region, and 3) to develop neutronics calculation tools which facilitate to take into account the neutron generation process by the spallation reaction and the delayed neutron behavior. (author)

  14. Nuclear power history calculation for subcritical systems using Euler-MacLaurin formula

    International Nuclear Information System (INIS)

    Henrice Junior, Edson; Goncalves, Alessandro da Cruz

    2013-01-01

    This paper presents an efficient method for calculating the reactivity using inverse point kinetic equation for subcritical systems by applying the Euler-MacLaurin summation formula to calculate the nuclear power history. In accordance with the accuracy of the numerical results, this method does not require a large number of points for calculation, providing accurate results with low computational cost. (author)

  15. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  16. MCNPX and MCB coupled methodology for the burnup calculation of the KIPT accelerator driven subcritical system

    International Nuclear Information System (INIS)

    Zhong, Z.; Gohar, Y.; Talamo, A.

    2009-01-01

    Argonne National Laboratory (ANL) of USA and Kharkov Inst. of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical facility (ADS). The facility will be utilized for basic research, medical isotopes production, and training young nuclear specialists. The burnup methodology and analysis of the KIPT ADS are presented in this paper. MCNPX and MCB Monte Carlo computer codes have been utilized. MCNPX has the capability of performing electron, photon and neutron coupled transport problems, but it lacks the burnup capability for driven subcritical systems. MCB has the capability for performing the burnup calculation of driven subcritical systems, while it cannot transport electrons. A calculational methodology coupling MCNPX and MCB has been developed, which can exploit the electrons transport capability of MCNPX for neutron production and the burnup capability of MCB for driven subcritical systems. In this procedure, a neutron source file is generated using MCNPX transport calculation, preserving the neutrons yield from photonuclear reactions initiated by electrons, and this source file is utilized by MCB for the burnup analyses with the same geometrical model. In this way, the ADS depletion calculation can be accurately. (authors)

  17. Inverse kinetics method with source term for subcriticality measurements during criticality approach in the IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Loureiro, Cesar Augusto Domingues; Santos, Adimir dos

    2009-01-01

    In reactor physics tests which are performed at the startup after refueling the commercial PWRs, it is important to monitor subcriticality continuously during criticality approach. Reactivity measurements by the inverse kinetics method are widely used during the operation of a nuclear reactor and it is possible to perform an online reactivity measurement based on the point reactor kinetics equations. This technique is successful applied at sufficiently high power level or to a core without an external neutron source where the neutron source term in point reactor kinetics equations may be neglected. For operation at low power levels, the contribution of the neutron source must be taken into account and this implies the knowledge of a quantity proportional to the source strength, and then it should be determined. Some experiments have been performed in the IPEN/MB-01 Research Reactor for the determination of the Source Term, using the Least Square Inverse Kinetics Method (LSIKM). A digital reactivity meter which neglects the source term is used to calculate the reactivity and then the source term can be determined by the LSIKM. After determining the source term, its value can be added to the algorithm and the reactivity can be determined again, considering the source term. The new digital reactivity meter can be used now to monitor reactivity during the criticality approach and the measured value for the reactivity is more precise than the meter which neglects the source term. (author)

  18. BRAHMMA - accelerator driven subcritical facility

    International Nuclear Information System (INIS)

    Roy, Tushar; Shukla, Shefali; Shukla, M.; Ray, N.K.; Kashyap, Y.S.; Patel, T.; Gadkari, S.C.

    2017-01-01

    Accelerator Driven Subcritical systems are being studied worldwide for their potential in burning minor actinides and reducing long term radiotoxicity of spent nuclear fuels. In order to pursue the physics studies of Accelerator Driven Subcritical systems, a thermal subcritical assembly BRAHMMA (BeOReflectedAndHDPeModeratedMultiplying Assembly) has been developed at Purnima Labs, BARC. The facility consists of two major components: Subcritical core and Accelerator (DT/ DD Purnima Neutron Generator)

  19. A study of physics of sub-critical multiplicative systems driven by sources and the utilization of deterministic codes in calculation of this systems

    International Nuclear Information System (INIS)

    Antunes, Alberi

    2008-01-01

    This work presents the Physics of Source Driven Systems (ADS). It shows some statics and K i netics parameters of the reactor Physics and when it is sub critical, that are important in evaluation and definition of these systems. The objective is to demonstrate that there are differences in parameters when the reactor is critical. Moreover, the work shows the differences observed in the parameters for different calculation models. Two calculation methodologies are shown In this dissertation: Gandini and Salvatores and Dulla, and some parameters are calculated. The ANISN deterministic transport code is used in calculation in order to compare these parameters. In a subcritical configuration of IPEN-MB-01 Reactor driven by an external source some parameters are calculated. The conclusions about calculation realized are presented in end of work. (author)

  20. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument’s LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  1. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Klein, Steven Karl; Determan, John C.

    2015-01-01

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument's LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  2. About the possibility of use of different types of targets as a neutron source for subcritical nuclear reactor driven by particle beam accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Avdeev, E.F.; Dorokhovich, S.L.; Chusov, I.A. [Obninsk Institute of Nuclear Power Engineering (Russian Federation)

    1995-10-01

    The schemes of jet gas and liquid targets as well as the gastargets with a solid phase dispersion are introduced to use to receive the neutrons admitted to a subcritical reactor core. The possible variants of target position in the reactor are considered, target characteristics are calculated. The authors pay a great attention to the estimation of radioactive products yield receiving due to the interaction of the beam with the target.

  3. Concept of the thorium fuelled accelerator driven subcritical system for both energy production and TRU incineration - 'TASSE'

    International Nuclear Information System (INIS)

    Slessarev, I.; Berthou, V.; Salvatores, M.; Tchistiakov, A.

    1999-01-01

    The TASSE is the concept of the subcritical accelerator driven system with 'TRU-free' fuel cycle and the continuous Th-feed regime. The tightness of Th neutronics call inevitably the subcritical mode of work. Two types of neutron spectra are recommended: fast and super-thermal (well thermalized) ones. TASSE fuel cycle could have the following options: (i) without any fuel recycling and reprocessing (once-through fuel cycle option) for maximum fuel cycle simplicity. However, subcriticality level (1- K eff ) is essential and it requires high power accelerators; (ii) with the partial or, eventually, full U recycling 'on line' including the separation (U + Pa + Th) component from TRU + FP component which can be considered as wastes. Relatively small mass of fuel have to be reprocessed. Moreover, the requirement to separation is very soft. In this case, recycling allows to minimise subcriticality and smaller accelerators can be acceptable. The TASSE is oriented on 'clean' nuclear energy production and TRU burning with the following attractive features: (1) For the long term perspective, TASSEs have a rather limited mass of long-lived radioactive wastes, consisting mostly of Th, U and Pa nuclides. One can see the considerable reduction of waste toxicity by the factor of 1000 (or even more) in the magnitude regarding current PWR's and by the factor of 10-100 regarding (PWR's + dedicated burners) scenario. (2) Relatively low amounts of Th would have to be mined: approximately a factor of 100 lower than the U mined for PWR's. With TASSEs, nuclear power has practically inexhaustible (for a long future) and cheap fuel resources, taking into account that Thorium reserves exceed Uranium PWR fuel reserves by factor of 10 3 . (3) TASSEs are able to burnout all previously accumulated transuraniums as well as weapons grade materials during PWR's replacement over a period of approximately 50 years. No actinide fuel waste is foreseen for this period of time. There is no need to

  4. Modeling of the CTEx subcritical unit using MCNPX code

    International Nuclear Information System (INIS)

    Santos, Avelino; Silva, Ademir X. da; Rebello, Wilson F.; Cunha, Victor L. Lassance

    2011-01-01

    The present work aims at simulating the subcritical unit of Army Technology Center (CTEx) namely ARGUS pile (subcritical uranium-graphite arrangement) by using the computational code MCNPX. Once such modeling is finished, it could be used in k-effective calculations for systems using natural uranium as fuel, for instance. ARGUS is a subcritical assembly which uses reactor-grade graphite as moderator of fission neutrons and metallic uranium fuel rods with aluminum cladding. The pile is driven by an Am-Be spontaneous neutron source. In order to achieve a higher value for k eff , a higher concentration of U235 can be proposed, provided it safely remains below one. (author)

  5. Minimising the economic cost and risk to accelerator-driven subcritical reactor technology: The case of designing for flexibility: Part 1

    International Nuclear Information System (INIS)

    Steer, Steven J.; Cardin, Michel-Alexandre; Nuttall, William J.; Parks, Geoffrey T.; Gonçalves, Leonardo V.N.

    2012-01-01

    Highlights: ► Accelerator performance is a risk to ADSR reactor technology demonstration. ► Sensitivity of ADSR economic value to accelerator performance is assessed. ► Economic value of ADSRs with and without accelerator redundancy is tested. ► Real options identify design flexibility to accelerator performance uncertainty. ► Multiple ADSR “park” with a single integrated accelerator system is proposed. - Abstract: Demonstrating the generation of electricity for commercial markets with accelerator-driven subcritical reactor (ADSR) technology will incur substantial financial risk. This risk will arise from traditional uncertainties associated with the construction of nuclear power stations and also from new technology uncertainties such as the reliability of the required accelerator system. The sensitivity of the economic value of ADSRs to the reliability of the accelerator system is assessed. Using linear accelerators as an example of choice for the accelerator technology, the economic assessment considers an ADSR with either one or two accelerators driving it. The extent to which a second accelerator improves the accelerator system reliability is determined, as are the costs for that reliability improvement. Two flexible designs for the accelerator system are also considered, derived from the real options analysis technique. One seeks to achieve the benefits of both the single and dual accelerator ADSR configurations through initially planning to build a second accelerator, but only actually constructing it once it is determined to be economically beneficial to do so. The other builds and tests an accelerator before committing to constructing a reactor. Finally, a phased multiple-reactor park with an integrated system of accelerators is suggested and discussed. The park uses the principles of redundancy as for the Dual accelerator ADSR and flexibility as for the real options design, but for a lower cost per unit of electricity produced.

  6. A three-dimensional thermal and fluid dynamics analysis of a gas cooled subcritical fast reactor driven by a D-T fusion neutron source

    International Nuclear Information System (INIS)

    Angelo, G.; Andrade, D.A.; Angelo, E.; Carluccio, T.; Rossi, P.C.R.; Talamo, A.

    2011-01-01

    Highlights: → A thermal fluid dynamics numerical model was created for a gas cooled subcritical fast reactor. → Standard k-ε model, Eddy Viscosity Transport Equation model underestimates the fuel temperature. → For a conservative assumption, SSG Reynolds stress model was chosen. → Creep strength is the most important parameter in fuel design. - Abstract: The entire nuclear fuel cycle involves partitioning classification and transmutation recycling. The usage of a tokamak as neutron sources to burn spent fuel in a gas cooled subcritical fast reactor (GCSFR) reduces the amount of long-lived radionuclide, thus increasing the repository capacity. This paper presents numerical thermal and fluid dynamics analysis for a gas cooled subcritical fast reactor. The analysis aim to determine the operational flow condition for this reactor, and to compare three distinct turbulence models (Eddy Viscosity Transport Equation, standard k-ε and SSG Reynolds stress) for this application. The model results are presented and discussed. The methodology used in this paper was developed to predict the coolant mass flow rate. It can be applied to any other gas cooled reactor.

  7. Subcritical assemblies, use and their feasibility assessment

    International Nuclear Information System (INIS)

    Haroon, M.R.

    1982-03-01

    In developing countries, subcritical assemblies can be a useful tool for training and research in the field of nuclear technology with minimum cost. The historical development of subcritical assemblies and the reactor physics experiments which can be carried out using this facility are outlined. The different types of subcritical assemblies have been described and material requirements for each assembly have been pointed out. (author)

  8. Noise method for monitoring the sub-criticality in accelerator driven systems

    International Nuclear Information System (INIS)

    Rugama, Y.; Munoz-Cobo, J.L.; Valentine, T.E.; Mihalczo, J.T.; Perez, R.B.; Perez-Navarro, A.

    2001-01-01

    In this paper, an absolute measurements technique for the sub-criticality determination is presented. The development of ADS, requires of methods to monitor and control the sub-criticality of this kind of systems, without interfering it's normal operation mode. This method is based on the Stochastic Neutron and Photon Transport Theory developed by Munoz-Cobo et al., and which can be implemented in presently available neutron transport codes. As a by-product of the methodology a monitoring measurement technique has been developed and verified using two coupled Monte Carlo programs. The spallation collisions and the high-energy transport are simulated with LAHET. The neutrons transports with energies less than 20 MeV and the estimation of the count statistics for neutron and/or gamma ray counters in fissile systems, is simulated with MCNP-DSP. It is possible to get the kinetics parameters and the k eff value of the sub-critical system through the analysis of the counter detectors. (author)

  9. Disposition of nuclear waste using subcritical accelerator-driven systems

    International Nuclear Information System (INIS)

    Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

    1998-01-01

    Studies have shown that the repository long-term radiological risk is from the long-lived transuranics and the fission products Tc-99 and I-129, thermal loading concerns arise mainly form the short-lived fission products Sr-90 and Cs-137. In relation to the disposition of nuclear waste, ATW is expected to accomplish the following: (1) destroy over 99.9% of the actinides; (2) destroy over 99.9% of the Tc and I; (3) separate Sr and Cs (short half-life isotopes); (4) separate uranium; (5) produce electricity. In the ATW concept, spent fuel would be shipped to a ATW site where the plutonium, other transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their only pass through the facility. This approach contrasts with the present-day reprocessing practices in Europe and Japan, during which high purity plutonium is produced and used in the fabrication of fresh mixed-oxide fuel (MOX) that is shipped off-site for use in light water reactors

  10. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Klein, Steven Karl; Day, Christy M.; Determan, John C.

    2015-01-01

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  11. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Day, Christy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  12. Modeling and Simulation Monte Carlo by the MCNP code for determining neutron parameters of the nuclear reactor-subcritical assembly in CNSTN

    International Nuclear Information System (INIS)

    Romdhani, Ibtissem

    2014-01-01

    As part of developing its nuclear infrastructure base, the National Science and Technology Center Nuclear (CNSTN) examines the technical feasibility of setting up a new installation of subcritical assembly. Our study focuses on determining the neutron parameters of a nuclear zero power reactor based on Monte Carlo simulation MCNP. The objective of the simulation is to model the installation, determine the effective multiplication factor, and spatial distribution of neutron flux.

  13. Concept of a subcritical transmutation system with fast neutron spectrum and liquid fuel

    International Nuclear Information System (INIS)

    Tittelbach, S.

    2002-11-01

    The annual amount of nearly 9500 t of spent fuel from worldwide industrial nuclear energy utilization has to be disposed as high level waste. The retention of nuclear waste from the biosphere has to be assured until the radiological risk decreases to tolerable levels. The long-term radiological risk of spent fuel is dominated by actinide elements, i.e. plutonium, americium and curium. It is intended to reduce this amount of high level waste by Partitioning and Transmutation, so that the radiotoxicity of the disposed waste falls short of the reference value of fresh fuel decaying naturally after about thousand years. For this time period the retention of high level waste can be assured by technical means. The scope of this work is the design of a subcritical fast transmutation system with liquid metal cooling and liquid metal fuel. The lead bismuth eutectic has been choosen as the liquid metal coolant and fuel carrier. To dissolve at least 3 at% of transuran elements, a minimum fuel temperature of 600 C is required. The calculations were carried out with a fuel composition, which results from two plutonium recycling steps in a thorium fuel cycle. Two homogeneous and two heterogeneous blankets have been designed and evaluated leading to one preferred heterogeneous blanket design, which has been investigated in more detail. This blanket design merges the positive properties of a solid fuel system (better control of fuel and reactivity because of smaller and closed fuel volumina) and a liquid fuel system (continous charge and discharge or extraction of fission products). The blanket design is based on the core design of fast breeder liquid metal reactors. It consists of hexagonal fuel elements housing up to six annular shaped fuel cylinders. The hexagonal shape of the fuel elements leads to three fuel zones positioned concentrically around the central spallation target. There is a strong heterogeneous distribution of power and heat flux in this blanket design. Besides

  14. Subcritical nuclear assembly

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a {sup 239}PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

  15. Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

    International Nuclear Information System (INIS)

    Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C.

    2017-01-01

    Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S N ) formulation and the response matrix method was applied to solve the forward and the adjoint S N problems. Numerical results are given to verify the present. (author)

  16. Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

    Energy Technology Data Exchange (ETDEWEB)

    Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: lrcmoraes@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: ricardob@iprj.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pós-Graduação em Modelagem Computacional

    2017-07-01

    Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S{sub N}) formulation and the response matrix method was applied to solve the forward and the adjoint S{sub N} problems. Numerical results are given to verify the present. (author)

  17. Dynamic analysis of an accelerator-based subcritical radioactive waste burning system

    International Nuclear Information System (INIS)

    Woosley, M.L. Jr.; Rydin, R.A.

    1997-01-01

    There has been a recent revival of interest in accelerator-driven subcritical fluid-fueled systems for radioactive waste management. This motivates the need for dynamic analysis of the nuclear kinetics of such systems. A physical description of the Los Alamos Accelerator-Based Conversion (ABC) concept is provided. This system is used as the basis for the kinetic study in this research. The current approach to the dynamic simulation of an accelerator-driven subcritical fluid-fueled system includes four functional blocks: A discrete ordinates model is used to calculate the flux distribution for the source-driven system (DORT); A nodal convection model is used to calculate time-dependent isotope and temperature distributions which impact reactivity (ABCcore); A nodal importance weighting model is used to calculate the reactivity impact of temperature and isotope distributions and to feed this information back to the time-dependent nodal convection model (ABCvip); A transient driver simulates system transients and records simulation data (ABCtrans). Specific transients which have been analyzed with the current modeling system are discussed. These transients include loss-of-flow and loss-of-cooling accidents, xenon and samarium transients, and cold-plug and overfueling events. The results of various transients have uncovered unpredictable behavior, unresolved design issues, and the need for active control. 11 refs., 6 figs., 1 tab

  18. Dynamic analysis of an accelerator-driven fluid-fueled subcritical radioactive waste burning system

    International Nuclear Information System (INIS)

    Woosley, M.L. Jr.; Rydin, R.A.

    1998-01-01

    The recent revival of interest in accelerator-driven subcritical fluid-fueled systems is documented. Several important applications of these systems are mentioned, and this is used to motivate the need for dynamic analysis of the nuclear kinetics of such systems. A physical description of the Los alamos National Laboratory accelerator-based conversion (ABC) concept is provided. This system is used as the basis for the kinetics study in this research. The current approach to the dynamic simulation of an accelerator-driven subcritical fluid-fueled system includes four functional elements: a discrete ordinates model is used to calculate the flux distribution for the source-driven system; a nodal convection model is used to calculate time-dependent isotope and temperature distributions that impact reactivity; a nodal importance weighting model is used to calculate the reactivity impact of temperature and isotope distributions and to feed this information back to the time-dependent nodal convection model; and a transient driver is used to simulate transients, model the balance of plant, and record simulation data. Specific transients that have been analyzed with the current modeling system are discussed. These transients include loss-of-flow and loss-of-cooling accidents, xenon and samarium transients, and cold-plug and overfueling events. The results of various transients have uncovered unpredictable behavior, unresolved design issues, and the need for active control. The need for the development of a nodal-coupling spatial kinetics model is mentioned

  19. Design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium; Diseno de un reactor nuclear subcritico heterogeneo con sales fundidas a base de torio

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Hernandez A, P.; Letechipia de L, C.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080-A (Venezuela, Bolivarian Republic of)

    2015-09-15

    This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a {sup 252}Cf source, whose dose levels at the periphery allows its use in teaching and research activities. The design was realized by the Monte Carlo method, where the geometry, dimensions and the fuel was varied in order to obtain the best design. The result was a cubic reactor of 110 cm of side, with graphite moderator and reflector. In the central part having 9 ducts of 3 cm in diameter, eight of them are 110 cm long, which were placed on the Y axis; the separation between each duct is 10 cm. The central duct has 60 cm in length and this contains the {sup 252}Cf source, also there are two irradiation channels and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For the design the k{sub eff} was calculated, neutron spectra and ambient dose equivalent. In the first instance the above was calculated for a virgin fuel, was called case 1; then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose of comparing two different fuels operating within the reactor. For the two irradiation ducts three positions are used: center, back and front, in each duct in order to have different flows. (Author)

  20. Physical aspects of an electronuclear set-up on the basis of the subcritical zone of the IBR-30 reactor and the 660 MeV proton phasotron

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Pogodaev, G.N.; Polanski, A.; Popov, Yu.P.; Puzynin, I.V.; Sisakyan, A.N.; Sosnin, A.N.

    1998-01-01

    Mathematical modeling and thermal flux estimations show that a combination of installations available at present at JINR - the plutonium reactor IBR-30 and the 660 MeV proton phasotron with the current of the extracted beam 0.25 mkA, i.e. 10% of its average value, - allows one to construct an air-cooled electronuclear set-up with the multiplication coefficient K eff ≅ 0.94, the neutron yield N tot ≅ 10 14 - 10 15 and the heat generation about to 10 kW. This set-up will demonstrate a possibility to construct subcritical transmutation-power generating electronuclear systems safe and stable in operation and applicable for utilization of weapon grade and technical plutonium. Kinetics of the electronuclear system will be investigated, in particular, fluctuations of the value of K eff for various parameters of the proton beam. Cross sections of nuclear reactions which are important for the estimations of an efficiency of various conditions of the nuclear waste transmutation and the neutron fluxes together with the heat distributions inside and outside the plutonium core will be measured. A comparison of these data with theoretical calculations allows one to check up and to develop significantly the methods of mathematical modeling of electronuclear systems. We suppose also to estimate the possibilities of ARC-method for the burning of radioactive wastes and to study the influence of various reflectors and multipartitioning of the core which increases the neutron yield. (author)

  1. Decay constants of subcritical system by diffusion theory for two groups

    International Nuclear Information System (INIS)

    Moura Neto, C. de.

    1977-01-01

    The effects of a neutronic pulse applied to a subcritical multiplicative medium are analysed on the basis of the diffusion theory for one and two groups. The decay constants of the system for various values of geometric buckling were determined from the experimental data. A natural uranium-light water lattice was pulsed employing a Texas Nuclear 9905 neutron generator. The least square method was employed in the data reduction procedures to determine the decay constants. The separation of the decay constants associated with thermal and epithermal fluxes is attempted through two groups formulation. (author)

  2. Decay constants of a subcritical system by two-group diffusion theory

    International Nuclear Information System (INIS)

    Moura Neto, C. de.

    1979-08-01

    The effects of a neutronic pulse applied to a subcritical multiplicative medium are analyzed on the basis of the diffusion theory for one and two groups. The decay constants of the system were determined from the experimental data, for various values geometric buckling. A natural uranium light-water configuration was pulsed employing a Texas Nuclear 9905 neutron generator. The least square method was employed in the data reduction procedures to determine the decay constants. The separation of the decay constants associated with thermal and epithermal fluxes are verified through two groups formulation. (Author) [pt

  3. Numerical model for thermoeconomic diagnosis in commercial transcritical/subcritical booster refrigeration systems

    International Nuclear Information System (INIS)

    Ommen, Torben; Elmegaard, Brian

    2012-01-01

    Highlights: ► A transcritical booster refrigeration plant is modelled. ► We examine changes in cost flow at different operation parameters. ► The use of characteristic curves for diagnosis is studied. - Abstract: Transcritical/subcritical booster refrigeration systems are increasingly installed and used in Danish supermarkets. The systems operate in both transcritical and subcritical conditions dependent on the heat rejection performance and the ambient conditions. The plant consists of one refrigerant cycle supplying refrigerant for evaporators in both chilled and frozen display cases. In the paper, thermoeconomic theory is used to establish the cost of cooling at each individual temperature level based on operating costs. With a high amount of operating systems, faulty operation becomes an economic, and environmental, interest. A general solution for evaluation of these systems is considered, with the objective to reduce cost and power consumption of malfunctioning equipment in operation. An analysis of the use of thermoeconomic diagnosis methods is required, as these methods may prove applicable. To accommodate the analysis, a numerical model of a transcritical booster refrigeration plant is considered in this paper. Additionally the characteristic curves method is applied to the high pressure compressor unit of the refrigeration plant. The approach successfully determine whether an anomaly is intrinsic or induced in the component when no uncertainties are introduced in the steady state model.

  4. High order statistical signatures from source-driven measurements of subcritical fissile systems

    International Nuclear Information System (INIS)

    Mattingly, J.K.

    1998-01-01

    This research focuses on the development and application of high order statistical analyses applied to measurements performed with subcritical fissile systems driven by an introduced neutron source. The signatures presented are derived from counting statistics of the introduced source and radiation detectors that observe the response of the fissile system. It is demonstrated that successively higher order counting statistics possess progressively higher sensitivity to reactivity. Consequently, these signatures are more sensitive to changes in the composition, fissile mass, and configuration of the fissile assembly. Furthermore, it is shown that these techniques are capable of distinguishing the response of the fissile system to the introduced source from its response to any internal or inherent sources. This ability combined with the enhanced sensitivity of higher order signatures indicates that these techniques will be of significant utility in a variety of applications. Potential applications include enhanced radiation signature identification of weapons components for nuclear disarmament and safeguards applications and augmented nondestructive analysis of spent nuclear fuel. In general, these techniques expand present capabilities in the analysis of subcritical measurements

  5. Source-jerk method for application on ADS neutronics study The ADS is stated for Accelerator Driven sub-critical System

    CERN Document Server

    Zhu Qing Fu; Li Yi; Xia Pu; Zheng Wu Qing; Zhu Guo Sheng

    2003-01-01

    The paper is concerned in the source-jerk method used to measure the sub-criticality, and the sub-critical experiment facility, which is used for the study on the neutronics of ADS, driven by external neutron source sup 2 sup 5 sup 2 Cf. The effects of the location of neutron source and material buffer where is at the location of the pipe of proton beam and target of fission-product dispersion on the sub-criticality of reactor are studied by source-jerk method

  6. Study of the External Neutron Source Effect on TRU Burning in a Sub-critical Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zafar, Zafar Iqbal; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    One of the drawback points of nuclear power is the production of highly radioactive and long lasting waste isotopes during power production. Therefore, most important design requirement of future nuclear option should have a potential to burn selectively long-lived fission products (LLFP) and long-lived minor actinides (LLMA). However, there is no way to burn them selectively in the reactor core. Practical method of waste transmutation should rely on selective separation of them from spent nuclear fuel of power plants. Under the proliferation concern, direct separation of trans-uranic isotopes (TRU) from pyro-reprocessing plant became a feasible option in our country. Even though social political agreement is not matured as well as technical feasibility, current study is done based on basic assumptions; TRU and LLFP is separated from spent fuel of nuclear power plants. The remaining neutrons (among the external 3%) very few in number (less than 1% in any case) being very energetic (above three MeV or so) do cause much more fissions per neutron than their counterparts but, because of their overall low population they do not have any significant and decisive influence in the overall reactor performance. Currently, entire study is limited to the source neutron energy of 20 MeV only. In future, it is expected to get reasonably plausible fixed source dependent difference in the TRU burning by using tabulated data for the neutrons of higher energy (up to 250 MeV at least). Secondly, a clearer picture is expected if the TRU loading was increased from the current value of 133 kg to few metric tons, as is the case in most of the existing reactors.

  7. Study on variance-to-mean method as subcriticality monitor for accelerator driven system operated with pulse-mode

    International Nuclear Information System (INIS)

    Yamauchi, Hideto; Kitamura, Yasunori; Yamane, Yoshihiro; Misawa, Tsuyoshi; Unesaki, Hironobu

    2003-01-01

    Two types of the variance-to-mean methods for the subcritical system that was driven by the periodic and pulsed neutron source were developed and their experimental examination was performed with the Kyoto University Critical Assembly and a pulsed neutron generator. As a result, it was demonstrated that the prompt neutron decay constant could be measured by these methods. From this fact, it was concluded that the present variance-to-mean methods had potential for being used in the subcriticality monitor for the future accelerator driven system operated with the pulse-mode. (author)

  8. Proposals on the organization of a fuel cycle of the cascade sub-critical molten salt reactor (CSMSR)

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Kormilitsyn, M.V.; Melnik, M.I.; Babikov, L.G.; Ponomarev, L.I.

    2002-01-01

    At present the approach of burning out long-lived radioactive waste (RW) in the reactor core neutron flux is the most feasible one. Currently the way of closing nuclear fuel cycle (NFC) on the basis of the nuclear chemical concept of the cascade sub-critical molten salt reactor (CSMSR) is considered as the most promising one. It is characterised by a number of advantages. CSMSR controlled by a beam of protons or electrons is the optimal reactor for closing the NFC using non-aqueous fluoride methods of fuel reprocessing. They, in comparison with aqueous methods, are characterised by a small waste quantity and are less laborious because of the absence of severe requirements to the product purity. A high productivity of high-temperature electrochemical processes allows the implementation of the fuel recycling process as part of the CSMSR total technological cycle. It can be conducted in the 'on-line' mode in the bypass molten salt circuit that brings the transportation volume of high-activity materials to a minimum. In order to reprocess the CSMSR irradiated molten salt fuel on the basis of salt composition LiF-NaF-(BeF 2 ) an option, based on the following three main operations of the melt treatment, was proposed at SSC RF RIAR: (i) On-line argon treatment of molten salt fuel for removal of gaseous fission products (FP) and also FP that form volatile fluorides and aerosols; (ii) Organisation of the fuel-active metal (probably with a fine-dispersed plutonium alloy) interaction in the on-line mode for removal of 'noble' and 'semi-noble' FP and corrosion products such as Ni, Fe, Cr (when using Pu alloy it allows to regenerate at the same time of the burned-out plutonium component); (iii) Portion-by-portion (fuel composition partially being removed from the CSMSR molten salt circuit) pyroelectrochemical reprocessing of the molten salt composition aimed at the removal of lanthanides - FP followed by a return of actinides to the CSMSR fuel cycle. This technology will allow

  9. Accelerator driven sub-critical core

    Science.gov (United States)

    McIntyre, Peter M; Sattarov, Akhdiyor

    2015-03-17

    Systems and methods for operating an accelerator driven sub-critical core. In one embodiment, a fission power generator includes a sub-critical core and a plurality of proton beam generators. Each of the proton beam generators is configured to concurrently provide a proton beam into a different area of the sub-critical core. Each proton beam scatters neutrons within the sub-critical core. The plurality of proton beam generators provides aggregate power to the sub-critical core, via the proton beams, to scatter neutrons sufficient to initiate fission in the sub-critical core.

  10. Estimation of the sub-criticality of the sodium-cooled fast reactor Monju using the modified neutron source multiplication method

    International Nuclear Information System (INIS)

    Truchet, G.; Van Rooijen, W. F. G.; Shimazu, Y.; Yamaguchi, K.

    2012-01-01

    The Modified Neutron Source Method (MNSM) is applied to the Monju reactor. This static method to estimate sub-criticality has already given good results on commercial Pressurized Water Reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement. Among Monju's particularities that have a big influence on MNSM factors are: the presence of two californium sources and the position of the detector which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and (α, n) reactions which have increased during the shutdown period of 15 years will also be discussed. The relative position of detectors and sources deeply affect the correction factors in some regions. In order to evaluate the detector count rate, an analytical propagation has been conducted from the reactor vessel. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). (authors)

  11. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  12. Optimization by simulation of the coupling between a sub-critical reactor and its spallation source. Towards a pilot reactor; Optimisation par simulation du couplage entre un reacteur sous-critique et sa source de spallation. Application a un demonstrateur

    Energy Technology Data Exchange (ETDEWEB)

    Kerdraon, D

    2001-10-01

    Accelerator Driven Systems (ADS), based on a proton accelerator and a sub-critical core coupled with a spallation target, offer advantages in order to reduce the nuclear waste radiotoxicity before repository closure. Many studies carried out on the ADS should lead to the definition of an experimental plan which would federate the different works in progress. This thesis deals with the neutronic Monte Carlo simulations with the MCNPX code to optimize such a system in view of a pilot reactor building. First, we have recalled the main neutronic properties of an hybrid reactor. The concept of gas-cooled eXperimental Accelerator Driven System (XADS) chosen for our investigations comes from the preliminary studies done by the Framatome company. In order to transmute minor actinides, we have considered the time evolution of the main fuels which could be reasonably used for the demonstration phases. The neutronic parameters of the reactor, concerning minor actinide transmutation, are reported. Also, we have calculated the characteristic times and the transmutation rates in the case of {sup 99}Tc and {sup 129}I isotopes. We have identified some neutronic differences between an experimental and a power ADS according to the infinite multiplication coefficient, the shape factor and the level of flux to extend the demonstrator concept. We have proposed geometric solutions to keep the radial shape factor of a power ADS acceptable. In the last part, beyond the experimental XADS scope, we have examined the possible transition towards an uranium/thorium cycle based on Molten Salt Reactors using a power ADS in order to generate the required {sup 233}U proportion. (author)

  13. Workshop on Subcritical Neutron Production

    International Nuclear Information System (INIS)

    Walter Sadowski; Roald Sagdeev

    2006-01-01

    Executive Summary of the Workshop on Subcritical Neutron Production A workshop on Subcritical Neutron Production was sponsored by the East-West Center of the University of Maryland on October 11-13, 2004. The subject of the workshop was the application of subcritical neutrons to transmutation of actinides. The workshop was attended by members of the fission, accelerator and fusion communities. Papers on the state of development of neutron production by accelerators, fusion devices, and fission reactors were presented. Discussions were held on the potential of these technologies to solve the problems of spent nuclear waste storage and nuclear non-proliferation presented by current and future nuclear power reactors. A list of participants including their affiliation and their E-Mail addresses is attached. The workshop concluded that the technologies, presently available or under development, hold out the exciting possibility of improving the environmental quality and long term energy resources of nuclear power while strengthening proliferation resistance. The workshop participants agreed on the following statements. The workshop considered a number of technologies to deal with spent nuclear fuels and current actinide inventories. The conclusion was reached that substantial increase in nuclear power production will require that the issue of spent nuclear fuel be resolved. The Workshop concluded that 14 MeV fusion neutrons can be used to destroy nuclear reactor by-products, some of which would otherwise have to be stored for geologic periods of time. The production of 14 MeV neutrons is based on existing fusion technologies at different research institutions in several countries around the world. At the present time this technology is used to produce 14 MeV neutrons in JET. More development work will be required, however, to bring fusion technology to the level where it can be used for actinide burning on an industrial scale. The workshop concluded that the potential

  14. Determination of limits for smallest detectable and largest subcritical leakage cracks in piping systems

    International Nuclear Information System (INIS)

    Bieselt, R.; Wolf, M.

    1995-01-01

    Nuclear power plant piping systems - those still in their original as-built condition as well as upgraded designs - are subject to safety analysis. In order to limit the consequences of postulated piping failures, the basic safety concept incorporating rupture preclusion criteria is applied to specific high-energy piping systems. Leak-before-break analyses are also conducted within the framework of this concept. These analyses serve to determine the potential consequences of jet and reaction forces due to maximum subcritical leakage cracks while also establishing the minimum crack sizes that would be reliably detectable by the leakage rates resulting from these cracks. The boundary conditions for these analyses are not clearly defined. Using various examples as a basis, this paper presents and discusses how the leak-before-break concept can be applied. (orig.)

  15. A portable measurement system for subcriticality measurements by the Cf-source-driven neutron noise analysis method

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Ragan, G.E.; Blakeman, E.D.

    1987-01-01

    A portable measurement system consisting of a personal computer used as a Fourier analyzer and three detection channels (with associated electronics that provide the signals to analog-to-digital (A/D) convertors) has been assembled to measure subcriticality by the 252 Cf-source-driven neutron noise analysis method. 8 refs

  16. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  17. Conceptual study of high power proton linac for accelerator driven subcritical nuclear power system

    International Nuclear Information System (INIS)

    Yu Qingchang; Ouyang Huafu; Xu Taoguang

    2002-01-01

    As a prior option of the next generation of energy source, the accelerator driven subcritical nuclear power system (ADS) can use efficiently the uranium and thorium resource, transmute the high-level long-lived radioactive wastes and raise nuclear safety. The ADS accelerator should provide the proton beam with tens megawatts. The superconducting linac is a good selection of ADS accelerator because of its high efficiency and low beam loss rate. The ADS accelerator presented by the authors consists of a 5 MeV radio-frequency quadrupole, a 100 MeV independently phased superconducting cavity linac and a 1 GeV elliptical superconducting cavity linac. The accelerating structures and main parameters are determined and the research and development plan is considered

  18. Conceptual study of high power proton linac for accelerator driven subcritical nuclear power system

    CERN Document Server

    Yu Qi; Ouyang Hua Fu; Xu Tao Guang

    2001-01-01

    As a prior option of the next generation of energy source, the accelerator driven subcritical nuclear power system (ADS) can use efficiently the uranium and thorium resource, transmute the high-level long-lived radioactive wastes and raise nuclear safety. The ADS accelerator should provide the proton beam with tens megawatts. The superconducting linac is a good selection of ADS accelerator because of its high efficiency and low beam loss rate. The ADS accelerator presented by the consists of a 5 MeV radio-frequency quadrupole, a 100 MeV independently phased superconducting cavity linac and a 1 GeV elliptical superconducting cavity linac. The accelerating structures and main parameters are determined and the research and development plan is considered

  19. Equilibria of ternary system Acetic Acid—Water—CO2 under subcritical conditions

    DEFF Research Database (Denmark)

    JIMENEZ GUTIERREZ, Jose M. (Chema); Mussatto, Solange I.; TSOU, Joana

    Carbon dioxide has been subject of research in the past decades, with special attention targeting different uses of this “greenhouse” gas as raw material, technological fluid, building block or as a carbon supply for fuels, turning it from a pollutant to a green resource. Albeit likely...... it will be returned to the atmosphere (as part of the carbon cycle), CO2 is an inexpensive and clean source with numerous industrial applications in diverse fields: from chemical processes to biotechnological purposes [1]. Many of these studies have been focused on supercritical CO2, due to its broad potential uses...... in a very wide range of applications. However, those conditions, especially the levels of high pressure required at larger scale, involve certain equipment limitations. An alternative to overcome those restrictions is to use subcritical carbon dioxide. In order to understand the different systems...

  20. Estimation of subcriticality and fuel concentration by using 'pattern matching' of neutron flux distribution under non uniformed system

    International Nuclear Information System (INIS)

    Ishitani, Kazuki; Yamane, Yoshihiro

    1999-01-01

    In nuclear fuel reprocessing plants, monitoring the spatial profile of neutron flux to infer subcriticality and distribution of fuel concentration using detectors such as PSPC, is very beneficial in sight of criticality safety. In this paper a method of subcriticality and fuel concentration estimation which is supposed to use under non-uniformed system is proposed. Its basic concept is the pattern matching between measured neutron flux distribution and beforehand calculated one. In any kind of subcriticality estimation, we can regard that measured neutron counts put any kind of black box, and then this black box outputs subcriticality. We proposed the use of artificial neural network or 'pattern matching' as black box which have no theoretical clear base. These method are wholly based on the calculated value as recently advancement of computer code accuracy for criticality safety. The most difference between indirect bias estimation method and our method is that our new approach target are the unknown non-uniform system. (J.P.N.)

  1. Obtainment of the subcritical reactivity by mean of measurement of APSD and CPSD employing pulse mode detectors in the IPEN/MB-01 reactor; Obtencao de reatividade subcritica por meio de medidas de APSD e CPSD utilizando detectores modo pulso no Reator IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Min

    2014-07-01

    This work presents a new experimental approach to determine the reactivity levels of subcritical systems. The method employs the subcritical kinetic model developed by Gandini and Salvatores and it is based only on measured quantities such as counting rates of the detectors employed in the experiments and the parameters arising from the least squares fitting of the APSD (Auto Power Spectral Density) and CPSD (Cross Power Spectral Density). Detector efficiencies, quantity required in other procedures such as Neutron Source Multiplication (NSM) method, are not needed in the proposed method. The only hypothesis made in the method was the independence of the effective delayed neutron fraction and the prompt neutron generation time to the subcriticality level of the system. The proposed method was applied to measure the reactivity of several subcritical configurations of the IPEN/MB-01 reactor. Measurements of APSD and CPSD were performed in several degrees of subcriticality (up to around -7000 pcm). The spectral densities data were least squares fitted to get the prompt decay mode (α) and other quantities. Beside the startup source of the facility, an external neutron source of Am-Be was installed near the core in order to improve neutron counting statistics. The final experimental results are of good quality. The proposed experimental method shows clearly that the classical point kinetic theory cannot describe the measured reactivity. Instead, the reactivity inferred from this model follows closely the subcriticality index (ζ) for the source arrangements in the experiment. The agreement of the MCNP5 and GPT-TORT results, both with ENDF/B-VII.0 as the basic nuclear data library, when compared to the corresponding experimental ones was also good. (author)

  2. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  3. Accelerator driven radiation clean nuclear power system conceptual research symposium

    International Nuclear Information System (INIS)

    Zhao Zhixiang

    2000-06-01

    The R and D of ADS (Accelerators Driven Subcritical System) in China introduced. 31 theses are presented. It includes the basic principle of ADS, accelerators, sub-critical reactors, neutron physics, nuclear data, partitioning and transmutation

  4. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  5. Model-based adaptive sliding mode control of the subcritical boiler-turbine system with uncertainties.

    Science.gov (United States)

    Tian, Zhen; Yuan, Jingqi; Xu, Liang; Zhang, Xiang; Wang, Jingcheng

    2018-05-25

    As higher requirements are proposed for the load regulation and efficiency enhancement, the control performance of boiler-turbine systems has become much more important. In this paper, a novel robust control approach is proposed to improve the coordinated control performance for subcritical boiler-turbine units. To capture the key features of the boiler-turbine system, a nonlinear control-oriented model is established and validated with the history operation data of a 300 MW unit. To achieve system linearization and decoupling, an adaptive feedback linearization strategy is proposed, which could asymptotically eliminate the linearization error caused by the model uncertainties. Based on the linearized boiler-turbine system, a second-order sliding mode controller is designed with the super-twisting algorithm. Moreover, the closed-loop system is proved robustly stable with respect to uncertainties and disturbances. Simulation results are presented to illustrate the effectiveness of the proposed control scheme, which achieves excellent tracking performance, strong robustness and chattering reduction. Copyright © 2018. Published by Elsevier Ltd.

  6. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  7. Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2016-04-01

    Full Text Available An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan, a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft–Walton type accelerator, which generates the external neutron source by deuterium–tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

  8. Monte Carlo analysis of the accelerator-driven system at Kyoto University Research Reactor Institute

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Kyeong; Lee, Deok Jung [Nuclear Engineering Division, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Hyun Chul [VHTR Technology Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Pyeon, Cheol Ho [Nuclear Engineering Science Division, Kyoto University Research Reactor Institute, Osaka (Japan); Shin, Ho Cheol [Core and Fuel Analysis Group, Korea Hydro and Nuclear Power Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft-Walton type accelerator, which generates the external neutron source by deuterium-tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

  9. Subcritical multiplication measurements with a BeO reflected 233U uranyl nitrate solution system

    International Nuclear Information System (INIS)

    Job, P.K.; Srinivasan, M.; Nargundkar, V.R.; Chandramoleshwar, K.; Pasupathy, C.S.; Das, S.; Mayankutty, P.C.

    1978-01-01

    A series of subcritical multiplication measurements were carried out in PURNIMA with 233 U uranyl nitrate solution contained in all 11 x 11 cm 2 square sectional tank and reflected by 30 cm thickness of BeO on all sides. The objective of these experiments was to determine the 'Minimum critical mass' of the system in rectangular parellelopiped geometry. The rectangular aluminium core tank was attached to the bottom of an alpha tight glove box. BeO reflector was arranged below the glove box outside the core tank. The system multiplication was measured as a function of solution concentration and core volume by means of neutron detectors placed outside the assembly. The extrapolated critical mass was obtained through conventional inverse counts plot. The maximum amount of 233 U used was 120 gms. The rectangular geometry was estimated to be 235 +- 10 gms, in the concentration range of 80 to 120 gms/litre of 233 U. The experimental set up, procedure adopted, method of analysis and the details of the results are described. (author)

  10. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  11. Reactor safety assessment system

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  12. Design of a molten heavy-metal coolant and target for fast-thermal accelerator driven sub-critical system (ADS)

    International Nuclear Information System (INIS)

    Satyamurthy, P.; Degwekar, S.B.; Nema, P.K.

    2001-01-01

    Accelerator Driven sub-critical Systems (ADS) have evoked considerable interest in recent years. The Energy Amplifier concept developed by C. Rubbia and others at CERN incorporates a buoyancy driven, lead-coolant primary system for extracting the heat generated in the fast reactor as well as that in neutron spallation target. In earlier publications, our BARC group has proposed a one-way coupled booster reactor system which could be operated at proton beam currents as low as 1-2 mA for a power output of 750 MW th . Here, the basic idea is to have a fast booster reactor zone of low power (- 100 MW th ) which is separated by a large gap from the main thermal reactor zone. In this arrangement, the spallation neutron source feeds neutrons to the fast reactor zone where neutrons are further multiplied. Further in this system, the neutrons from the booster region enter the main reactor but very few neutrons from main reactor return to booster, thus ensuring one-way coupling. In earlier work, several possible configurations of the booster and thermal regions were presented. In the present work, we describe an engineering design particularly with respect to thermal hydraulics of lead/lead-bismuth eutectic coolant also acting as spallation neutron source. This hybrid ADS reactor consists of fast and thermal reactor zones producing about 100 MW th and 650 MW th respectively. The scheme of the system is shown. The fast core consists of 48 hexagonal fuel bundles each containing 169 fuel pins of 8.2 mm diameter arranged in 11.4 mm triangular array pitch. The average thermal power per fuel pin is about 13.46 kw. However, due to neutron flux peaking effect, the maximum fuel pin power can be up to 2.5 times this average power. The thermal reactor consists of heavy water as moderator and coolant similar to a typical CANDU type Indian PHWR except for fuel composition. Though the gap between fast and thermal zones essentially provides one way coupling of neutron flux, a thermal

  13. New breathing functions for the transverse breathing crack of the cracked rotor system: Approach for critical and subcritical harmonic analysis

    Science.gov (United States)

    Al-Shudeifat, Mohammad A.; Butcher, Eric A.

    2011-01-01

    The actual breathing mechanism of the transverse breathing crack in the cracked rotor system that appears due to the shaft weight is addressed here. As a result, the correct time-varying area moments of inertia for the cracked element cross-section during shaft rotation are also determined. Hence, two new breathing functions are identified to represent the actual breathing effect on the cracked element stiffness matrix. The new breathing functions are used in formulating the time-varying finite element stiffness matrix of the cracked element. The finite element equations of motion are then formulated for the cracked rotor system and solved via harmonic balance method for response, whirl orbits and the shift in the critical and subcritical speeds. The analytical results of this approach are compared with some previously published results obtained using approximate formulas for the breathing mechanism. The comparison shows that the previously used breathing function is a weak model for the breathing mechanism in the cracked rotor even for small crack depths. The new breathing functions give more accurate results for the dynamic behavior of the cracked rotor system for a wide range of the crack depths. The current approach is found to be efficient for crack detection since the critical and subcritical shaft speeds, the unique vibration signature in the neighborhood of the subcritical speeds and the sensitivity to the unbalance force direction all together can be utilized to detect the breathing crack before further damage occurs.

  14. Physics design of an accelerator for an accelerator-driven subcritical system

    Directory of Open Access Journals (Sweden)

    Zhihui Li

    2013-08-01

    Full Text Available An accelerator-driven subcritical system (ADS program was launched in China in 2011, which aims to design and build an ADS demonstration facility with the capability of more than 1000 MW thermal power in multiple phases lasting about 20 years. The driver linac is defined to be 1.5 GeV in energy, 10 mA in current and in cw operation mode. To meet the extremely high reliability and availability, the linac is designed with much installed margin and fault tolerance, including hot-spare injectors and local compensation method for key element failures. The accelerator complex consists of two parallel 10-MeV injectors, a joint medium-energy beam transport line, a main linac, and a high-energy beam transport line. The superconducting acceleration structures are employed except for the radio frequency quadrupole accelerators (RFQs which are at room temperature. The general design considerations and the beam dynamics design of the driver linac complex are presented here.

  15. FFTF reactor assembly system technology

    International Nuclear Information System (INIS)

    Mangelsdorf, T.A.

    1975-01-01

    An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

  16. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  17. A portable measurement system for subcriticality measurements by the CF-source-driven neutron noise analysis method

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Ragan, G.E.; Blakeman, E.D.

    1988-01-01

    A portable measurement system consisting of a personal computer used as a Fourier analyzer and three detection channels (with associated electronics that provide the signals to analog-to-digital (A/D) convertors) has been assembled to measure subcriticality by the /sup 252/Cf-source-driven neutron noise analysis method. The /sup 252/Cf-source-driven neutron noise analysis method for obtaining the subcritical neutron multiplication factor of a configuration of fissile material requires measurement of the frequency-dependent cross-power spectral density (CPSD), G/sub 23/(ω), between a pair of detectors (Nos. 2 and 3) located in or near the fissile material and CPSDs G/sub 12/(ω) and G/sub 13/(ω) between these same detectors and a source of neutrons emanating from an ionization chamber (No. 1) containing /sup 252/Cf, also positioned in or near the fissile material. The auto-power spectral density (APSD), G/sub 11/(ω), of the source is also required. A particular ratio of spectral densities, G/sub 12//sup */G/sub 13//G/sub 11/G/sub 23/ (/sup */ denotes complex conjugation), is then formed. This ratio is related to the subcritical neutron multiplication factor and is independent of detector efficiencies

  18. Nuclear reactor monitoring system

    International Nuclear Information System (INIS)

    Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

    1976-01-01

    The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

  19. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  20. The Muse-4 experiment: measurement of the kinetic parameters of a subcritical system; L'experience MUSE-4: mesure des parametres cinetiques d'un systeme sous-critique

    Energy Technology Data Exchange (ETDEWEB)

    Vollaire, J

    2004-10-01

    Accelerator Driven Systems (ADS) which are based on an external neutron source coupled to a subcritical core, offer advantages for the incineration of radioactive waste. In order to understand the neutronic specificity of such a system, during the MUSE IV experimental program, the experimental reactor MASURCA (CEA Cadarache) has been coupled to the neutrons source GENEPI. This setup has enabled the development of an on-line measurement technique of the effective multiplication factor. This measurement benefits from the characteristics of the reactor response depending on the multiplication factor at the prompt fission and delayed fission time scales. The analysis of those experiments shows that the proposed method give results in agreement with the one deduced using classical reactivity measurement techniques which can not however be used in a power ADS. (author)

  1. 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)

  2. Reactor protection system

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

    1977-10-01

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  3. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

    This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

  4. Impact of proton beam trips and pulsation on accelerator-driven subcritical system (ADSS) control

    International Nuclear Information System (INIS)

    Sullivan, P.; Rydin, R.A.

    2001-01-01

    The full text follows. ADSS (accelerator driven systems) concepts use a source consisting of a proton beam directed into a high-Z target, such as tungsten or lead, driving a (p,n) spallation reaction. This results in a neutron source which is meant to sustain a constant rate of fission power production in the fuel. However, conceptual ADSS studies to date have not taken two special attributes of these sources into account, usually using the simplifying assumption that the spallation source is constant and continuously variable. These are: (1) proton beams under consideration for ADSS sources are pulsed at a certain frequency, which would suggest that the neutron source for the ADSS is also pulsed since the spallation reaction is prompt; (2) while shutting off the beam is the usual answer to safety questions, this poses its own set of issues. Even state-of the-art proton beams are prone to routine and frequent trips, with beam restart times ranging from seconds to hours. Both of these attributes could place system components under severe thermal stress. The strong, subcriticality-level-dependent feedback effects that occur in an ADSS can introduce transient power swings and oscillations that may need to be controlled or abated by source modulation and/or control rod motion. Earlier work indicates that such feedback effects may lead to unpredictable behavior, affecting restart performance and requiring active control measures to prevent or mitigate such effects. And there is a further consideration. Since efficient operation of ADSS-based systems may suggest that a system operates near, though below, critical, the actual operation of such a system means that the operator would need to walk a fine line between safety and efficiency. The intent of this effort is to examine control and safety issues posed by pulsed and trip-prone neutron sources in near-critical ADSS's, using spread sheet-based simulations [informed by the results of earlier work] to develop scenarios

  5. Criticality Analysis of the U-H2O Subcritical Assembly Modified for Rand D of the High Temperature Reactor

    International Nuclear Information System (INIS)

    Syarip; Tri-Wulan-Tjiptono; Tegas-Sutondo

    2000-01-01

    A criticality analysis of the natural uranium - light water sub-criticalassembly available at the P3TM-BATAN Yogyakarta, converted into a naturaluranium - graphite system has been performed. The purpose of this study is toprovide the research facility on the basic static and kinetics studies forthe high temperature reactor (HTR) in which the HTR fuel system is underdevelopment at the P3TM. For the purpose of this study, a neutroniccalculation was performed using WIMSD/4 code, to determine the neutronmultiplication factor for various fuel configurations of the sub-criticalassemblies. The results show that the effective neutron multiplication factor(k ef ) for U-Be-H 2 O and U-Be-He systems are 1.0474 and 1.4666 respectively,while for the graphite moderated systems with coolants of H 2 O or He(U-C-H 2 O and U-C-He) systems, the corresponding k ef are 0.787 and 0.4211respectively. The results conclude that the modification of U-H 2 O toU-C-H 2 O system, in accordance with neutronic is quite feasible, safe, cheapand practical, and in addition, the treatment of H 2 O is relatively easy.(author)

  6. Neutronic design and analysis on dual-cooled waste transmutation blanket for the fusion driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican; Gao Chunjing; Xu Dezheng; Li Jingjing; Zhu Xiaoxiang

    2004-01-01

    Neutronics design and analysis of dual-cooled multi-functional waste transmutation blanket (DWTB) for the fusion driven sub-critical system (FDS) are performed to ensure the system be able to meet the requirements of fuel-sufficiency and more waste transmutation ratio with low initial loading fuel inventory, which is based on 1-D burn-up calculations with home-developed code Visual BUS and the multi-group (175 neutron groups-42 Gamma groups coupled) data library HENDL1.0/MG (Hybrid Evaluated Nuclear Data Library). (authors)

  7. Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassanzadeh, M. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Feghhi, S.A.H., E-mail: a_feghhi@sbu.ac.ir [Department of Radiation Application, Shahid Beheshti University, G.C., Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of)

    2013-09-15

    Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ{sup †}) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ{sup †}) and importance weighted neutron generation time (Λ{sup †}) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ{sup †} is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.

  8. Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

    International Nuclear Information System (INIS)

    Hassanzadeh, M.; Feghhi, S.A.H.; Khalafi, H.

    2013-01-01

    Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ † ) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ † ) and importance weighted neutron generation time (Λ † ) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ † is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function

  9. Theoretical Analysis for Heat Transfer Optimization in Subcritical Electrothermal Energy Storage Systems

    Directory of Open Access Journals (Sweden)

    Peng Hu

    2017-02-01

    Full Text Available Electrothermal energy storage (ETES provides bulk electricity storage based on heat pump and heat engine technologies. A subcritical ETES is described in this paper. Based on the extremum principle of entransy dissipation, a geometry model is developed for heat transfer optimization for subcritical ETES. The exergy during the heat transfer process is deduced in terms of entropy production. The geometry model is validated by the extremum principle of entropy production. The theoretical analysis results show that the extremum principle of entransy dissipation is an effective criterion for the optimization, and the optimum heat transfer for different cases with the same mass flux or pressure has been discussed. The optimum heat transfer can be achieved by adjusting the mass flux and pressure of the working fluid. It also reveals that with the increase of mass flux, there is a minimum exergy in the range under consideration, and the exergy decreases with the increase of the pressure.

  10. Nuclear reactor refueling system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

  11. Development and Validation of NODAL-LAMBDA Program for the Calculation of the Sub-criticality of LAMDA MODES By Nodal Methods in BWR reactors

    International Nuclear Information System (INIS)

    Munoz-Cobo, J. L.; Merino, R.; Escriva, A.; Melara, J.; Concejal, A.

    2014-01-01

    We have developed a 3D code with two energy groups and diffusion theory that is capable of calculating eigenvalues lambda of a BWR reactor using nodal methods and boundary conditions that calculates ALBEDO NODAL-LAMBDA from the properties of the reflector code itself. The code calculates the sub-criticality of the first harmonic, which is involved in the stability against oscillations reactor out of phase, and which is needed for calculating the decay rate for data out of phase oscillations. The code is very fast and in a few seconds is able to make a calculation of the first eigenvalues and eigenvectors, discretized solving the problem with different matrix elements zero. The code uses the LAPACK and ARPACK libraries. It was necessary to modify the LAPACK library to perform various operations with five non-diagonal matrices simultaneously in order to reduce the number of calls to bookstores and simplify the procedure for calculating the matrices in compressed format CSR. The code is validated by comparing it with the results for SIMULATE different cases and making 3D BENCHMAR of the IAEA. (Author)

  12. Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium; Calculo de espectros de neutrones y dosis en un reactor nuclear subcritico a base de Torio

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080A (Venezuela, Bolivarian Republic of)

    2015-10-15

    This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of {sup 252}Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of {sup 252}Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For design the k{sub eff}, neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k{sub eff} of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)

  13. Nuclear reactor trip system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out four'' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the ''two out of four''configuration would be reduced to a ''one out of three'' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a ''two out of three'' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

  14. Point kinetics equations for subcritical systems based on the importance function associated to an external neutron source

    International Nuclear Information System (INIS)

    Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

    2015-01-01

    Highlights: • We define the new function importance. • We calculate the kinetic parameters Λ, β, Γ and Q to: 0.95, 0.96, 0.97, 0.98 and 0.99. • We compared the results with those obtained by the main important functions. • We found that the calculated kinetic parameters are physically consistent. - Abstract: This paper aims to determine the parameters for a new set of equations of point kinetic subcritical systems, based on the concept of importance of Heuristic Generalized Perturbation Theory (HGPT). The importance function defined here is related to both the subcriticality and the external neutron source worth (which keeps the system at steady state). The kinetic parameters defined in this work are compared with the corresponding parameters when adopting the importance functions proposed by Gandini and Salvatores (2002), Dulla et al. (2006) and Nishihara et al. (2003). Furthermore, the point kinetics equations developed here are solved for two different transients, considering the parameters obtained with different importance functions. The results collected show that there is a similar behavior of the solution of the point kinetics equations, when used with the parameters obtained by the importance functions proposed by Gandini and Salvatores (2002) and Dulla et al. (2006), specially near the criticality. However, this is not verified as the system gets farther from criticality

  15. Reactor system safety assurance

    International Nuclear Information System (INIS)

    Mattson, R.J.

    1984-01-01

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  16. PNS and statistical experiments simulation in subcritical systems using Monte-Carlo method on example of Yalina-Thermal assembly

    International Nuclear Information System (INIS)

    Sadovich, S.; Burnos, V.; Kiyavitskaya, H.; Fokov, Y.; Talamo, A.

    2013-01-01

    In subcritical systems driven by an external neutron source, the experimental methods based on pulsed neutron source (PNS) and statistical techniques play an important role for reactivity measurement. Simulation of these methods is very time-consumed procedure. For simulations in Monte-Carlo programs several improvements for neutronic calculations have been made. This paper introduces a new method for simulating PNS and statistical measurements. In this method all events occurred in the detector during simulation are stored in a file using PTRAC feature in the MCNP. After that with a special code (or post-processing) PNS and statistical methods can be simulated. Additionally different shapes of neutron pulses and its lengths as well as dead time of detectors can be included into the simulation. The methods described above have been tested on the sub-critical assembly Yalina-Thermal, located in the Joint Institute for Power and Nuclear Research SOSNY in Minsk (Belarus). A good agreement between experiment and simulation was shown. (authors)

  17. BWR reactor management system

    International Nuclear Information System (INIS)

    Makino, Kakuji; Kawamura, Atsuo; Yoshioka, Ritsuo; Neda, Toshikatsu.

    1979-01-01

    It is necessary to grasp the delicate state of operation in reactor cores in view of the control of burn-up and power output at the time of the operation management of BWRs. Enormous labor has been required for the collection, processing and evaluation of the data. It is desirable to obtain the safer, more efficient and faster method of operation control by predicting the states in cores including the change of xenon and reflecting them to operation plans as well as by tracing with high accuracy the past burn-up history for a long period. At present, the on-line evaluation of the states in cores is carried out with the process computers attached to respective units, but the amount of data required for core operation management of high degree far exceeds their capacity. From such viewpoints, the research and development on the reactor management system were carried out. The data processing concerning core operation management is performed with newly installed computers utilizing the data from existing process computers, and the operation of reactor cores, the qualitative improvement of management works, labor saving, and fast, efficient operation control are feasible with it. This system was installed in an actual plant in October, 1977. The composition of the system, the prediction of the change in local output distribution accompanying control rod operation, the prediction of the change in the states in cores due to the flow rate of coolant, and the function of collecting plant data are explained. (Kako, I.)

  18. Nuclear reactor sealing system

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

  19. Thermodynamic performance analysis of a coupled transcritical and subcritical organic Rankine cycle system for waste heat recovery

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Xi Wu [Zhejiang Ocean University, Zhejian (China); Wang, Xiao Qiong; Li, You Rong; Wu, Chun Mei [Chongqing University, Chongqing (China)

    2015-07-15

    We present a novel coupled organic Rankine cycle (CORC) system driven by the low-grade waste heat, which couples a transcritical organic Rankine cycle with a subcritical organic Rankine cycle. Based on classical thermodynamic theory, a detailed performance analysis on the novel CORC system was performed. The results show that the pressure ratio of the expander is decreased in the CORC and the selection of the working fluids becomes more flexible and abundant. With the increase of the pinch point temperature difference of the internal heat exchanger, the net power output and thermal efficiency of the CORC all decrease. With the increase of the critical temperature of the working fluid, the system performance of the CORC is improved. The net power output and thermal efficiency of the CORC with isentropic working fluids are higher than those with dry working fluids.

  20. Application of Hastelloy X in gas-cooled reactor systems

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given

  1. Passive safety system of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

    2015-08-15

    Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

  2. Status of intense permanent magnet proton source for China-accelerator driven sub-critical system Linac.

    Science.gov (United States)

    Wu, Q; Ma, H Y; Yang, Y; Sun, L T; Zhang, X Z; Zhang, Z M; Zhao, H Y; He, Y; Zhao, H W

    2016-02-01

    Two compact intense 2.45 GHz permanent magnet proton sources and their corresponding low energy beam transport (LEBT) system were developed successfully for China accelerator driven sub-critical system in 2014. Both the proton sources operate at 35 kV potential. The beams extracted from the ion source are transported by the LEBT, which is composed of two identical solenoids, to the 2.1 MeV Radio-Frequency Quadrupole (RFQ). In order to ensure the safety of the superconducting cavities during commissioning, an electrostatic-chopper has been designed and installed in the LEBT line that can chop the continuous wave beam into a pulsed one. The minimum width of the pulse is less than 10 μs and the fall/rise time of the chopper is about 20 ns. The performance of the proton source and the LEBT, such as beam current, beam profile, emittance and the impact to RFQ injection will be presented.

  3. Status of intense permanent magnet proton source for China-accelerator driven sub-critical system Linac

    Science.gov (United States)

    Wu, Q.; Ma, H. Y.; Yang, Y.; Sun, L. T.; Zhang, X. Z.; Zhang, Z. M.; Zhao, H. Y.; He, Y.; Zhao, H. W.

    2016-02-01

    Two compact intense 2.45 GHz permanent magnet proton sources and their corresponding low energy beam transport (LEBT) system were developed successfully for China accelerator driven sub-critical system in 2014. Both the proton sources operate at 35 kV potential. The beams extracted from the ion source are transported by the LEBT, which is composed of two identical solenoids, to the 2.1 MeV Radio-Frequency Quadrupole (RFQ). In order to ensure the safety of the superconducting cavities during commissioning, an electrostatic-chopper has been designed and installed in the LEBT line that can chop the continuous wave beam into a pulsed one. The minimum width of the pulse is less than 10 μs and the fall/rise time of the chopper is about 20 ns. The performance of the proton source and the LEBT, such as beam current, beam profile, emittance and the impact to RFQ injection will be presented.

  4. Status of intense permanent magnet proton source for China-accelerator driven sub-critical system Linac

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Q., E-mail: wuq@impcas.ac.cn; Ma, H. Y.; Yang, Y.; Sun, L. T.; Zhang, X. Z.; Zhang, Z. M.; Zhao, H. Y.; He, Y.; Zhao, H. W. [Institute of Modern Physics (IMP), Chinese Academy of Sciences, Lanzhou 730000 (China)

    2016-02-15

    Two compact intense 2.45 GHz permanent magnet proton sources and their corresponding low energy beam transport (LEBT) system were developed successfully for China accelerator driven sub-critical system in 2014. Both the proton sources operate at 35 kV potential. The beams extracted from the ion source are transported by the LEBT, which is composed of two identical solenoids, to the 2.1 MeV Radio-Frequency Quadrupole (RFQ). In order to ensure the safety of the superconducting cavities during commissioning, an electrostatic-chopper has been designed and installed in the LEBT line that can chop the continuous wave beam into a pulsed one. The minimum width of the pulse is less than 10 μs and the fall/rise time of the chopper is about 20 ns. The performance of the proton source and the LEBT, such as beam current, beam profile, emittance and the impact to RFQ injection will be presented.

  5. Subcritical and supercritical water oxidation of organic, wet wastes for carbon cycling in regenerative life support systems

    Science.gov (United States)

    Ronsse, Frederik; Lasseur, Christophe; Rebeyre, Pierre; Clauwaert, Peter; Luther, Amanda; Rabaey, Korneel; Zhang, Dong Dong; López Barreiro, Diego; Prins, Wolter; Brilman, Wim

    2016-07-01

    For long-term human spaceflight missions, one of the major requirements is the regenerative life support system which has to be capable of recycling carbon, nutrients and water from both solid and liquid wastes generated by the crew and by the local production of food through living organisms (higher plants, fungi, algae, bacteria, …). The European Space Agency's Life Support System, envisioned by the MELiSSA project, consists of a 5 compartment artificial ecosystem, in which the waste receiving compartment (so-called compartment I or briefly 'CI') is based on thermophilic fermentation. However, as the waste generated by the crew compartment and food production compartment contain typical plant fibres (lignin, cellulose and hemicellulose), these recalcitrant fibres end up largely unaffected in the digestate (sludge) generated in the C-I compartment. Therefore, the C-I compartment has to be supplemented with a so-called fibre degradation unit (in short, FDU) for further oxidation or degradation of said plant fibres. A potential solution to degrading these plant fibres and other recalcitrant organics is their oxidation, by means of subcritical or supercritical water, into reusable CO2 while retaining the nutrients in an organic-free liquid effluent. By taking advantage of the altered physicochemical properties of water above or near its critical point (647 K, 22.1 MPa) - including increased solubility of non-polar compounds and oxygen, ion product and diffusivity - process conditions can be created for rapid oxidation of C into CO2. In this research, the oxidizer is provided as a hydrogen peroxide solution which, at elevated temperature, will dissociated into O2. The purpose of this study is to identify ideal process conditions which (a) ensure complete oxidation of carbon, (b) retaining the nutrients other than C in the liquid effluent and (c) require as little oxidizer as possible. Experiments were conducted on a continuous, tubular heated reactor and on batch

  6. Hysteretic and intermittent regimes in the subcritical bifurcation of a quasi-one-dimensional system of interacting particles

    Science.gov (United States)

    Dessup, Tommy; Coste, Christophe; Saint Jean, Michel

    2016-01-01

    In this article, we study the effects of white Gaussian additive thermal noise on a subcritical pitchfork bifurcation. We consider a quasi-one-dimensional system of particles that are transversally confined, with short-range (non-Coulombic) interactions and periodic boundary conditions in the longitudinal direction. In such systems, there is a structural transition from a linear order to a staggered row, called the zigzag transition. There is a finite range of transverse confinement stiffnesses for which the stable configuration at zero temperature is a localized zigzag pattern surrounded by aligned particles, which evidences the subcriticality of the bifurcation. We show that these configurations remain stable for a wide temperature range. At zero temperature, the transition between a straight line and such localized zigzag patterns is hysteretic. We have studied the influence of thermal noise on the hysteresis loop. Its description is more difficult than at T =0 K since thermally activated jumps between the two configurations always occur and the system cannot stay forever in a unique metastable state. Two different regimes have to be considered according to the temperature value with respect to a critical temperature Tc(τobs) that depends on the observation time τobs. An hysteresis loop is still observed at low temperature, with a width that decreases as the temperature increases toward Tc(τobs) . In contrast, for T >Tc(τobs) the memory of the initial condition is lost by stochastic jumps between the configurations. The study of the mean residence times in each configuration gives a unique opportunity to precisely determine the barrier height that separates the two configurations, without knowing the complete energy landscape of this many-body system. We also show how to reconstruct the hysteresis loop that would exist at T =0 K from high-temperature simulations.

  7. Reactor safety systems

    International Nuclear Information System (INIS)

    Kafka, P.

    1975-01-01

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de

  8. Design of a homogeneous subcritical nuclear reactor based on thorium with a source of californium 252; Diseno de un reactor nuclear subcritico homogeneo a base de Torio con una fuente de Californio 252

    Energy Technology Data Exchange (ETDEWEB)

    Delgado H, C. E.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: ce_delgado89@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of)

    2015-10-15

    Full text: One of the energy alternatives to fossil fuels which do not produce greenhouse gases is the nuclear energy. One of the drawbacks of this alternative is the generation of radioactive wastes of long half-life and its relation to the generation of nuclear materials to produce weapons of mass destruction. An option to these drawbacks of nuclear energy is to use Thorium as part of the nuclear fuel which it becomes in U{sup 233} when capturing neutrons, that is a fissile material. In this paper Monte Carlo methods were used to design a homogeneous subcritical reactor based on thorium. As neutron reflector graphite was used. The reactor core is homogeneous and is formed of 70% light water as moderator, 12% of enriched uranium UO{sub 2}(NO{sub 3}){sub 4} and 18% of thorium Th(NO{sub 3}){sub 4} as fuel. To start the nuclear fission chain reaction an isotopic source of californium 252 was used with an intensity of 4.6 x 10{sup 7} s{sup -1}. In the design the value of the effective multiplication factor, whose value turned out k{sub eff} <1 was calculated. Also, the neutron spectra at different distances from the source and the total fluence were calculated, as well as the values of the ambient dose equivalent in the periphery of the reactor. (Author)

  9. Reactor cooling system

    International Nuclear Information System (INIS)

    Kato, Etsuji.

    1979-01-01

    Purpose: To eliminate cleaning steps in the pipelines upon reactor shut-down by connecting a filtrating and desalting device to the cooling system to thereby always clean up the water in the pipelines. Constitution: A filtrating and desalting device is connected to the pipelines in the cooling system by way of drain valves and a check valve. Desalted water is taken out from the exit of the filtrating and desalting device and injected to one end of the cooling system pipelines by way of the drain valve and the check valve and then returned by way of another drain valve to the desalting device. Water in the pipelines is thus always desalted and the cleaning step in the pipelines is no more required in the shut-down. (Kawakami, Y.)

  10. Design and producing of fine-group cross section library HENDL3.0/FG for subcritical system

    International Nuclear Information System (INIS)

    Zou, J.; Zeng, Q.; Xu, D.; Hu, L.; Long, P.

    2012-01-01

    To improve the accuracy of the neutron analyses for subcritical system with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-VII, JEFF3.1 and JENDL3.3 was produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. It showed that the nuclear data library had accuracy and availability. (authors)

  11. Reactor system on barge

    International Nuclear Information System (INIS)

    Hayashi, Kingo; Yamada, Nobuyuki

    1987-01-01

    Floating electrical power plants or power plant barges add new dimensions to utility planners and agencies in the world. Intrinsically safe and economical reactors (ISER) employ steel reactor pressure vessels, which significantly reduce the weight as compared with PIUS, and provide siting versatility including barge-mounted plants. In this paper, the outline of power plant barges and barge-mounted ISERs is described. Besides their mobility, power plant barges have the salient advantages such as short delivery time and better quality control due to the outfitting in shipyards. These power plant barges may be temporarily moored or permanently grounded in shallow water at the centers of industrial complexes or the suitable areas adjacent to them, and satisfy the increasing needs for electric power. A cost-effective and technically perfect barge positioning system should be designed to meet the specific requirement for the location and its condition. Offshore siting away from coast may be applicable only to large plants of 1,000 MWe or more, and inshore siting and coastal or river siting are considered for an ISER-200 barge-mounted plant. The system of a barge-mounted ISER plant is discussed in the case of a floating type and the type on a seismic base isolator. (Kako, I.)

  12. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1979-01-01

    Purpose: To allow sufficient removal of radioactive substance released in the reactor containment shell upon loss of coolants accidents thus to sufficiently decrease the exposure dose to human body. Constitution: A clean-up system is provided downstream of a heat exchanger and it is branched into a pipeway to be connected to a spray nozzle and further connected by way of a valve to a reactor container. After the end of sudden transient changes upon loss of coolants accidents, the pool water stored in the pressure suppression chamber is purified in the clean-up system and then sprayed in the dry-well by way of a spray nozzle. The sprayed water dissolves to remove water soluble radioactive substances floating in the dry-well and then returns to the pressure suppression chamber. Since radioactive substances in the dry-well can thus removed rapidly and effectively and the pool water can be reused, public hazard can also be decreased. (Horiuchi, T.)

  13. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

    Juhl, N.H.; Marwick, E.F.

    1983-01-01

    A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

  14. Feasibility analysis of constant TRU feeding in waste transmutation system using accelerator-driven subcritical system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kun Jai; Cho, Nam Zin; Jo, Chang Keun; Park, Chang Je; Kim, Do Sam; Park, Jeong Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    It is probable that the issue of nuclear spent fuel and high-level waste can have negative impact on the future expansion of nuclear power programs. Accelerator-driven nuclear waste transmutation with constant composition TRU feeding which satisfies non-proliferation condition will help establish the long-range nuclear waste disposal strategy. In this study, current status of accelerator-driven transmutation of waste technology, and feasibility analysis of constant composition TRU feeding system were investigated. We ascertained that solid system using constant composition TRU is feasible with the the capability of transmutation. (author). 13 refs., 53 figs., 20 tabs.

  15. Space-time dependent impulse response of a subcritical cylindrical reactor; Reponse impulsionnelle spatio-temporelle d'un reacteur cylindrique en regime sous-critique

    Energy Technology Data Exchange (ETDEWEB)

    Cazemajou, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    In this paper, a new formulation of the spatial dependent impulse response of a subcritical reactor in a cylindrical geometry is proposed. An expression of the transfer function between a point source at the center of coordinates and the flux at a given point (r,z) is obtained by solving: by means of Laplace transform, the one group diffusion equation. In this transfer function, variables r and p (p being the Laplace variable) remain linked within a modified Bessel function. Taking the inverse Laplace transform is done by two different ways: - using the Mellin-Fourier method which separates variables r and t. This method makes it possible to establish that there is identity between the classical formulation and the new one. - using an inverse Laplace transform which keeps variables r and t linked. This method requires to approximate the inverse Laplace transform of the end factor. It is then possible to replace the radial harmonics modes series of the classical expression by a single function. This new formulation seems to be of particular interest when dealing with reactors of large size and lifetime. It is also interesting each time the harmonics play an important role. (author) [French] Dans le present rapport, on propose une nouvelle formulation de la reponse impulsionnelle spatio-temporelle d'un reacteur sous-critique, en geometrie cylindrique. Une expression de la fonction de transfert entre une source ponctuelle placee au centre des coordonnees et le flux au point courant (r,z) est obtenue en resolvant, par transformation de Laplace, l'equation de la diffusion a un seul groupe d'energie. Dans cette fonction de transfert, les variables r et p (variable de Laplace) demeurent groupees dans une fonction de Bessel modifiee. Le retour a l'original est effectue de deux manieres: - la methode de Mellin-Fourier qui separe les variables r et t, permet d'etablir l'identite entre la nouvelle formulation et la formulation classique. - un original conservant les variables

  16. Modelling of two-zone accelerator-driven systems

    Directory of Open Access Journals (Sweden)

    V. A. Babenko

    2012-09-01

    Full Text Available Neutron-physical modelings of two-zone subcritical reactor driven by high-intensity neutron generator are considered. The cascade principle in subcritical reactors, the use of which can hypothetically substantially amplify the neutron flux from the external source is discussed in this article. The theoretical preconditions of the cascade principle are discussed, and the directions of practical realization of the cascade subcritical system are considered, namely the possible methods of neutron feedback between reactor sections elimination. The results of Monte Carlo neutron-physical modeling of the cascade subcritical systems are presented and discussed.

  17. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  18. Monitor for reactor feedwater systems

    International Nuclear Information System (INIS)

    Takizawa, Yoji; Tomizawa, Teruaki

    1983-01-01

    Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

  19. Dynamic Simulation of the Water-steam System in Once-through Boilers - Sub-critical Power Boiler Case -

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seongil; Choi, Sangmin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2017-05-15

    The dynamics of a water-steam system in a once-through boiler was simulated based on the physics-based modeling approach, representing the system in response to large load change or scale disturbance simulations. The modeling considered the mass, energy conservation, and momentum equation in the water pipe and the focus was limited to the sub-critical pressure region. An evaporator tube modeling was validated against the reference data. A simplified boiler system consisting of economizer, evaporator, and superheater was constructed to match a 500 MW power boiler. The dynamic response of the system following a disturbance was discussed along with the quantitative response characteristics. The dynamic response of the boiler system was further evaluated by checking the case of an off-design point operation of the feedwater-to-fuel supply ratio. The results re-emphasized the significance of controlling the feedwater-to-fuel supply ratio and additional design requirements of the water-steam separator and spray attemperator.

  20. Dynamic Simulation of the Water-steam System in Once-through Boilers - Sub-critical Power Boiler Case -

    International Nuclear Information System (INIS)

    Kim, Seongil; Choi, Sangmin

    2017-01-01

    The dynamics of a water-steam system in a once-through boiler was simulated based on the physics-based modeling approach, representing the system in response to large load change or scale disturbance simulations. The modeling considered the mass, energy conservation, and momentum equation in the water pipe and the focus was limited to the sub-critical pressure region. An evaporator tube modeling was validated against the reference data. A simplified boiler system consisting of economizer, evaporator, and superheater was constructed to match a 500 MW power boiler. The dynamic response of the system following a disturbance was discussed along with the quantitative response characteristics. The dynamic response of the boiler system was further evaluated by checking the case of an off-design point operation of the feedwater-to-fuel supply ratio. The results re-emphasized the significance of controlling the feedwater-to-fuel supply ratio and additional design requirements of the water-steam separator and spray attemperator.

  1. High sensitivity isotope analysis with a 252Cf--235U fueled subcritical multiplier and low background photon detector systems

    International Nuclear Information System (INIS)

    Wogman, N.A.; Rieck, H.G. Jr.; Laul, J.C.; MacMurdo, K.W.

    1976-09-01

    A 252 Cf activation analysis facility has been developed for routine multielement analysis of a wide variety of solid and liquid samples. The facility contains six sources of 252 Cf totaling slightly over 100 mg. These sources are placed in a 93 percent 235 U-enriched uranium core which is subcritical with a K effective of 0.985 (multiplication factor of 66). The system produces a thermal flux on the order of 10 +1 neutrons per square centimeter per second. A pneumatic rabbit system permits automatic irradiation, decay, and counting regimes to be performed unattended on the samples. The activated isotopes are analyzed through their photon emissions with state-of-the-art intrinsic Ge detectors, Ge(Li) detectors, and NaI(Tl) multidimensional gamma ray spectrometers. High efficiency (25 percent), low background, anticoincidence shielded Ge(Li) gamma ray detector systems have been constructed to provide the lowest possible background, yet maintain a peak to Compton ratio of greater than 1000 to 1. The multidimensional gamma ray spectrometer systems are composed of 23 cm diameter x 20 cm thick NaI(Tl) crystals surrounded by NaI(Tl) anticoincidence shields. The detection limits for over 65 elements have been determined for this system. Over 40 elements are detectable at the 1 part per million level at a precision of +-10 percent

  2. Measurement of subcritical multiplication by the interval distribution method

    International Nuclear Information System (INIS)

    Nelson, G.W.

    1985-01-01

    The prompt decay constant or the subcritical neutron multiplication may be determined by measuring the distribution of the time intervals between successive neutron counts. The distribution data is analyzed by least-squares fitting to a theoretical distribution function derived from a point reactor probability model. Published results of measurements with one- and two-detector systems are discussed. Data collection times are shorter, and statistical errors are smaller the nearer the system is to delayed critical. Several of the measurements indicate that a shorter data collection time and higher accuracy are possible with the interval distribution method than with the Feynman variance method

  3. Analysis of the Techniques for Measuring the Reactivity of Far Sub-Critical Multiplying Systems

    International Nuclear Information System (INIS)

    Vandeplas, P.

    1968-01-01

    The methods of measuring control-rod worth on the basis of reactor response to a periodic excitation source are analysed and it is shown that the ratio of the integrals of the prompt and delayed neutron densities over one period is independent of the shape of the signal used. This ratio will thus be equal to the ratio of the neutron densities in the presence of a time-independent source. The pulsed source, sinusoidal source and source removed methods therefore give identical worth values. Experimentation and numerical analysis show that the ratio of prompt to delayed neutron density, which is a linear combination of successive eigenvalues of the system, is largely dependent on the position of the source. It is demonstrated analytically that with a suitably chosen volume of integration, the contribution of the space harmonics is drastically reduced in the integrals of both the prompt neutron density and the delayed neutron density. Numerical studies show that the ratio of these two integrals is practically equal to the reactivity corresponding to the fundamental static mode and is almost independent of the position of the source. Integration of the neutron densities over a volume can be reduced to an integration along a straight line by a suitable choice of source location. Such an integration can be performed by means of an integral detector, thereby only requiring one measurement. Optimization of the position of the source and of the limits of integration to give the best value for the reactivity can be achieved on the basis of purely experimental criteria. (author) [fr

  4. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  5. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  6. Subcritical calculation of the nuclear material warehouse

    International Nuclear Information System (INIS)

    Garcia M, T.; Mazon R, R.

    2009-01-01

    In this work the subcritical calculation of the nuclear material warehouse of the Reactor TRIGA Mark III labyrinth in the Mexico Nuclear Center is presented. During the adaptation of the nuclear warehouse (vault I), the fuel was temporarily changed to the warehouse (vault II) and it was also carried out the subcritical calculation for this temporary arrangement. The code used for the calculation of the effective multiplication factor, it was the Monte Carlo N-Particle Extended code known as MCNPX, developed by the National Laboratory of Los Alamos, for the particles transport. (Author)

  7. Resonance self-shielding effect analysis of neutron data libraries applied for the dual-cooled waste transmutation blanket of the fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Liu Haibo; Wu Yican; Zheng Shanliang; Zhang Chunzao

    2004-01-01

    Based on the Fusion-Driven Subcritical System (FDS-I), the 25 groups, 175 groups and 620 groups neutron nuclear data libraries with/without resonance self-shielding correction are made with the Njoy and Transx codes, and the K eff and reaction rates are calculated with the Anisn code. The conclusion indicates that the resonance self-shielding effect affects the reaction rates strongly. (authors)

  8. Large subcriticality measurement by pulsed neutron method

    International Nuclear Information System (INIS)

    Yamane, Y.; Yoshida, A.; Nishina, K.; Kobayashi, K.; Kanda, K.

    1985-01-01

    To establish the method determining large subcriticalities in the field of nuclear criticality safety, the authors performed pulsed neutron experiments using the Kyoto University Critical Assembly (KUCA) at Research Reactor Institute, Kyoto University and the Cockcroft-Walton type accelerator attached to the assembly. The area-ratio method proposed by Sjoestrand was employed to evaluate subcriticalities from neutron decay curves measured. This method has the shortcomings that the neutron component due to a decay of delayed neutrons remarkably decreases as the subcriticality of an objective increases. To overcome the shortcoming, the authors increased the frequency of pulsed neutron generation. The integral-version of the area-ratio method proposed by Kosaly and Fisher was employed in addition in order to remove a contamination of spatial higher modes from the decay curve. The latter becomes significant as subcriticality increases. The largest subcriticality determined in the present experiments was 125.4 dollars, which was equal to 0.5111 in a multiplication factor. The calculational values evaluated by the computer code KENO-IV with 137 energy groups based on the Monte Carlo method agreed well with those experimental values

  9. Proposed sub-criticality level for an 80 MWTHd-bismuth-cooled Ads

    International Nuclear Information System (INIS)

    Mansani, L.; Monti, R.; Neuhold, P.

    2003-01-01

    The degree of operational sub-criticality of an Accelerator-driven System (ADS) on the one hand directly affects key accelerator system parameters, such as the proton beam current required to sustain the selected rated power level and, on the other, the likelihood of approaching or attaining criticality under abnormal or accident conditions. Then, if in all such conditions the safety goal is pursued to design the sub-critical core so that it stays away from criticality with adequate margin, the required operational sub-criticality level must be determined by a properly balanced approach between excessively demanding accelerator system performances and risk of accidental criticality. The approach must necessarily include evaluation and appropriate combination of the relevant reactivity effects (e.g. from system cool-down, postulated accident scenarios, geometrical variations) and proper consideration of specific design features (such as, for instance, the absence of safety rods, intended as neutron absorbing devices having a role equivalent to the shutdown rods in critical reactors). The paper presents a possible approach to the determination of the operational sub-criticality level of an 80 MWth Lead-Bismuth-cooled pool type ADS, initially conceived and developed by a team of Italian Organisations led by Ansaldo, with funding from the Ministry of University and Scientific and Technological Research, and currently in the process of being assessed, versus a gas-cooled concept, in the frame of a contract with the Commission of the European Communities. After a brief description of the Lead-Bismuth-cooled ADS concept relevant features and of the key safety goals in terms of required sub-criticality margin, the evaluated reactivity effects are presented, a method to combine them is discussed and a proposed operational sub-criticality level is derived. (author)

  10. 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.)

  11. Subcritical Multiplication Parameters of the Accelerator-Driven System with 100 MeV Protons at the Kyoto University Critical Assembly

    Directory of Open Access Journals (Sweden)

    Jae-Yong Lim

    2012-01-01

    Full Text Available Basic experiments on the accelerator-driven system (ADS at the Kyoto University Critical Assembly are carried out by combining a solid-moderated and -reflected core with the fixed-field alternating gradient accelerator. The reaction rates are measured by the foil activation method to obtain the subcritical multiplication parameters. The numerical calculations are conducted with the use of MCNPX and JENDL/HE-2007 to evaluate the reaction rates of activation foils set in the core region and at the location of the target. Here, a comparison between the measured and calculated eigenvalues reveals a relative difference of around 10% in C/E values. A special mention is made of the fact that the reaction rate analyses in the subcritical systems demonstrate apparently the actual effect of moving the tungsten target into the core on neutron multiplication. A series of further ADS experiments with 100 MeV protons needs to be carried out to evaluate the accuracy of subcritical multiplication parameters.

  12. Development and Investigation of Reactivity Measurement Methods in Subcritical Cores

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Johanna

    2005-05-01

    Subcriticality measurements during core loading and in future accelerator driven systems have a clear safety relevance. In this thesis two subcriticality methods are treated: the Feynman-alpha and the source modulation method. The Feynman-alpha method is a technique to determine the reactivity from the relative variance of the detector counts during a measurement period. The period length is varied to get the full time dependence of the variance-to-mean. The corresponding theoretical formula was known only with stationary sources. In this thesis, due to its relevance for novel reactivity measurement methods, the Feynman-alpha formulae for pulsed sources for both the stochastic and the deterministic cases are treated. Formulae neglecting as well as including the delayed neutrons are derived. The formulae neglecting delayed neutrons are experimentally verified with quite good agreement. The second reactivity measurement technique investigated in this thesis is the so-called source modulation technique. The theory of the method was elaborated on the assumption of point kinetics, but in practice the method will be applied by using the signal from a single local neutron detector. Applicability of the method therefore assumes point kinetic behaviour of the core. Hence, first the conditions of the point kinetic behaviour of subcritical cores was investigated. After that the performance of the source modulation technique in the general case as well as and in the limit of exact point kinetic behaviour was examined. We obtained the unexpected result that the method has a finite, non-negligible error even in the limit of point kinetic behaviour, and a substantial error in the operation range of future accelerator driven subcritical reactors (ADS). In practice therefore the method needs to be calibrated by some other method for on-line applications.

  13. Development and Investigation of Reactivity Measurement Methods in Subcritical Cores

    International Nuclear Information System (INIS)

    Wright, Johanna

    2005-05-01

    Subcriticality measurements during core loading and in future accelerator driven systems have a clear safety relevance. In this thesis two subcriticality methods are treated: the Feynman-alpha and the source modulation method. The Feynman-alpha method is a technique to determine the reactivity from the relative variance of the detector counts during a measurement period. The period length is varied to get the full time dependence of the variance-to-mean. The corresponding theoretical formula was known only with stationary sources. In this thesis, due to its relevance for novel reactivity measurement methods, the Feynman-alpha formulae for pulsed sources for both the stochastic and the deterministic cases are treated. Formulae neglecting as well as including the delayed neutrons are derived. The formulae neglecting delayed neutrons are experimentally verified with quite good agreement. The second reactivity measurement technique investigated in this thesis is the so-called source modulation technique. The theory of the method was elaborated on the assumption of point kinetics, but in practice the method will be applied by using the signal from a single local neutron detector. Applicability of the method therefore assumes point kinetic behaviour of the core. Hence, first the conditions of the point kinetic behaviour of subcritical cores was investigated. After that the performance of the source modulation technique in the general case as well as and in the limit of exact point kinetic behaviour was examined. We obtained the unexpected result that the method has a finite, non-negligible error even in the limit of point kinetic behaviour, and a substantial error in the operation range of future accelerator driven subcritical reactors (ADS). In practice therefore the method needs to be calibrated by some other method for on-line applications

  14. Noise analysis role in reactor safety, Spectral analysis (PSD)

    International Nuclear Information System (INIS)

    Jovanovic, S.; Velickovic, Lj.

    1967-11-01

    Spectral power density of a zero power reactor is frequency dependent and related to transfer function of the reactor and to spectral density of the input disturbance. Measurement of spectral power density of a critical system is used to obtain the ratio (β/l), β is the effective yield of delayed neutrons, and l is the effective mean neutron lifetime. When reactor is subcritical, if the effective yie ald of delayed neutrons, the effective mean neutron lifetime are known, the shutdown margin can be determined by relation α = (1 - k (1- β0)/l, k is the effective multiplication factor. Output neutron spectrum at the RB reactor in Vinca was measured for a few reactor core configurations and for a few levels of heavy water at subcritical state. Measured values were satisfactory when the reactor was critical, but the reactor noise of subcritical system was covered by the white noise of the detector and electronic equipment. The Ra-Be source was under the reactor vessel when measurements of subcritical system were done. More efficient detector or external random stimulus for increasing the intensity of neutron fluctuations would be needed to obtain results for subcritical system

  15. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

    An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

  16. Deep subcritical levels measurements dependents upon kinetic distortion factors

    International Nuclear Information System (INIS)

    Pan Shibiao; Li Xiang; Fu Guo'en; Huang Liyuan; Mu Keliang

    2013-01-01

    The measurement of deep subcritical levels, with the increase of subcriticality, showed that the results impact on the kinetic distortion effect, along with neutron flux strongly deteriorated. Using the diffusion theory, calculations have been carried out to quantify the kinetic distortion correction factors in subcritical systems, and these indicate that epithermal neutron distributions are strongly affected by kinetic distortion. Subcriticality measurements in four different rod-state combination at the zero power device was carried out. The test data analysis shows that, with increasing subcriticality, kinetic distortion effect correction factor gradually increases from 1.052 to 1.065, corresponding reactive correction amount of 0.78β eff ∼ 3.01β eff . Thus, it is necessary to consider the kinetic distortion effect in the deep subcritical reactivity measurements. (authors)

  17. Safety system for reactor container

    International Nuclear Information System (INIS)

    Shimizu, Miwako; Seki, Osamu; Mano, Takio.

    1995-01-01

    A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

  18. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

  19. Physics design of a 10 MeV injector test stand for an accelerator-driven subcritical system

    Directory of Open Access Journals (Sweden)

    Fang Yan

    2015-05-01

    Full Text Available The 10 MeV accelerator-driven subcritical system (ADS Injector I test stand at Institute of High Energy Physics (IHEP is a testing facility dedicated to demonstrate one of the two injector design schemes [Injector Scheme-I, which works at 325 MHz], for the ADS project in China. The injector is composed of two parts, the linac part and the beam dump line. The former is designed on the basis of 325 MHz four-vane type copper structure radio frequency quadrupole and superconducting (SC spoke cavities with β=0.12. The latter is designed to transport the beam coming out of the SC section of the linac to the beam dump, where the beam transverse profile is fairly enlarged and unformed to simplify the beam target design. The SC section consists of two cryomodules with 14 β=0.12 Spoke cavities, 14 solenoid and 14 BPMs in total. The first challenge in the physics design comes from the necessary space required for the cryomodule separation where the periodical lattice is destroyed at a relatively lower energy of ∼5  MeV. Another challenge is the beam dump line design, as it will be the first beam dump line being built by using a step field magnet for the transverse beam expansion and uniformity in the world. This paper gives an overview of the physics design study together with the design principles and machine construction considerations. The results of an optimized design, fabrication status and end to end simulations including machine errors are presented.

  20. Physics design of a 10 MeV injector test stand for an accelerator-driven subcritical system

    Science.gov (United States)

    Yan, Fang; Pei, Shilun; Geng, Huiping; Meng, Cai; Zhao, Yaliang; Sun, Biao; Cheng, Peng; Yang, Zheng; Ouyang, Huafu; Li, Zhihui; Tang, Jingyu; Wang, Jianli; Sui, Yefeng; Dai, Jianping; Sha, Peng; Ge, Rui

    2015-05-01

    The 10 MeV accelerator-driven subcritical system (ADS) Injector I test stand at Institute of High Energy Physics (IHEP) is a testing facility dedicated to demonstrate one of the two injector design schemes [Injector Scheme-I, which works at 325 MHz], for the ADS project in China. The injector is composed of two parts, the linac part and the beam dump line. The former is designed on the basis of 325 MHz four-vane type copper structure radio frequency quadrupole and superconducting (SC) spoke cavities with β =0.12 . The latter is designed to transport the beam coming out of the SC section of the linac to the beam dump, where the beam transverse profile is fairly enlarged and unformed to simplify the beam target design. The SC section consists of two cryomodules with 14 β =0.12 Spoke cavities, 14 solenoid and 14 BPMs in total. The first challenge in the physics design comes from the necessary space required for the cryomodule separation where the periodical lattice is destroyed at a relatively lower energy of ˜5 MeV . Another challenge is the beam dump line design, as it will be the first beam dump line being built by using a step field magnet for the transverse beam expansion and uniformity in the world. This paper gives an overview of the physics design study together with the design principles and machine construction considerations. The results of an optimized design, fabrication status and end to end simulations including machine errors are presented.

  1. Study of two-zone reactor system using a pulsed neutron technique

    Energy Technology Data Exchange (ETDEWEB)

    Shishin, B P; Platovskikh, Yu A; Didejkin, T S

    1977-05-01

    Theoretical and experimental investigations of a neutron flux time dependence after a sport fast neutron pulse in a reactor core - neutron reflector multiplying system have been conducted. A correlation between eigenvalues governing neutron flux decrease at t..-->..infinity for the two-zone system and eigenvalues for each zone has been established in terms of the one-group diffusion approximation. Experiments have been performed in an experimental subcritical assembly comprising a cylindrical uranium core surrounded by a radial water reflector with different boric acid concentrations.

  2. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  3. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    1989-06-01

    Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

  4. Numerical model for thermoeconomic diagnosis in commercial transcritical/subcritical booster refrigeration systems

    DEFF Research Database (Denmark)

    Ommen, Torben; Elmegaard, Brian

    2012-01-01

    cycle supplying refrigerant for evaporators in both chilled and frozen display cases. In the paper, thermoeconomic theory is used to establish the cost of cooling at each individual temperature level based on operating costs.With a high amount of operating systems, faulty operation becomes an economic......, and environmental, interest. A general solution for evaluation of these systems is considered, with the objective to reduce cost and power consumption of malfunctioning equipment in operation. An analysis of the use of thermoeconomic diagnosis methods is required, as these methods may prove applicable...

  5. Solvent refined coal reactor quench system

    Science.gov (United States)

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  6. Critical and subcritical parameters of the system simulating plutonium metal dissolution

    International Nuclear Information System (INIS)

    Vasilev, Yury Yu.; Ryazanov, Boris G.; Sviridov, Victor I.; Mozhayeva, Lubov I.

    2003-01-01

    Dissolution of plutonium metal was simulated using the Monte Carlo computer code to calculate criticality safety limits for the process. Calculations were made for the constant masses of plutonium charged to the dissolving vessel considering distribution of plutonium in metal and solution phases. Critical parameters and limits were calculated as a function of dissolving vessel volume and plutonium metal mass. 240 Pu content was assumed to be from 0% to 10% (mass). Critical parameters were evaluated for the system with a water reflector. Results of this paper may be used in the designing process equipment for plutonium metal dissolution. (author)

  7. Investigations of subcritical crack propagation of the Empress 2 all-ceramic system.

    Science.gov (United States)

    Mitov, Gergo; Lohbauer, Ulrich; Rabbo, Mohammad Abed; Petschelt, Anselm; Pospiech, Peter

    2008-02-01

    The mechanical properties and slow crack propapagation of the all-porcelain system Empress 2 (Ivoclar Vivadent, Schaan, Liechtenstein) with its framework compound Empress 2 and the veneering compounds "Empress 2 and Eris were examined. For all materials, the fracture strength, Weibull parameter and elastic moduli were experimentally determined in a four-point-bending test. For the components of the Empress 2 system, the fracture toughness K(IC) was determined, and the crack propagation parameters n and A were determined in a dynamic fatigue method. Using these data, life data analysis was performed and lifetime diagrams were produced. The development of strength under static fatigue conditions was calculated for a period of 5 years. The newly developed veneering ceramic Eris showed a higher fracture strength (sigma(0)=66.1 MPa) at a failure probability of P(F)=63.2%, and crack growth parameters (n=12.9) compared to the veneering ceramic Empress 2 (sigma(0)=60.3 MPa). For Empress 2 veneer the crack propagation parameter n could only be estimated (n=9.5). This is reflected in the prognosis of long-term resistance presented in the SPT diagrams. For all materials investigated, the Weibull parameter m values (Empress 2 framework m=4.6; Empress 2 veneer m=7.9; Eris m=6.9) were much lower than the minimum demanded by the literature (m=15). The initial fracture strength value alone is not sufficient to characterize the mechanical resistance of ceramic materials, since their stressability is time-dependent. Knowledge about the crack propagation parameters n and A are of great importance when preclinically predicting the clinical suitability of dental ceramic materials. The use of SPT diagrams for lifetime calculation of ceramic materials is a valuable method for comparing different ceramics.

  8. Neutronics optimization of LiPb-He dual-cooled fuel breeding blanket for the fusion-driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican

    2002-01-01

    The concept of the liquid Li 17 Pb 83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239 Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

  9. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    Spiegelberg, R.

    1992-01-01

    The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

  10. Accelerator-driven sub-critical reactor system (ADS) for nuclear ...

    Indian Academy of Sciences (India)

    High-energy accelerator; energy generation; radioactive waste elimination. PACS Nos 29.17. .... Withstands irradiation and thermal effects along with its container. (b) High-power proton ... That is, its availability is >99%. 946. Pramana – J.

  11. K/sub infinity/-meter concept verified via subcritical-critical TRIGA experiments

    International Nuclear Information System (INIS)

    Ocampo Mansilla, H.

    1983-01-01

    This work presents a technique for building a device to measure the k/sub infinity/ of a spent nuclear fuel assembly discharged from the core of a nuclear power plant. The device, called a k/sub infinity/-meter, consists of a cross-shaped subcritical assembly, two artificial neutron sources, and two separate neutron counting systems. The central position of the subcritical assembly is used to measure k/sub infinity/ of the spent fuel assembly. The initial subcritical assembly is calibrated to determine its k/sub eff/ and verify the assigned k/sub infinity/ of a selected fuel assembly placed in the central position. Count rates are taken with the fuel assembly of known k/sub infinity/'s placed in the central position and then repeated with a fuel assembly of unknown k/sub infinity/ placed in the central position. The count rate ratio of the unknown fuel assembly to the known fuel assembly is used to determine the k/sub infinity/ of the unknown fuel assembly. The k/sub infinity/ of the unknown fuel assembly is represented as a polynomial function of the count rate ratios. The coefficients of the polynomial equation are determined using the neutronic codes LEOPARD and EXTERMINATOR-II. The analytical approach has been validated by performing several subcritical/critical experiments, using the Penn State Breazeale TRIGA Reactor (PSBR), and comparing the experimental results with the calculations

  12. Some neutronics of innovative subcritical assembly with fast neutron spectrum

    International Nuclear Information System (INIS)

    Kiyavitskaya, H.; Fokov, Yu.; Rutkovskaya, Ch.; Sadovich, S.; Kasuk, D.; Gohar, Y.; Bolshinsky, I.

    2013-01-01

    Conclusion: • New assembly can be used to: • develop the experimental techniques and adapt the existing ones for monitoring the sub-criticality level, neutron spectra measurements, etc; • study the spatial kinetics of sub-critical and critical systems with fast neutron spectra; • measure the transmutation reaction rates of minor-actinides etc

  13. Performance comparison and parametric optimization of subcritical Organic Rankine Cycle (ORC) and transcritical power cycle system for low-temperature geothermal power generation

    International Nuclear Information System (INIS)

    Shengjun, Zhang; Huaixin, Wang; Tao, Guo

    2011-01-01

    Research highlights: → We conduct the thermodynamic and economic performance comparison of the fluids in both subcritical ORC and transcritical power cycle. → We perform parameter optimization based on five indicators. → The optimum operation parameters and working fluids are not the same for different indicators. → The LEC value is used as the determining factor for fluids screening. → The transcritical power cycle with R125 as the working fluid was a cost-effective approach. - Abstract: Organic Rankine Cycle (ORC) is a promising technology for converting the low-grade energy to electricity. This paper presents an investigation on the parameter optimization and performance comparison of the fluids in subcritical ORC and transcritical power cycle in low-temperature (i.e. 80-100 o C) binary geothermal power system. The optimization procedure was conducted with a simulation program written in Matlab using five indicators: thermal efficiency, exergy efficiency, recovery efficiency, heat exchanger area per unit power output (APR) and the levelized energy cost (LEC). With the given heat source and heat sink conditions, performances of the working fluids were evaluated and compared under their optimized internal operation parameters. The optimum cycle design and the corresponding operation parameters were provided simultaneously. The results indicate that the choice of working fluid varies the objective function and the value of the optimized operation parameters are not all the same for different indicators. R123 in subcritical ORC system yields the highest thermal efficiency and exergy efficiency of 11.1% and 54.1%, respectively. Although the thermal efficiency and exergy efficiency of R125 in transcritical cycle is 46.4% and 20% lower than that of R123 in subcritical ORC, it provides 20.7% larger recovery efficiency. And the LEC value is relatively low. Moreover, 22032L petroleum is saved and 74,019 kg CO 2 is reduced per year when the LEC value is used as

  14. Hybrid systems for transuranic waste transmutation in nuclear power reactors: state of the art and future prospects

    Science.gov (United States)

    Yurov, D. V.; Prikhod'ko, V. V.

    2014-11-01

    The features of subcritical hybrid systems (HSs) are discussed in the context of burning up transuranic wastes from the U-Pu nuclear fuel cycle. The advantages of HSs over conventional atomic reactors are considered, and fuel cycle closure alternatives using HSs and fast neutron reactors are comparatively evaluated. The advantages and disadvantages of two HS types with neutron sources (NSs) of widely different natures -- nuclear spallation in a heavy target by protons and nuclear fusion in magnetically confined plasma -- are discussed in detail. The strengths and weaknesses of HSs are examined, and demand for them for closing the U-Pu nuclear fuel cycle is assessed.

  15. Investigation of the properties of the nuclei using on the new generation reactor technology systems

    International Nuclear Information System (INIS)

    Tel, E.; Sahin, H. M.; Yalcin, S.; Altinok, T.; Kaplan, A.; Aydin, A.

    2007-01-01

    The application fields of the fast neutron are Accelerator-Driven subcritical Systems (ADS) for fission energy production and hybrid reactor systems. The technical design hybrid reactor and ADS systems potentialities require the knowledge of a wide range of better data and much effort. Thorium (Th) and Uranium (U) are nuclear fuels in these reactor systems. Lead (Pb), Bismuth (Bi) and Tungsten (W) are the target nuclei in the ADS reactor systems. The Hartree-Fock (H-F) method with an effective interaction with Skyrme forces is widely used for studying the properties of nuclei such as binding energy, Root Mean Square (RMS) charge radii, mass radii, neutron density, proton density, electromagnetic multipole moments, etc. In this study, by using H-F method with interaction Skyrme RMS charge radii, RMS mass radii, neutron density and proton density were calculated for the 2 32Th, 2 38U, 2 07Pb, 2 09Bi and 1 84W isotopes used on the new generation reactor systems. The calculation results of charge radii have been compared with experimental data and obtained other results have been discussed for hybrid and ADS reactor systems

  16. Tokamak reactor systems studies

    International Nuclear Information System (INIS)

    1992-01-01

    A summary of work completed on the ARIES project during this report period is given. The main areas of effort were: neutronics, shield optimization and design, safety, systems, startup and shutdown, and ripple loss

  17. Reactor limit control system

    International Nuclear Information System (INIS)

    Rubbel, F.E.

    1982-01-01

    The very extensive use of limitations in the operational field between protection system and closed-loop controls is an important feature of German understanding of operational safety. The design of limitations is based on very large activities in the computational field but mostly on the high level of the plant-wide own commissioning experience of a turnkey contractor. Limitations combine intelligence features of closed-loop controls with the high availability of protection systems. (orig.)

  18. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

  19. Steady squares and hexagons on a subcritical ramp

    International Nuclear Information System (INIS)

    Hoyle, R.B.

    1995-01-01

    Steady squares and hexagons on a subcritical ramp are studied, both analytically and numerically, within the framework of the lowest-order amplitude equations. On the subcritical ramp, the external stress or control parameter varies continuously in space from subcritical to supercritical values. At the subcritical end of the ramp, pattern formation is suppressed, and patterns fade away into the conduction solution. It is shown that three-dimensional patterns may change shape on a subcritical ramp. A square pattern becomes a pattern of rolls as it fades, with the roll axes aligned in the direction orthogonal to that in which the control parameter varies. Hexagons in systems with horizontal midplane symmetry become a pattern of rectangles before reaching the conduction solution. There is a suggestion that hexagons in systems which lack this symmetry might fade away through a roll pattern. Numerical simulations are used to illustrate these phenomena

  20. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  1. Ensuring the validity of calculated subcritical limits

    International Nuclear Information System (INIS)

    Clark, H.K.

    1977-01-01

    The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin

  2. Reactor protection and shut-down system

    International Nuclear Information System (INIS)

    Klar

    1980-01-01

    The reactor protection system being a part of the reactor safety system. The requirements on the reactor protection system are: high safety with regard to signal processing, high availability, self-reporting of faults etc. The functional sections of the reactor protection system are the analog section, the logic section and the generating of output signals. Description of the operation characteristics and of the extension of function. (orig.)

  3. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

  4. Reactor protection systems for the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Morris, C.R.

    2003-01-01

    The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

  5. YALINA facility a sub-critical Accelerator- Driven System (ADS) for nuclear energy research facility description and an overview of the research program (1997-2008).

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Smith, D. L.; Nuclear Engineering Division

    2010-04-28

    The YALINA facility is a zero-power, sub-critical assembly driven by a conventional neutron generator. It was conceived, constructed, and put into operation at the Radiation Physics and Chemistry Problems Institute of the National Academy of Sciences of Belarus located in Minsk-Sosny, Belarus. This facility was conceived for the purpose of investigating the static and dynamic neutronics properties of accelerator driven sub-critical systems, and to serve as a neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinide nuclei. This report provides a detailed description of this facility and documents the progress of research carried out there during a period of approximately a decade since the facility was conceived and built until the end of 2008. During its history of development and operation to date (1997-2008), the YALINA facility has hosted several foreign groups that worked with the resident staff as collaborators. The participation of Argonne National Laboratory in the YALINA research programs commenced in 2005. For obvious reasons, special emphasis is placed in this report on the work at YALINA facility that has involved Argonne's participation. Attention is given here to the experimental program at YALINA facility as well as to analytical investigations aimed at validating codes and computational procedures and at providing a better understanding of the physics and operational behavior of the YALINA facility in particular, and ADS systems in general, during the period 1997-2008.

  6. Dosimetry system of the RB reactor

    International Nuclear Information System (INIS)

    Lolic, B.; Vukadin, D.

    1962-01-01

    Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor

  7. Nuclear reactor vessel decontamination systems

    International Nuclear Information System (INIS)

    McGuire, P. J.

    1985-01-01

    There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

  8. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  9. Emergency system for nuclear reactors

    International Nuclear Information System (INIS)

    1976-01-01

    The invention concerns a circuit called 'of emergency help' intended to remove, in a safe and quick manner, the residual thermal power on the safety vessel of a fast neutron reactor cooled by a liquid metal flow, in the event of a failure occurring inside the main reactor vessel or on it. This system includes a network of spray nozzle tubes, distributed around and near the external surface of the safety vessel, to project on to the surface of the vessel a mist of a liquid having high latent vaporisation heat. The steam produced on contact with the safety vessel is collected in the space provided between the safety vessel and the external protection vessel by at least one collector pipe for dischaging this steam outside the vessel. Under a preferred design mode of the invention the liquid is water the use of which turns out to be particularly advantageous in practice owing to its favourable physical properties and its low cost [fr

  10. The Optimization of power reactor control system

    International Nuclear Information System (INIS)

    Danupoyo, S.D.

    1997-01-01

    A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

  11. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Bolton, John A.

    1970-01-01

    Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

  12. Study of two-zone reactor system using a pulsed neutron technique

    International Nuclear Information System (INIS)

    Shishin, B.P.; Platovskikh, Yu.A.; Didejkin, T.S.

    1977-01-01

    Theoretical and experimental investigations of a neutron flux time dependence after a sport fast neutron pulse in a reactor core - neutron reflector multiplying system have been conducted. A correlation between eigenvalues governing neutron flux decrease at t→infinity for the two-zone system and eigenvalues for each zone has been established in terms of the one-group diffusion approximation. Experiments have been performed in an experimental subcritical assembly comprising a cylindrical uranium core surrounded by a radial water reflector with different boric acid concentrations. The experiments show that the observed neutron flux decrease in the core is governed by an exponent exp(-Λ 1 t), whereas in the reflector by a sum of two exponents exp(-Λ 1 t) and exp(-Λ 2 t). The eigenvalue Λ 1 reflects multiplying properties of the reactor, and Λ 2 is determined by the reflector absorption cross section

  13. Research Programme for the 660 Mev Proton Accelerator Driven MOX-Plutonium Subcritical Assembly

    CERN Document Server

    Barashenkov, V S; Buttseva, G L; Dudarev, S Yu; Polanski, A; Puzynin, I V; Sissakian, A N

    2000-01-01

    The paper presents a research programme of the Experimental Acclerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton acceletator operating at the Laboratory of Nuclear Problems of the JINR, Dubna. MOX fuel (25% PuO_2 + 75% UO_2) designed for the BN-600 reactor use will be adopted for the core of the assembly. The present conceptual design of the experimental subcritical assembly is based on a core of a nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient k_eff = 0.945, energetic gain G = 30 and the accelerator beam power 0.5 kW.

  14. Research programme for the 660 MeV proton accelerator driven MOX-plutonium subcritical assembly

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Buttsev, V.S.; Buttseva, G.L.; Dudarev, S.Yu.; Polanski, A.; Puzynin, I.V.; Sissakyan, A.N.

    2000-01-01

    The paper presents a research programme of the Experimental Accelerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton accelerator operating at the Laboratory of Nuclear Problems of the JINR, Dubna. MOX fuel (25% PuO 2 + 75% UO 2 ) designed for the BN-600 reactor use will be adopted for the core of the assembly. The present conceptual design of the experimental subcritical assembly is based on a core of a nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient k eff = 0.945, energetic gain G=30 and the accelerator beam power 0.5 kW

  15. Digital control system of advanced reactor

    International Nuclear Information System (INIS)

    Peng Huaqing; Zhang Rui; Liu Lixin

    2001-01-01

    This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

  16. Propose Reactor Control and Monitoring System for RTP

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha

    2011-01-01

    Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

  17. Recirculation system for nuclear reactors

    International Nuclear Information System (INIS)

    Braun, H. E.; Dollard, W. J.; Tower, S. N.

    1980-01-01

    A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

  18. New digital control and power protection system of VR 1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Juoeickova, M.

    2005-01-01

    The contribution describes the new VR-1 training reactor control and power protection system at the Czech Technical University in Prague. The control system provides safety and control functions, calculates average values of the important variables and sends data and system status to the human-machine interface. The upgraded control system is based on a high quality industrial PC. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The software was developed according to requirements in MS Visual C. The independent power protection system is a component of the reactor safety (protection) system with high quality and reliability requirements. The digital system is redundant; each channel evaluates the reactor power and the velocity of power changes and provides safety functions. The digital part of the channel is multiprocessor-based. The software was developed with respect to nuclear standards. The software design was coded in the C language regarding the NRC restrictions. Configuration management, verification and validation accompanied the software development. Both systems were thoroughly tested. Firstly, the non active tests were carried out. During these tests, the active core of the reactor was subcritical; the input signals were generated from HPIB and VXI controlled instruments to simulate different operational and safety events. The software for instruments control and tests evaluation utilized Agilent VEE development system. After the successful non active checking, the active tests followed. (author)

  19. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  20. Breeding description for fast reactors and symbiotic reactor systems

    International Nuclear Information System (INIS)

    Hanan, N.A.

    1979-01-01

    A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

  1. Pourbaix diagrams for the iron–water system extended to high-subcritical and low-supercritical conditions

    International Nuclear Information System (INIS)

    Cook, William G.; Olive, Robert P.

    2012-01-01

    Highlights: ► Pourbaix diagrams for iron–water are extended to low-supercritical temperatures. ► Thermodynamic properties for use in R-HKF model re-evaluated. ► Above the critical point, magnetite solubility is between 10 −11 and 10 −10 mol/kg. - Abstract: The supercritical water-cooled reactor (SCWR) is a Generation IV reactor concept that will operate at temperatures and pressures above water’s thermodynamic critical point. Pourbaix diagrams for the iron–water system at temperatures slightly below and above the critical point at 25 MPa have been constructed to aid the evaluation and development of potential construction materials. High temperature data extrapolation was performed using a revised Helgeson–Kirkham–Flowers model and fit to data on magnetite and hematite solubility in high-temperature water. A low-concentration diagram at 350 °C reveals the importance of water chemistry control to avoid transitioning to an active corrosion region.

  2. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

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

  3. An experimental accelerator driven system based on plutonium subcritical assembly and 660 MeV protons accelerator

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Puzynin, I.V.; Sisakyan, A.N.; Polanski, A.

    1999-01-01

    We present a Plutonium Based Energy Amplifier Testing Concept, which employs a plutonium subcritical assembly and a 660 MeV proton accelerator operating in the JINR Laboratory of Nuclear Problems. Fuel designed for the pulsed neutron source IREN (Laboratory of Neutron Physics, JINR) will be adopted for the core of the assembly. To make the present conceptual design of the Plutonium Energy Amplifier we have chosen a nominal unit capacity of 20 kW (thermal). This corresponds to the multiplication coefficient K eff ranging between 0.94 and 0.95 and the energetic gain about 20. Accelerated current is in the range of 1-1.6μA

  4. An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

    International Nuclear Information System (INIS)

    Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

    1976-01-01

    It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

  5. Reaction kinetics of cellulose hydrolysis in subcritical and supercritical water

    Science.gov (United States)

    Olanrewaju, Kazeem Bode

    The uncertainties in the continuous supply of fossil fuels from the crisis-ridden oil-rich region of the world is fast shifting focus on the need to utilize cellulosic biomass and develop more efficient technologies for its conversion to fuels and chemicals. One such technology is the rapid degradation of cellulose in supercritical water without the need for an enzyme or inorganic catalyst such as acid. This project focused on the study of reaction kinetics of cellulose hydrolysis in subcritical and supercritical water. Cellulose reactions at hydrothermal conditions can proceed via the homogeneous route involving dissolution and hydrolysis or the heterogeneous path of surface hydrolysis. The work is divided into three main parts. First, the detailed kinetic analysis of cellulose reactions in micro- and tubular reactors was conducted. Reaction kinetics models were applied, and kinetics parameters at both subcritical and supercritical conditions were evaluated. The second major task was the evaluation of yields of water soluble hydrolysates obtained from the hydrolysis of cellulose and starch in hydrothermal reactors. Lastly, changes in molecular weight distribution due to hydrothermolytic degradation of cellulose were investigated. These changes were also simulated based on different modes of scission, and the pattern generated from simulation was compared with the distribution pattern from experiments. For a better understanding of the reaction kinetics of cellulose in subcritical and supercritical water, a series of reactions was conducted in the microreactor. Hydrolysis of cellulose was performed at subcritical temperatures ranging from 270 to 340 °C (tau = 0.40--0.88 s). For the dissolution of cellulose, the reaction was conducted at supercritical temperatures ranging from 375 to 395 °C (tau = 0.27--0.44 s). The operating pressure for the reactions at both subcritical and supercritical conditions was 5000 psig. The results show that the rate-limiting step in

  6. The mechanical design and fabrication of 162.5 MHz buncher for China accelerator driven sub-critical system injector II

    Energy Technology Data Exchange (ETDEWEB)

    Niu, Hai Hua; Li, Youtang [Lanzhou University of Technology, Lanzhou (China); He, Yuan; Zhang, Bin; Huang, Shichun; Yuan, Chenzhang; Jia, Huan; Zhang, Shenghu [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou (China)

    2017-08-15

    A buncher is one of the main pieces of equipment in the medium energy beam transport line (MEBT) for China accelerator driven sub-critical system (C-ADS) Injector II. To focus the beam longitudinally and match the beam for the acceptance of the superconducting linac section, two room temperature quarter wave resonator (QWR) bunchers with frequency of 162.5 MHz have been designed as parts of the MEBT. According to the beam transmission matching of the MEBT and the geometric parameters requirements of bunchers, the unique mechanical structure and the main processing technology of buncher cavities and their couplers and tuners are described in this paper. The fabrication of bunchers and their parts have been completed and tested at high power, the test results agree well with the design requirements. These bunchers work well for about two years in Institute of Modern Physics, Chinese Academy of Sciences.

  7. The mechanical design and fabrication of 162.5 MHz buncher for China accelerator driven sub-critical system injector II

    Directory of Open Access Journals (Sweden)

    Haihua Niu

    2017-08-01

    Full Text Available A buncher is one of the main pieces of equipment in the medium energy beam transport line (MEBT for China accelerator driven sub-critical system (C-ADS Injector II. To focus the beam longitudinally and match the beam for the acceptance of the superconducting linac section, two room temperature quarter wave resonator (QWR bunchers with frequency of 162.5 MHz have been designed as parts of the MEBT. According to the beam transmission matching of the MEBT and the geometric parameters requirements of bunchers, the unique mechanical structure and the main processing technology of buncher cavities and their couplers and tuners are described in this paper. The fabrication of bunchers and their parts have been completed and tested at high power, the test results agree well with the design requirements. These bunchers work well for about two years in Institute of Modern Physics, Chinese Academy of Sciences.

  8. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  9. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Zhang-cheng; Xie, Heng

    2014-01-01

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  10. Final report for fuel acquisition and design of a fast subcritical blanket facility

    International Nuclear Information System (INIS)

    Clikeman, F.M.; Ott, K.O.

    1976-01-01

    A summary is presented of work leading to the design of a subcritical facility for the study of fast reactor blankets. Included are activities related to fuel acquisition, design of the facility, and experiment planning

  11. Superfluid helium at subcritical active core

    International Nuclear Information System (INIS)

    Vasil'ev, V.V.; Lopatkin, A.V.; Muratov, V.G.; Rakhno, I.L.

    2002-01-01

    Power range and neutron flux wherein super thermal source was realized at high volume of superfluid helium were investigated. MCU, BRAND, MCNP codes were used for the calculation of reactors. It is shown that the availability of full-size diameter for cryogenic source of ultracold neutrons, as the source with superfluid helium is considered, is possible in the reflector of subcritical assembly. Results obtained from the MCNP-4B code application demonstrated that the density of thermal neutron flux in helium must be not higher than 2.3 x 10 11 s -1 cm -2 [ru

  12. Activity report of working party on reactor physics of accelerator-driven system. July 1999 to March 2001

    International Nuclear Information System (INIS)

    2002-02-01

    Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) was set in July 1999 to review and investigate special subjects related to reactor physics research for the Accelerator-Driven Subcritical System (ADS). The ADS-WP, at the first meeting, discussed a guideline of its activity for two years and decided to concentrate upon three subjects: (1) neutron transport calculations in high energy range, (2) static and kinetic (safety-related) characteristics of subcritical system, and (3) system design including ADS concepts and elemental technology developments required. The activity of ADS-WP continued from July 1999 to March 2001. In this duration, the members of ADS-WP met together four times and discussed the above subjects. In addition, the ADS-WP conducted a questionnaire on requests and proposals for the plan of Transmutation Physics Experimental Facility in the High-Intensity Proton Accelerator Project, which is a joint project between JAERI and KEK (High Energy Accelerator Research Organization). This report summarizes the results obtained by the above ADS-WP activity. (author)

  13. Physics of subcritical multiplying regions and experimental validation

    International Nuclear Information System (INIS)

    Salvatores, M.

    1996-01-01

    The coupling of a particle accelerator with a spallation target and with a subcritical multiplying region has been proposed in the fifties and is called here a hybrid system. This article gives some ideas about the energetic balance of such a system. The possibilities of experimental validation of some properties of a subcritical multiplying region by using MASURCA facility at CEA-Cadarache are examined. The results of a preliminary experiment called MUSE are presented. (A.C.)

  14. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  15. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    1999-01-01

    Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

  16. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter

    2003-01-01

    Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible for these ......Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible...... to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...

  17. RSAS: a Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Dixon, B.W.; Bray, M.A.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

  18. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  19. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Science.gov (United States)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  20. To the question of definition of fissile material mass and neutron multiplication in deep sub-critical systems

    International Nuclear Information System (INIS)

    Dulin, V.V.

    2006-01-01

    A method of determination neutrons multiplication in deep sub-critical multiplying media has been developed. It is based on a modified of Rossi - alpha method. It will consist in use of integral on time (a method of the areas) from correlated parts of distribution and integral in area, independent of time a part of distribution (area of a constant background). It allows to spend the calculated analysis, using the integrated equation on time for a neutrons flux and to not use representation of point kinetic model. A calculation spatially-correlation factor the adjoint (relative the detector count rate) inhomogeneous equation is used. Its calculation takes into account fission both in multiplying media and in a spontaneous neutron source. Measurements with plutonium-steel and uranium-steel blocks, and blocks from uranium and plutonium dioxide of different enrichment are have been carried out. The measured values of neutrons multiplication in a range 1.03-1.82 will be well coordinated to results of calculations. The question on an opportunity of definition of weight of the measured blocks of fissile material is considered [ru

  1. Nuclear reactor auxiliary heat removal system

    International Nuclear Information System (INIS)

    Thompson, R.E.; Pierce, B.L.

    1977-01-01

    An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

  2. Subcritical reactivity measurement at Angra 1 nuclear power plant

    International Nuclear Information System (INIS)

    Kuramoto, Renato Yoichi Ribeiro; Miranda, Anselmo Ferreira

    2011-01-01

    In order to speed up the Angra 1 NPP physics tests, this work intends to develop a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements at Angra 1 NPP. In the first part of this work, we have applied the Modified Neutron Source Multiplication (MNSM) Method with fundamental mode extraction, in order to improve the monitoring of the subcriticality at Angra 1 NPP during the criticality approach. In the second part, we developed a preliminary subcritical reactivity meter algorithm based on the point-reactor inverse kinetic model with six delayed neutron groups and external neutron source. The source strength was obtained through the Least Squares Inverse Kinetics Method (LSIKM). (author)

  3. MCNPX, MONK, and ERANOS analyses of the YALINA Booster subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto, E-mail: alby@anl.go [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Gohar, Y.; Aliberti, G.; Cao, Y.; Smith, D.; Zhong, Z. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I. [Joint Institute for Power and Nuclear Research - Sosny, National Academy of Sciences of Belarus, 99 Acad. Krasin Str., Minsk 220109 (Belarus)

    2011-05-15

    This paper compares the numerical results obtained from various nuclear codes and nuclear data libraries with the YALINA Booster subcritical assembly (Minsk, Belarus) experimental results. This subcritical assembly was constructed to study the physics and the operation of accelerator-driven subcritical systems (ADS) for transmuting the light water reactors (LWR) spent nuclear fuel. The YALINA Booster facility has been accurately modeled, with no material homogenization, by the Monte Carlo codes MCNPX (MCNP/MCB) and MONK. The MONK geometrical model matches that of MCNPX. The assembly has also been analyzed by the deterministic code ERANOS. In addition, the differences between the effective neutron multiplication factor and the source multiplication factors have been examined by alternative calculational methodologies. The analyses include the delayed neutron fraction, prompt neutron lifetime, generation time, neutron flux profiles, and spectra in various experimental channels. The accuracy of the numerical models has been enhanced by accounting for all material impurities and the actual density of the polyethylene material used in the assembly (the latter value was obtained by dividing the total weight of the polyethylene by its volume in the numerical model). There is good agreement between the results from MONK, MCNPX, and ERANOS. The ERANOS results show small differences relative to the other results because of material homogenization and the energy and angle discretizations.The MCNPX results match the experimental measurements of the {sup 3}He(n,p) reaction rates obtained with the californium neutron source.

  4. MCNPX, MONK, and ERANOS analyses of the YALINA Booster subcritical assembly

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gohar, Y.; Aliberti, G.; Cao, Y.; Smith, D.; Zhong, Z.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I.

    2011-01-01

    This paper compares the numerical results obtained from various nuclear codes and nuclear data libraries with the YALINA Booster subcritical assembly (Minsk, Belarus) experimental results. This subcritical assembly was constructed to study the physics and the operation of accelerator-driven subcritical systems (ADS) for transmuting the light water reactors (LWR) spent nuclear fuel. The YALINA Booster facility has been accurately modeled, with no material homogenization, by the Monte Carlo codes MCNPX (MCNP/MCB) and MONK. The MONK geometrical model matches that of MCNPX. The assembly has also been analyzed by the deterministic code ERANOS. In addition, the differences between the effective neutron multiplication factor and the source multiplication factors have been examined by alternative calculational methodologies. The analyses include the delayed neutron fraction, prompt neutron lifetime, generation time, neutron flux profiles, and spectra in various experimental channels. The accuracy of the numerical models has been enhanced by accounting for all material impurities and the actual density of the polyethylene material used in the assembly (the latter value was obtained by dividing the total weight of the polyethylene by its volume in the numerical model). There is good agreement between the results from MONK, MCNPX, and ERANOS. The ERANOS results show small differences relative to the other results because of material homogenization and the energy and angle discretizations.The MCNPX results match the experimental measurements of the 3 He(n,p) reaction rates obtained with the californium neutron source.

  5. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    Parcy, J.P.

    1982-09-01

    A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented [fr

  6. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    Ipek, O.

    1997-01-01

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  7. Burning of spent fuel of an accelerator-driven modular HTGR in sub-critical condition

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Chang Hong; Wu Zongxin; Gu Yuxiang

    2002-01-01

    The modular high temperature gas cooled reactor (MHTGR) has good safety characteristics because of the use of coated particles in the fuel element. After the particles cool outside of the reactor for some time, the spent fuel can be re-utilized. The author describes a physics feasibility study for the burning of spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor in an accelerator-driven sub-critical reactor. A conceptual design is given for the 30 MW accelerator-driven sub-critical reactor. The neutron transport in the sub-critical reactor was simulated using the MCNP code, and the burnup was calculated using the ORIGEN2 code. The results show that the accelerator-driven sub-critical gas-cooled reactor has reliable sub-criticality and low power density and that the spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor can be burned to provide 20% more energy

  8. Measurements relevant to simulating subcriticality in ADS facilities with blanket

    International Nuclear Information System (INIS)

    Titarenko, Yu. E.; Batyaev, V.F.; Borovlev, S.P.; Gladkikh, N.G.; Igumnov, M.M.; Legostaev, V.O.; Karpikhin, E.I.; Konev, V.N.; Kushnerev, Yu.T.; Popkov, V.N.; Ryazhsky, V.I.; Spiridonov, V.G.; Chernyavsky, E.V.; Shvedov, O.V.

    2009-10-01

    The work presents the results of determining the blanket subcriticality for a zero-power heavy water reactor MAKET at the Institute for Theoretical and Experimental Physics, Moscow. The blanket is hexagonal lattice made of 36 90%-enriched 235U fuel rods spaced 173mm apart. The subcriticality was varied from ∼0.3% to 5% by adjusting the heavy water level. The subcriticality values were calibrated using the dependence of reactivity on heavy water level. The pulsed neutron source technique was used to measure the temporal dependence of neutron field at different blanket points for the calibrated subcriticality values. The subciticality values obtained in terms of the 'inverse clock' formulae using the decay constants of the measured dependences proved to differ from the calibrated subcriticalities by not more than 7% at the average. The MCNP code-aided simulations of the experiment made has given the calibrated keff values at prescribed heavy water levels and led to the neutron field decay constants at given points, which differ on the average from their experimental values by not more than 7% too. (author)

  9. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  10. Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dierckx, Marc; Van-Dyck, Dries

    2013-06-01

    We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

  11. Innovative inspection system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Mertens, K.; Trautmann, H.

    1999-01-01

    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [de

  12. Proceedings of workshop on reactor shutdown system

    International Nuclear Information System (INIS)

    1997-03-01

    India has gained considerable experience in design, development, construction and operation of research and power reactors during the last four decades. Reactor shutdown system (RSS) is the most important engineered safety system of any reactor. A lot of technological developments have taken place to improve the reactor shutdown systems, particularly with advancement in reliability analysis and instrumentation and control. If the reactor is not shutdown, the fuel may melt, releasing radioactivity and possibly reactivity addition as in the case of Fast Breeder Reactor (FBR). Apart from radiological safety consequences, large investment has to be written off. The function of the RSS is to stop fission chain reaction and prevent breach of fuel. The design of RSS is multidisciplinary. It requires reactor physics analysis, design of absorber rods, drive mechanisms, safety logic to order shutdown and instrumentation to detect unsafe conditions. High reliability is essential and this requires two independent shutdown systems. This book contains the proceedings of the workshop on reactor shutdown system and papers relevant to INIS are indexed separately

  13. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  14. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

  15. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  16. Reactor core design aiding system

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Hamaguchi, Yukio; Nakao, Takashi; Kondo, Yasuhide

    1995-01-01

    A two-dimensional radial power distribution and an axial one-dimensional power distribution are determined based on a distribution of a three-dimensional infinite multiplication factor, to obtain estimated power distribution estimation values. The estimation values are synthesized to obtain estimated three-dimensional power distribution values. In addition, the distribution of a two-dimensional radial multiplication factor and the distribution of an one-dimensional axial multiplication factor are determined based on the three-dimensional power distribution, to obtain estimated values for the multiplication factor distribution. The estimated values are synthesized to form estimated values for the three-dimensional multiplication factor distribution. Further, estimated fuel loading pattern value is determined based on the three-dimensional power distribution or the two-dimensional radial power distribution. Since the processes for determining the estimated values comprise only additive and multiplying operations, processing time can be remarkably saved compared with calculation based on a detailed physical models. Since the estimation is performed on every fuel assemblies, a nervous circuit network not depending on the reactor core system can be constituted. (N.H.)

  17. Additional reactor protection system of RBMK-1500

    International Nuclear Information System (INIS)

    1999-01-01

    Analysis of anticipated transients without scram of RBMK-1500 reactor showed that additional reactor protection system is required. Data of accident analysis in the case of loose of external electric power and loose of vacuum in condensers of turbines are provided

  18. 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)

  19. New technology for reactor protection system of CAREM reactor

    International Nuclear Information System (INIS)

    Dezzutti, J.C.; Verrastro, C.; Estryk, D.

    2009-01-01

    The use of FPGA in safety functions in a nuclear power plant, increase the reliability of software based systems, without loose any of the function required by the supervision and control systems. In this work the architecture of a Reactor Protection System is described, it use four independent measurement channels in 2 oo 4 configuration, each channel is based on diverse approach in 1 oo 2 configuration, the reliability of this system is near the same than the hardwired logic, with full performance like software based system. (author)

  20. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  1. Distributed expert systems for nuclear reactor control

    International Nuclear Information System (INIS)

    Otaduy, P.J.

    1992-01-01

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed

  2. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth

    2018-01-30

    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

  3. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

    Helmick, H.H.; Schwenk, F.C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

  4. Optimization of Biomass-Fuelled Combined Cooling, Heating and Power (CCHP Systems Integrated with Subcritical or Transcritical Organic Rankine Cycles (ORCs

    Directory of Open Access Journals (Sweden)

    Daniel Maraver

    2014-04-01

    Full Text Available This work is focused on the thermodynamic optimization of Organic Rankine Cycles (ORCs, coupled with absorption or adsorption cooling units, for combined cooling heating and power (CCHP generation from biomass combustion. Results were obtained by modelling with the main aim of providing optimization guidelines for the operating conditions of these types of systems, specifically the subcritical or transcritical ORC, when integrated in a CCHP system to supply typical heating and cooling demands in the tertiary sector. The thermodynamic approach was complemented, to avoid its possible limitations, by the technological constraints of the expander, the heat exchangers and the pump of the ORC. The working fluids considered are: n-pentane, n-heptane, octamethyltrisiloxane, toluene and dodecamethylcyclohexasiloxane. In addition, the energy and environmental performance of the different optimal CCHP plants was investigated. The optimal plant from the energy and environmental point of view is the one integrated by a toluene recuperative ORC, although it is limited to a development with a turbine type expander. Also, the trigeneration plant could be developed in an energy and environmental efficient way with an n-pentane recuperative ORC and a volumetric type expander.

  5. Stack Monitoring System At PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Zamrul Faizad Omar; Mohd Sabri Minhat; Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Izhar Abu Hussin

    2014-01-01

    This paper describes the current Stack Monitoring System at PUSPATI TRIGA Reactor (RTP) building. A stack monitoring system is a continuous air monitor placed at the reactor top for monitoring the presence of radioactive gaseous in the effluent air from the RTP building. The system consists of four detectors that provide the reading for background, particulate, Iodine and Noble gas. There is a plan to replace the current system due to frequent fault of the system, thus thorough understanding of the current system is required. Overview of the whole system will be explained in this paper. Some current results would be displayed and moving forward brief plan would be mentioned. (author)

  6. Sampling system for a boiling reactor NPP

    International Nuclear Information System (INIS)

    Zabelin, A.I.; Yakovleva, E.D.; Solov'ev, Yu.A.

    1976-01-01

    Investigations and pilot running of the nuclear power plant with a VK-50 boiling reactor reveal the necessity of normalizing the design system of water sampling and of mandatory replacement of the needle-type throttle device by a helical one. A method for designing a helical throttle device has been worked out. The quantitative characteristics of depositions of corrosion products along the line of reactor water sampling are presented. Recommendations are given on the organizaton of the sampling system of a nuclear power plant with BWR type reactors

  7. Experimental study on neutronics in bombardment of thick targets by high energy proton beams for accelerator-driven sub-critical system

    CERN Document Server

    Guo Shi Lun; Shi Yong Qian; Shen Qing Biao; Wan Jun Sheng; Brandt, R; Vater, P; Kulakov, B A; Krivopustov, M I; Sosnin, A N

    2002-01-01

    The experimental study on neutronics in the target region of accelerator-driven sub-critical system is carried out by using the high energy accelerator in Joint Institute for Nuclear Research, Dubna, Russia. The experiments with targets U(Pb), Pb and Hg bombarded by 0.533, 1.0, 3.7 and 7.4 GeV proton beams show that the neutron yield ratio of U(Pb) to Hg and Pb to Hg targets is (2.10 +- 0.10) and (1.76 +- 0.33), respectively. Hg target is disadvantageous to U(Pb) and Pb targets to get more neutrons. Neutron yield drops along 20 cm thick targets as the thickness penetrated by protons increases. The lower the energy of protons, the steeper the neutron yield drops. In order to get more uniform field of neutrons in the targets, the energy of protons from accelerators should not be lower than 1 GeV. The spectra of secondary neutrons produced by different energies of protons are similar, but the proportion of neutrons with higher energy gradually increases as the proton energy increases

  8. ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

    Directory of Open Access Journals (Sweden)

    Alexander Labutin

    2017-02-01

    Full Text Available The problem of the analytical synthesis of the synergetic control system of chemical reactor for the realization of a complex series-parallel exothermal reaction has been solved. The synthesis of control principles is performed using the analytical design method of aggregated regulators. Synthesized nonlinear control system solves the problem of stabilization of the concentration of target component at the exit of reactor and also enables one to automatically transfer to new production using the equipment.

  9. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  10. Reactor protection systems of 500 MWe PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

    1997-03-01

    The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

  11. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  12. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Neil Todreas; Pavel Hejzlar

    2008-01-01

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  13. Development of an automatic reactor inspection system

    International Nuclear Information System (INIS)

    Kim, Jae Hee; Eom, Heung Seop; Lee, Jae Cheol; Choi, Yoo Raek; Moon, Soon Seung

    2002-02-01

    Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

  14. Accelerator-driven systems (ADS) and fast reactors (FR) in advanced nuclear fuel cycles

    International Nuclear Information System (INIS)

    2002-01-01

    The long-term hazard of radioactive waste arising from nuclear energy production is a matter of continued discussion and public concern in many countries. Through partitioning and transmutation (P and T) of the actinides and some of the long-lived fission products, the radiotoxicity of high-level waste (HLW) can be reduced by a factor of 100 compared with the current once-through fuel cycle. This requires very effective reactor and fuel cycle strategies, including fast reactors (FR) and/or accelerator-driven, sub-critical systems (ADS). The present study compares FR- and ADS-based actinide transmutation systems with respect to reactor properties, fuel cycle requirements, safety, economic aspects and (R and D) needs. Several advanced fuel cycle strategies are analysed in a consistent manner to provide insight into the essential differences between the various systems in which the role of ADS is emphasised. The report includes a summary aimed at policy makers and research managers as well as a detailed technical section for experts in this domain. (authors)

  15. REACTOR CONTROL ROD OPERATING SYSTEM

    Science.gov (United States)

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  16. Reactor core operation management system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Tomomi.

    1992-05-28

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

  17. Reactor core operation management system

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1992-01-01

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

  18. Cooling system upon reactor isolation

    International Nuclear Information System (INIS)

    Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

    1992-01-01

    A water level indicator for detecting the upper limit value for a range of using a suppression pool and a thermometer for detecting the temperature of water at the cooling water inlet of an auxiliary device are disposed. When a detection signal is intaken and the water level in the suppression pool reach the upper limit value for the range of use, a secondary flow rate control value is opened and a primary flow rate control valve is closed. When the temperature of the water at the cooling water inlet of the auxiliary device reaches the upper limit value, the primary and the secondary flow rate control valves are opened. During a stand-by state, the first flow rate control valve is set open and the secondary flow rate control valve is set closed respectively. After reactor isolation, if a reactor water low level signal is received, an RCIC pump is actuated and cooling water is sent automatically under pressure from a condensate storage tank to the reactor and the auxiliary device requiring coolants by way of the primary flow rate control valve. Rated flow rate is ensured in the reactor and cooling water of an appropriate temperature can be supplied to the auxiliary device. (N.H.)

  19. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  20. 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.)

  1. Reprocessing of gas-cooled reactor particulate graphite fuel in a multi-strata transmutation system

    International Nuclear Information System (INIS)

    Laidler, J.J.

    2001-01-01

    Spent nuclear fuel discharged for light water reactors (LWRs) contains significant quantities of plutonium and other transuranic elements. Recent practice in Europe and Japan has been to recover the plutonium from spent fuel and recycle it to LWRs in the form of mixed uranium-plutonium oxide (MOX) fuel. Irradiation of the recycle fuel results in the generation of further plutonium and an increase in the isotopic concentration of the higher isotopes of plutonium, those having much lover fission cross sections than 239 Pu. This restricts plutonium recycle to one or two cycles, after which use of the plutonium becomes economically unfavorable. Recycle of the highly-transmuted plutonium in fast spectrum reactors can be an efficient method of fissioning this plutonium as well as other minor transuranics such as neptunium, americium and perhaps even curium. Those minor transuranics that are not conveniently burned in a fast reactor can be sent to an accelerator driven subcritical transmutation device for ultimate destruction. The preceding describes what has become known as a 'dual strata' or 'multi-strata' system. It is driven by the incentives to realize the maximum amount of energy from nuclear fuel and to eliminate the discharge of radio-toxic transuranic elements to the environment. Its implementation will be dependent in the long run upon the economic viability of the system and on the value placed by society on the elimination of radio-toxic materials that can conceivably be used in the manufacture of weapons of mass destruction. (author)

  2. Numerical simulation of stochastic point kinetic equation in the dynamical system of nuclear reactor

    International Nuclear Information System (INIS)

    Saha Ray, S.

    2012-01-01

    Highlights: ► In this paper stochastic neutron point kinetic equations have been analyzed. ► Euler–Maruyama method and Strong Taylor 1.5 order method have been discussed. ► These methods are applied for the solution of stochastic point kinetic equations. ► Comparison between the results of these methods and others are presented in tables. ► Graphs for neutron and precursor sample paths are also presented. -- Abstract: In the present paper, the numerical approximation methods, applied to efficiently calculate the solution for stochastic point kinetic equations () in nuclear reactor dynamics, are investigated. A system of Itô stochastic differential equations has been analyzed to model the neutron density and the delayed neutron precursors in a point nuclear reactor. The resulting system of Itô stochastic differential equations are solved over each time-step size. The methods are verified by considering different initial conditions, experimental data and over constant reactivities. The computational results indicate that the methods are simple and suitable for solving stochastic point kinetic equations. In this article, a numerical investigation is made in order to observe the random oscillations in neutron and precursor population dynamics in subcritical and critical reactors.

  3. Gaseous fuel reactors for power systems

    Science.gov (United States)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  4. Nuclear reactor machine refuelling system

    International Nuclear Information System (INIS)

    Cashen, W.S.; Erwin, D.

    1977-01-01

    Part of an on-line fuelling machine for a CANDU pressure-tube reactor is described. The present invention provides a refuelling machine wherein the fuelling components, including the fuel carrier and the closure adapter, are positively positioned and retained within the machine magazine or positively secured to the machine charge tube head, and cannot be accidentally disengaged as in former practice. The positive positioning devices include an arcuate keeper plate. Simplified hooked fingers are used. (NDH)

  5. Reactor vessel stud closure system

    International Nuclear Information System (INIS)

    Spiegelman, S.R.; Salton, R.B.; Beer, R.W.; Malandra, L.J.; Cognevich, M.L.

    1982-01-01

    A quick-acting stud tensioner apparatus for enabling the loosening or tightening of a stud nut on a reactor vessel stud. The apparatus is adapted to engage the vessel stud by closing a gripper around an upper end of the vessel stud when the apparatus is seated on the stud. Upon lifting the apparatus, the gripper releases the vessel stud so that the apparatus can be removed

  6. Nuclear reactor system for ABWR

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Kitagawa, Koji

    1997-01-01

    Various tests and measurements were performed during the pre-operational test run of Unit No. 6 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station, the first advanced boiling water reactor (ABWR) unit in the world, and the design and performance adequacy of the ABWR were confirmed. The realization of the ABWR in Japan took about 20 years. It was decided that technologies for the reactor internal pump (RIP) and the fine-motion control rod drive (FMCRD), which had been applied in Europe, would be incorporated in the ABWR aiming at simplification of its structure and operation. These main components were evaluated, modified and verified in consideration of the unique Japanese environment, such as seismic conditions, through a joint study program with Japanese utilities as well as an improvement and standardization program in cooperation with the government. In addition to incorporating RIP and FMCRD technologies, the ABWR also has improved features in terms of the design of the reactor pressure vessel and internals, as well as automated servicing equipment for the RIP, FMCRD, and primary containment vessel. (author)

  7. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

    1981-01-01

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  8. Renewal of reactor cooling system of JMTR. Reactor building site

    International Nuclear Information System (INIS)

    Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

    2012-03-01

    The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

  9. Light water reactor safeguards system evaluation

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

    1978-01-01

    A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

  10. Choice of thermal reactor systems: a report

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    This is a report by the UK National Nuclear Corporation published by the UK Secretary of State for Energy (Mr. Benn) on 29th July 1977. It is concerned with the advantages and disadvantages of three thermal reactor systems -the AGR (advanced gas cooled reactor), the PWR (pressurised water reactor), and the SGHWR (steam generating heavy water reactor). The object was to help in the future choice of a thermal system for the UK to cover the next 25 years. The matter of export potential is also considered. A programme of four stations of 1100 to 1300 MW each over six years starting from 1979 was assumed. It is emphasised that a decision must be taken now both about reactor systems and actual orders. Headings are as follows: Extract from conclusions reached; Summary of main features of assessment; General conclusions regarding the following - safety, security of the investment, operational characteristics, development and launching requirements, effect on industry, and capital and generation costs. It is stated that in order to make an overall judgement on reactor choice the technical, commercial and social issues involved must be weighed in conjunction with cost differentials.

  11. Neutron flux measuring system for nuclear reactor

    International Nuclear Information System (INIS)

    Aoki, Kazuo.

    1977-01-01

    Purpose: To avoid the generation of an undesired scram signal due to abrupt changes in the neutron level given to the detectors disposed near the boundary between the moderator and the atmosphere. Constitution: In a nuclear reactor adapted to conduct power control by the change of the level in the moderator such as heavy water, the outputs from the neutron detectors disposed vertically are averaged and the nuclear reactor is scramed corresponding to the averaged value. In this system, moderator level detectors are additionally provided to the nuclear reactor and their outputs, moderator level signal, are sent to a power averaging device where the output signals of the neutron detectors are judged if they are delivered from neutrons in the moderator or not depending on the magnitude of the level signal and the outputs of the detectors out of the moderator are substantially excluded. The reactor interlock signal from the device is utilized as a scram signal. (Seki, T.)

  12. Modelling aerosol behavior in reactor cooling systems

    International Nuclear Information System (INIS)

    McDonald, B.H.

    1990-01-01

    This paper presents an overview of some of the areas of concern in using computer codes to model fission-product aerosol behavior in the reactor cooling system (RCS) of a water-cooled nuclear reactor during a loss-of-coolant accident. The basic physical processes that require modelling include: fission product release and aerosol formation in the reactor core, aerosol transport and deposition in the reactor core and throughout the rest of the RCS, and the interaction between aerosol transport processes and the thermalhydraulics. In addition to these basic physical processes, chemical reactions can have a large influence on the nature of the aerosol and its behavior in the RCS. The focus is on the physics and the implications of numerical methods used in the computer codes to model aerosol behavior in the RCS

  13. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    Eisawy, E.A.; Sallam, H.

    2012-01-01

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  14. Considerations on nuclear reactor passive safety systems

    International Nuclear Information System (INIS)

    2016-01-01

    After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

  15. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  16. 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

  17. State system experience with safeguarding power reactors

    International Nuclear Information System (INIS)

    Roehnsch, W.

    1982-01-01

    This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

  18. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    Kovacik, W.P.

    1977-01-01

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

  19. Management system requirements for small reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.A., E-mail: kenneth.jones@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2013-07-01

    This abstract identifies the management system requirements for the life cycle of small reactors from initial conception through completion of decommissioning. For small reactors, the requirements for management systems remain the same as those for 'large' reactors regardless of the licensee' business model and objectives. The CSA N-Series of standards provides an interlinked set of requirements for the management of nuclear facilities. CSA N286 provides overall direction to management to develop and implement sound management practices and controls, while other CSA nuclear standards provide technical requirements and guidance that support the management system. CSA N286 is based on a set of principles. The principles are then supported by generic requirements that are applicable to the life cycle of nuclear facilities. CNSC regulatory documents provide further technical requirements and guidance. (author)

  20. Subcriticality determination of nuclear fuel assembly by Mihalczo method

    International Nuclear Information System (INIS)

    Yamane, Yoshihiro; Watanabe, Shoji; Nishina, Kojiro; Miyoshi, Yoshinori; Suzaki, Takenori; Kobayashi, Iwao.

    1986-01-01

    To establish a technique of on-site subcriticality determination suitable for the criticality safety management of nuclear fuel assembly, the applicability of the method proposed by Mihalczo was examined with the Tank-type Critical Assembly (TCA) of the Japan Atomic Energy Research Institute. In the Mihalczo method, cross power spectral densities and auto power spectral densities are evaluated from the output currents of an ionization chamber containing 252 Cf neutron source and two neutron detectors. The principle of this method is that the spectral ratio formed by the power spectral densities mentioned can be related to the subcriticality by the help of a stochastic theory. Throughout our data processing, an improved formula taking account of the neutron extinction at a detection process was used. Up to the subcriticality of 15 dollars, the Mihalczo method agreed with the water-level worth method, which has been a standard method of reactivity determination at the TCA facility. The systems treated in the present report hold symmetry concerning the nuclear fuel configuration and the 252 Cf chamber position. It was clarified that, contrary to Mihalczo's assertion, the factor converting the spectral ratio to a subcriticality depends on subcriticality itself. (author)

  1. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  2. The computerized reactor period measurement system for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

    The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

  3. Computer measurement system of reactor period for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

    The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

  4. Development study on subcriticality monitor. 1. Report under business contract with Japan Nuclear Fuel Cycle Development Institute

    CERN Document Server

    Yamada, S

    2002-01-01

    In this trust fund, we reviewed subcriticality measuring methods and neutron or gamma ray measuring and date transmission systems appropriate for realizing inexpensive on-line criticality surveillance systems, which is required for ensuring the safety of nuclear fuel reprocessing plants. Since the neutron flux level in subcritical systems is fairly low without external neutron sources, it is desirable to use pulse type neutron detectors for subcritical measurement systems. This logically implies that subcriticality measurement methods based on the temporal domain should be used for developing an on-line criticality surveillance system. In the deep subcriticality conditions, a strong external neutron source is needed for eactivity measurement and a D-T tube can be used in order to improve the accuracy of the measurement. A D-T tube is convenient since it is free from Tritium problem since Tritium is sealed in an airtight container and also can be controlled by power supply. Hence, under deep subcritical condit...

  5. Power conditioning system for a nuclear reactor

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi; Joge, Toshio.

    1981-01-01

    Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

  6. Digital instrumentation system for nuclear research reactors

    International Nuclear Information System (INIS)

    Aghina, Mauricio A.C.; Carvalho, Paulo Vitor R.

    2002-01-01

    This work describes a proposal for a system of nuclear instrumentation and safety totally digital for the Argonauta Reactor. The system divides in the subsystems: channel of pulses, channel of current, conventional instrumentation and safety system. The connection of the subsystems is made through redundant double local net, using the protocol modbus/rtu. So much the channel of pulses, the current channel and safety's system use modules operating in triple redundancy. (author)

  7. The IAEA power reactor information system - PRIS

    International Nuclear Information System (INIS)

    Laue, H.J.; Qureshi, A.; Skjoeldebrand, R.; White, D.

    1983-01-01

    The IAEA Power Reactor Information System, PRIS, is based on a collection of basic design data and operating experience data which the IAEA started in 1970. PRIS is used for annual publications on 'Power Reactors in Member States', 'Operating Experience with Nuclear Power Stations in Member States', which gives annual operating information for individual plants, and a 'Performance Analysis Report' summarizing each year's and earlier experience. Since 1973 information has been collected in a systematic manner on significant plant outages (= more than 10 full power hours). There is now information on more than 10,000 outages in the system which permits some conclusions to be drawn both in regard to individual plants and to categories of plants on the significance of different outage reasons and different types of equipment failures. PRIS has not been intended to be a component reliability information system as an international data collection must stop short of the level of detail which would be needed for that purpose. The objectives of PRIS have been to provide a factual background for assumptions on parameters which are essential for economic evaluations and for systems operation planning (load factor and availability). The outage information does, however, lend itself to conclusions about generic problems in different categories of plants and it can be used by an individual operator to find other plants where information about particular problems can be obtained. It would also now be possible to use PRIS for setting availability goals based on experience and not only on theoretical design considerations. The paper demonstrates the conclusions which can be drawn from 662 reactor years of operation of light and heavy water pressurized reactors and 390 reactor years of boiling water reactors and, in particular, the role that the main heat removal system and its components have played in the equipment failure category

  8. 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.)

  9. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  10. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  11. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  12. Criticality Analysis of SAMOP Subcritical Assembly

    International Nuclear Information System (INIS)

    Tegas-Sutondo; Syarip; Triwulan-Tjiptono

    2005-01-01

    A critically analysis has been performed for homogenous system of uranyl nitrate solution, as part of a preliminary design assessment on neutronic aspect of SAMOP sub-critical assembly. The analysis is intended to determine some critical parameters such as the minimum of critical dimension and critical mass for the desired concentration. As the basis of this analysis, it has been defined a fuel system with an enrichment of 20% for cylindrical geometry of both bare and graphite reflected of 30 cm thickness. The MCNP code has been utilized for this purpose, for variation of concentrations ranging from 150 g/l to 500 g/l. It is found that the best concentration giving the minimum geometrical dimension is around 400 g/l, for both the bare and reflected systems. Whilst the best one, of minimum critical mass is corresponding to the concentration of around 200 g/l with critical mass around 14.1 kg and 4.2 kg for the bare and reflected systems respectively. Based on the result of calculations, it is concluded that by taking into consideration of the critical limit, the SAMOP subcritical assembly is neutronically can be made. (author)

  13. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  14. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  15. CRNL research reactor retrofit Emergency Filtration System

    International Nuclear Information System (INIS)

    Philippi, H.M.

    1990-01-01

    This paper presents a brief history of NRX and NRU research reactor effluent air treatment systems before describing the selection and design of an appropriate retrofit Emergency Filtration System (EFS) to serve these reactors and the future MX-10 isotope production reactor. The conceptual design of the EFS began in 1984. A standby concrete shielding filter-adsorber system, sized to serve the reactor with the largest exhaust flow, was selected. The standby system, bypassed under normal operating conditions, is equipped with normal exhaust stream shutoff and diversion valves to be activated manually when an emergency is anticipated, or automatically when emergency levels of gamma radiation are detected in the exhaust stream. The first phase of the EFS installation, that is the construction of the EFS and the connection of NRU to the system, was completed in 1987. The second phase of construction, which includes the connection of NRX and provisions for the future connection of MX-10, is to be completed in 1990

  16. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

    2002-01-01

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6 / ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  17. Advanced reactor systems: safety and regulatory aspects

    International Nuclear Information System (INIS)

    Gopalakrishnan, A.

    1994-01-01

    Safety features which are desirable in futuristic reactor systems have been the subject of several studies over the past decade by different expert groups. When one discusses this subject, therefore, in a somewhat non-specific and qualitative manner, it is best to make use of the already available collective wisdom and literature on the matter. (author). 3 refs

  18. Coolant cleanup system for a nuclear reactor

    International Nuclear Information System (INIS)

    Shiina, Atsushi; Usui, Naoshi; Yamamoto, Michiyoshi; Osumi, Katsumi.

    1983-01-01

    Purpose: To maintain the electric conductivity of reactor water lower and to minimize the heat loss in the cleanup system by providing a low temperature cleanup system and a high temperature cleanup system together. Constitution: A low temperature cleanup system using ion exchange resins as filter aids and a high temperature cleanup system using inorganic ion exchange materials as filter aids are provided in combination. A part of the reactor water in a reactor pressure vessel is passed through a conductivity meter, one portion of which flows into the high temperature cleanup system having no heat exchanger and filled with inorganic ion exchange materials by way of a first flow rate control valve and the other portion of which flows into the low temperature cleanup system having heat exchangers and filled with the ion exchange materials by way of a second control valve. The first control valve is adjusted so as to flow, for example, about more than 15% of the feedwater flow rate to the high temperature cleanup system and the second control valve is adjusted with its valve opening degree depending on the indication of the conductivity meter so as to flow about 2 - 7 % of the feedwater flow rate into the low temperature cleanup system, to thereby control the electric conductivity to between 0.055 - 0.3 μS/cm. (Moriyama, K.)

  19. Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

    2013-01-01

    Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively

  20. Nuclear reactor plants and control systems therefor

    International Nuclear Information System (INIS)

    de Boer, G.A.; de Hex, M.

    1976-01-01

    A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

  1. Rodded shutdown system for a nuclear reactor

    International Nuclear Information System (INIS)

    Golden, M.P.; Govi, A.R.

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature is described. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core

  2. Method of controlling ECCS system in reactors

    International Nuclear Information System (INIS)

    Oohashi, Hideaki; Ikehara, Morihiko.

    1982-01-01

    Purpose: To eliminate the risk of misoperation and thereby improve the reliability of ECCS system upon accident. Method: ECCS system for nuclear reactor is automatically started by either of signals from a water level detector in a pressure vessel or from a pressure detector in a reactor container. Further, the ECCS system is started or stopped by the manual operation irrespective of the signals, and the signals from the pressure detector are isolated from the ECCS-starting signal by the contacts which actuate interlocked with the stopping operation of the manual operation switch. Then, after stopping the ECCS system by the manual operation, the ECCS system is started by the signals from the water level detector irrespective of the signals from the pressure detector. (Seki, T.)

  3. Reactor alarm system development and application issues

    Energy Technology Data Exchange (ETDEWEB)

    Drexler, J E; Oicese, G O [INVAP S.E. (Argentina)

    1997-09-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ``Reactor Alarm System`` (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: (a) Alarm Panel Display; (b) Alarm Page; (c) Alarm Masking and Inhibition; (d) Alarms Color and Attributes; (e) Condition Classification; and (f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: (a) Multiple-copy alarm summaries; (b) Improved alarm handling; (c) Extended dictionary; and (d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs.

  4. Reactor alarm system development and application issues

    International Nuclear Information System (INIS)

    Drexler, J.E.; Oicese, G.O.

    1997-01-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

  5. Design and test of a superconducting magnet in a linear accelerator for an Accelerator Driven Subcritical System

    International Nuclear Information System (INIS)

    Peng, Quanling; Xu, Fengyu; Wang, Ting; Yang, Xiangchen; Chen, Anbin; Wei, Xiaotao; Gao, Yao; Hou, Zhenhua; Wang, Bing; Chen, Yuan; Chen, Haoshu

    2014-01-01

    A batch superconducting solenoid magnet for the ADS proton linear accelerator has been designed, fabricated, and tested in a vertical dewar in Sept. 2013. A total of ten superconducting magnets will be installed into two separate cryomodules. Each cryomodule contains six superconducting spoke RF cavities for beam acceleration and five solenoid magnets for beam focusing. The multifunction superconducting magnet contains a solenoid for beam focusing and two correctors for orbit correction. The design current for the solenoid magnet is 182 A. A quench performance test shows that the operating current of the solenoid magnet can reach above 300 A after natural quenching on three occasions during current ramping (260 A, 268 A, 308 A). The integrated field strength and leakage field at the nearby superconducting spoke cavities all meet the design requirements. The vertical test checked the reliability of the test dewar and the quench detection system. This paper presents the physical and mechanical design of the batch magnets, the quench detection technique, field measurements, and a discussion of the residual field resulting from persistent current effects

  6. Subcritical to supercritical flow transition in a horizontal stratified flow

    International Nuclear Information System (INIS)

    Asaka, H.; Kukita, Y.

    1995-01-01

    The conditions for a transition from hydraulically subcritical to supercritical flow in the hot legs of a pressurized water reactor (PWR) were studied using data obtained from a two-phase natural circulation experiment conducted at the ROSA-IV Large Scale Test Facility (LSTF). The LSTF is a 1/48 volumetrically-scaled simulator of a Westinghouse-type PWR. The conditions for the transition were compared with the theory of Gardner. While the model explains the trend in the experimental data, the quantitative agreement was not satisfactory. It was found that the conditions for the transition from the subcritical to supercritical flow were predicted well by introducing energy loss term into the theory. (author)

  7. Reactor power system deployment and startup

    International Nuclear Information System (INIS)

    Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

  8. Fault-tolerant reactor protection system

    International Nuclear Information System (INIS)

    Gaubatz, D.C.

    1997-01-01

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs

  9. Fuel transfer system for a nuclear reactor

    International Nuclear Information System (INIS)

    Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

    1977-01-01

    Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

  10. 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.)

  11. Distributed computer control system for reactor optimization

    International Nuclear Information System (INIS)

    Williams, A.H.

    1983-01-01

    At the Oldbury power station a prototype distributed computer control system has been installed. This system is designed to support research and development into improved reactor temperature control methods. This work will lead to the development and demonstration of new optimal control systems for improvement of plant efficiency and increase of generated output. The system can collect plant data from special test instrumentation connected to dedicated scanners and from the station's existing data processing system. The system can also, via distributed microprocessor-based interface units, make adjustments to the desired reactor channel gas exit temperatures. The existing control equipment will then adjust the height of control rods to maintain operation at these temperatures. The design of the distributed system is based on extensive experience with distributed systems for direct digital control, operator display and plant monitoring. The paper describes various aspects of this system, with particular emphasis on: (1) the hierarchal system structure; (2) the modular construction of the system to facilitate installation, commissioning and testing, and to reduce maintenance to module replacement; (3) the integration of the system into the station's existing data processing system; (4) distributed microprocessor-based interfaces to the reactor controls, with extensive security facilities implemented by hardware and software; (5) data transfer using point-to-point and bussed data links; (6) man-machine communication based on VDUs with computer input push-buttons and touch-sensitive screens; and (7) the use of a software system supporting a high-level engineer-orientated programming language, at all levels in the system, together with comprehensive data link management

  12. Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

    International Nuclear Information System (INIS)

    Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

    2013-01-01

    RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

  13. COMPRESS - a computerized reactor safety system

    International Nuclear Information System (INIS)

    Vegh, E.

    1986-01-01

    The computerized reactor safety system, called COMPRESS, provides the following services: scram initiation; safety interlockings; event recording. The paper describes the architecture of the system and deals with reliability problems. A self-testing unit checks permanently the correct operation of the independent decision units. Moreover the decision units are tested by short pulses whether they can initiate a scram. The self-testing is described in detail

  14. Advanced liquid metal reactor plant control system

    International Nuclear Information System (INIS)

    Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

    1993-01-01

    The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

  15. Dust removal system for fusion experimental reactors

    International Nuclear Information System (INIS)

    Onozuka, M.; Ueda, Y.; Takahashi, K.; Oda, Y.; Seki, Y.; Ueda, S.; Aoki, I.

    1995-01-01

    Development of a dust removal system using static electricity has been conducted. It is envisioned that the system can collect and transport dust under vacuum. In the system, the dust is charged by dielectric polarization and floated by an electrostatic attraction force that is generated by the DC electric field. The dust is then transported by the electric curtain formed by the three-phase AC electric field. Experimental investigation has been conducted to examine the characteristics of the system. Current research results indicate that the dust removal system using static electricity can be used for fusion experimental reactors

  16. Dust removal system for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M.; Ueda, Y.; Takahashi, K.; Oda, Y. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Seki, Y.; Ueda, S.; Aoki, I. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)

    1995-12-31

    Development of a dust removal system using static electricity has been conducted. It is envisioned that the system can collect and transport dust under vacuum. In the system, the dust is charged by dielectric polarization and floated by an electrostatic attraction force that is generated by the DC electric field. The dust is then transported by the electric curtain formed by the three-phase AC electric field. Experimental investigation has been conducted to examine the characteristics of the system. Current research results indicate that the dust removal system using static electricity can be used for fusion experimental reactors.

  17. Reactor shutdown back-up system

    International Nuclear Information System (INIS)

    Hirao, Seizo; Sakashita, Motoaki.

    1982-01-01

    Purpose: To prevent back flow of poison upon injection to a moderator recycling pipeway. Constitution: In a nuclear reactor comprising a moderator recycling system for recycling and cooling moderator through a control rod guide pipe and a rapid poison injection system for rapidly injecting a poison solution at high density into the moderator by way of the same control rod guide pipe as a reactor shutdown back-up system, a mechanism is provided for preventing the back flow of a poison solution at high density into the moderator recycling system upon rapid injection of poison. An orifice provided in the joining pipeway to the control rod guide pipe on the side of the moderator recycling system is utilized as the back flow preventing device for the poison solution and the diameter for the orifice is determined so as to provide a constant ratio between the pressure loss in the control rod guide pipe and the pressure loss in the moderator recycling system pipe line upon usual reactor operation. (Kawakami, Y.)

  18. On the safety performance of the advanced nuclear energy systems

    International Nuclear Information System (INIS)

    Li Shounan

    1999-01-01

    Some features on the safety performances of the Advanced Nuclear Energy Systems are discussed. The advantages and some peculiar problems on the safety of Advanced Nuclear Energy Systems with subcritical nuclear reactor driven by external neutron sources are also pointed out in comparison with conventional nuclear reactors

  19. Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

    Science.gov (United States)

    Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

    2017-07-01

    In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

  20. Molten-salt reactor information system

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

    1975-06-01

    The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

  1. Development of fluoride reprocessing technology for molten salt transmutation reactor systems in the Czech Republic

    International Nuclear Information System (INIS)

    Uhlir, J.; Hosnedl, P.; Matal, O.

    2000-01-01

    At present, the transmutation of spent nuclear fuel is considered a prospective alternative conception with respect to the current conception based on the non-reprocessed spent fuel disposal into a deep geological repository. The Czech research and development programme in the area of partitioning is directed primarily on the development of the fuel cycle technology for the accelerator - driven subcritical reactor with a liquid fuel based on fluoride melts. The final objective of the research programme is the development of pyrochemical technologies suitable for a continuous or semi-continuous separation process which would allow practically perfect utilization of the transmutation potentialities of the reactor system. The present research is directed particularly on the development of suitable fluoride separation methods the target of which is the removal of the uranium component from spent nuclear fuel and on the research of the electro-separation procedures and further on the development of appropriate construction materials and equipment for the technology of fluoride salt melts. (authors)

  2. Reliability analysis of reactor protection systems

    International Nuclear Information System (INIS)

    Alsan, S.

    1976-07-01

    A theoretical mathematical study of reliability is presented and the concepts subsequently defined applied to the study of nuclear reactor safety systems. The theory is applied to investigations of the operational reliability of the Siloe reactor from the point of view of rod drop. A statistical study conducted between 1964 and 1971 demonstrated that most rod drop incidents arose from circumstances associated with experimental equipment (new set-ups). The reliability of the most suitable safety system for some recently developed experimental equipment is discussed. Calculations indicate that if all experimental equipment were equipped with these new systems, only 1.75 rod drop accidents would be expected to occur per year on average. It is suggested that all experimental equipment should be equipped with these new safety systems and tested every 21 days. The reliability of the new safety system currently being studied for the Siloe reactor was also investigated. The following results were obtained: definite failures must be detected immediately as a result of the disturbances produced; the repair time must not exceed a few hours; the equipment must be tested every week. Under such conditions, the rate of accidental rod drops is about 0.013 on average per year. The level of nondefinite failures is less than 10 -6 per hour and the level of nonprotection 1 hour per year. (author)

  3. Reactor Design for Bioelectrochemical Systems

    KAUST Repository

    Mohanakrishna, G.

    2017-12-01

    Bioelectrochemical systems (BES) are novel hybrid systems which are designed to generate renewable energy from the low cost substrate in a sustainable way. Microbial fuel cells (MFCs) are the well studied application of BES systems that generate electricity from the wide variety of organic components and wastewaters. MFC mechanism deals with the microbial oxidation of organic molecules for the production of electrons and protons. The MFC design helps to build the electrochemical gradient on anode and cathode which leads for the bioelectricity generation. As whole reactions of MFCs happen at mild environmental and operating conditions and using waste organics as the substrate, it is defined as the sustainable and alternative option for global energy needs and attracted worldwide researchers into this research area. Apart from MFC, BES has other applications such as microbial electrolysis cells (MECs) for biohydrogen production, microbial desalinations cells (MDCs) for water desalination, and microbial electrosynthesis cells (MEC) for value added products formation. All these applications are designed to perform efficiently under mild operational conditions. Specific strains of bacteria or specifically enriched microbial consortia are acting as the biocatalyst for the oxidation and reduction of BES. Detailed function of the biocatalyst has been discussed in the other chapters of this book.

  4. Reactor Design for Bioelectrochemical Systems

    KAUST Repository

    Mohanakrishna, G.; Kalathil, Shafeer; Pant, Deepak

    2017-01-01

    Bioelectrochemical systems (BES) are novel hybrid systems which are designed to generate renewable energy from the low cost substrate in a sustainable way. Microbial fuel cells (MFCs) are the well studied application of BES systems that generate electricity from the wide variety of organic components and wastewaters. MFC mechanism deals with the microbial oxidation of organic molecules for the production of electrons and protons. The MFC design helps to build the electrochemical gradient on anode and cathode which leads for the bioelectricity generation. As whole reactions of MFCs happen at mild environmental and operating conditions and using waste organics as the substrate, it is defined as the sustainable and alternative option for global energy needs and attracted worldwide researchers into this research area. Apart from MFC, BES has other applications such as microbial electrolysis cells (MECs) for biohydrogen production, microbial desalinations cells (MDCs) for water desalination, and microbial electrosynthesis cells (MEC) for value added products formation. All these applications are designed to perform efficiently under mild operational conditions. Specific strains of bacteria or specifically enriched microbial consortia are acting as the biocatalyst for the oxidation and reduction of BES. Detailed function of the biocatalyst has been discussed in the other chapters of this book.

  5. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  6. power system reliability in supplying nuclear reactors

    International Nuclear Information System (INIS)

    Gad, M.M.M.

    2007-01-01

    this thesis presents a simple technique for deducing minimal cut set (MCS) from the defined minimal path set (MPS) of generic distribution system and this technique have been used to evaluate the basic reliability indices of Egypt's second research reactor (ETRR-2) electrical distribution network. the alternative system configurations are then studied to evaluate their impact on service reliability. the proposed MCS approach considers both sustained and temporary outage. the temporary outage constitutes an important parameter in characterizing the system reliability indices for critical load point in distribution system. it is also consider the power quality impact on the reliability indices

  7. Emergency cooling system for the PHENIX reactor

    International Nuclear Information System (INIS)

    Megy, J.M.; Giudicelli, A.G.; Robert, E.A.; Crette, J.P.

    Among various engineered safeguards of the reactor plant, the authors describe the protective system designed to remove the decay heat in emergency, in case of complete loss of all normal decay heat removal systems. First the normal decay heat rejection systems are presented. Incidents leading to the loss of these normal means are then analyzed. The protective system and its constructive characteristics designed for emergency cooling and based on two independent and highly reliable circuits entirely installed outside the primary containment vessel are described

  8. Integral reactor system and method for fuel cells

    Science.gov (United States)

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  9. REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

  10. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  11. RIMACS, Reactor Inspection Main Control System

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: RIMACS prepares for automatic inspection files on each inspection item for the reactor. These automatic inspection files provide the data to move RIROB (Reactor Inspection Robot) with laser by interpreting the coordinates of LASPO (Laser Positioner) and the laser detecting device of RIROB in three dimensional space. In addition, when RIROB arrives at the inspecting location, the files provide all values of the manipulator's motions to acquire the ultrasonic data. RIMACS provides various modules in order to perform these complex functions, and the functions are programmed on graphic user interface for the convenience of the user. RIMACS provides various functions, such as insertion of reactor production data, selection of the reactor for inspection, the creation of automatic inspection file, the selection of the inspection item, inspection simulation, and automatic inspection procedures. It also provides all other functions, which are necessary for the inspection, such as operating program download and manual control of LASPO and RIROB, the inspection simulation and the inspection status display by means of the graphic screen, and SODAS (ultra-Sonic Data Acquisition System) drive verification. 2 - Methods: Moving path and operation procedures for inspection robot are generated automatically with Kinematics algorithm. 3 - Restrictions on the complexity of the problem: A graphics display with MS-Window capability is required

  12. Overheating preventive system for reactor core fuels

    International Nuclear Information System (INIS)

    Ito, Daiju

    1981-01-01

    Purpose: To ensure the cooling function of reactor water in a cooling system in case of erroneous indication or misoperation by reliable temperature measurement for fuels and actuating relays through the conversion output obtained therefrom. Constitution: Thermometers are disposed laterally and vertically in a reactor core in contact with core fuels so as to correspond to the change of status in the reactor core. When there is a high temperature signal issued from one of the thermometers or one of conversion circuits, the function of relay contacts does not provide the closed state as a whole. When high temperature signals are issued from two or more thermometers of conversion circuits from independent OR circuits, the function of the relay contacts provides the closure state as a whole. Consequently, in the use of 2-out of 3-circuits, the entire closure state, that is, the misoperation of the relay contacts for the thermometer or the conversion circuits can be avoided. In this way, by the application of the output from the conversion circuits to the logic circuit and, in turn, application of the output therefrom to the relay groups in 2-out of 3-constitution, the reactor safety can be improved. (Horiuchi, T.)

  13. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

    2008-01-01

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

  14. Microchannel Reactor System for Catalytic Hydrogenation

    Energy Technology Data Exchange (ETDEWEB)

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  15. Staged membrane oxidation reactor system

    Science.gov (United States)

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2012-09-11

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  16. Cost benefit analysis of reactor safety systems

    International Nuclear Information System (INIS)

    Maurer, H.A.

    1984-01-01

    Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

  17. Transesterification of soybean oil with methanol and acetic acid at lower reaction severity under subcritical conditions

    International Nuclear Information System (INIS)

    Go, Alchris Woo; Sutanto, Sylviana; NguyenThi, Bich Thuyen; Cabatingan, Luis K.; Ismadji, Suryadi; Ju, Yi-Hsu

    2014-01-01

    Highlights: • (trans)Esterification of oils under subcritical conditions. • Acetic acid as catalyst and co-solvent in biodiesel production. • Influence of reactor hydrodynamic (loading and stirring) on FAME yield. • High methyl ester yield can be obtained at less severe reaction conditions. - Abstract: Soybean oil (56–80 g) was reacted with methanol (40–106 mL) to produce fatty acid methyl ester in the presence of 1–6% acetic acid under subcritical condition at 250 °C. Stirring and loading of the reaction system affected the yield and severity of the process. The presence of acetic acid improved the yield of FAME from 32.1% to 89.5% at a methanol to oil molar ratio of 20 mL/g. Acetic acid was found to act strongly as an acid catalyst and to some extent improved the solubility between oil and methanol. Reaction pressure higher than the supercritical pressure of methanol (7.85 MPa) was not required to achieve high FAME yield (89.5–94.8%) in short time (30–60 min)

  18. Laser fusion hybrid reactor systems study

    International Nuclear Information System (INIS)

    1976-07-01

    The work was performed in three phases. The first phase included a review of the many possible laser-reactor-blanket combinations and resulted in the selection of a ''demonstration size'' 500 MWe plant for further study. A number of fast fission blankets using uranium metal, uranium-molybdenum alloy, and uranium carbide as fuel were investigated. The second phase included design of the reactor vessel and internals, heat transfer system, tritium processing system, and the balance of plant, excluding the laser building and equipment. A fuel management scheme was developed, safety considerations were reviewed, and capital and operating costs were estimated. Costs developed during the second phase were unexpectedly high, and a thorough review indicated considerable unit cost savings could be obtained by scaling the plant to a larger size. Accordingly, a third phase was added to the original scope, encompassing the redesign and scaling of the plant from 500 MWe to 1200 MWe

  19. Power conditioning for space nuclear reactor systems

    Science.gov (United States)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  20. Space reactor electric systems: system integration studies, Phase 1 report

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-01-01

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

  1. Development of intellectual reactor design system IRDS

    International Nuclear Information System (INIS)

    Kugo, T.; Tsuchihashi, K.; Nakagawa, M.; Mori, T.

    1993-01-01

    An intellectual reactor design system IRDS has been developed to support feasibility study and conceptual design of new type reactors in the fields of reactor core design including neutronics, thermal-hydraulics and fuel design. IRDS is an integrated software system in which a variety of computer codes in the different fields are installed. An integration of simulation modules are performed by the information transfer between modules through design model in which the design information of the current design work is stored. An object oriented architecture is realized in frame representation of core configuration in a design data base. The knowledge relating to design tasks to be performed are encapsulated, to support the conceptual design work. The system is constructed on an engineering workstation, and supports efficiently design work through man-machine interface adopting the advanced information processing technologies. Optimization methods for design parameters with use of the artificial intelligence technique are now under study, to reduce the parametric study work. A function to search design window in which design is feasible is realized in the fuel pin design. (orig.)

  2. Reactor component inventory system at FFTF

    International Nuclear Information System (INIS)

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER

  3. Data acquisition system for nuclear reactor environment

    International Nuclear Information System (INIS)

    Tiwari, Akash; Tiwari, Railesha; Tiwari, S.S.; Panday, Lokesh; Suri, Nitin; Chouksey, Abhsek; Singh, Sarvesh Kumar; Dwivedi, Tarun; Agrawal, Ashish; Pandey, Pranav Kumar; Sharma, Brijnandan; Bhatia, Chirag

    2004-01-01

    We have designed an online real time data acquisition system for nuclear reactor environment monitoring. Data acquisition system has eight channels of analog signals and one channel of pulsed input signal from detectors like GM Tube, or any other similar input. Connectivity between the data acquisition system and environmental parameters monitoring computer is made through a wireless data communication link of 151 MHz/100 mW RF power and 10 km maximum communication range for remote data telemetry. Sensors used are gamma ionizing radiation sensor made from CsI:Tl scintillator, atmospheric pressure sensor with +/-0.1 mbar precision, temperature sensor with +/-l milli degree Celsius precision, relative humidity with +/-0.1RH precision, pulse counts with +/-1 count in 0-10000 Hz count rate measurement precision and +/-1 count is accumulated count measurement precision. The entire data acquisition system and wireless telemetry system is 9 V battery powered and the device is to be fitted on a wireless controlled mobile robot for scanning the nuclear reactor zone from remote. Wireless video camera has been planned for integration into the existing system on a later date for moving the robotics environmental data acquisition system beyond human vision reach. System development cost is Rs.25 Lacs and has been developed for Department of Atomic Energy, Government of India and Indian Defense use. (author)

  4. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  5. Dosimetry system of the RB reactor; Dozimetarski sistem reaktora RB

    Energy Technology Data Exchange (ETDEWEB)

    Lolic, B; Vukadin, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1962-07-01

    Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor.

  6. Heat exchangers in heavy water reactor systems

    International Nuclear Information System (INIS)

    Mehta, S.K.

    1988-01-01

    Important features of some major heat exchange components of pressurized heavy water reactors and DHRUVA research reactor are presented. Design considerations and nuclear service classifications are discussed

  7. Primary cooling system for BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

    1993-01-01

    The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

  8. Monitoring system in reactor dry well

    International Nuclear Information System (INIS)

    Horie, Akira; Suzuki, Shun-ichi; Yamamoto, Shinji; Kubokawa, Toshihiko; Takagi, Sakae; Yokosawa, Makoto.

    1991-01-01

    A failed portion of a dry well in a BWR type reactor is monitored and identified from a remote place by a simple structure. That is, laser beams are irradiated under scanning to a portion to be monitored. Then, the reflection light is monitored by a light receiving and monitoring system, and abnormalities such as defects or leaks of monitored portion are optically detected by a remote viewing equipment. With such a constitution, the portion to be monitored in poor operation circumstances of the reactor dry well can always be monitored efficiently from a remote place. The device of the present invention does not undergo the effect of radiation noises, etc. and it is excellent in heat resistance and radiation resistance. (I.S.)

  9. A stereoscopic television system for reactor inspection

    International Nuclear Information System (INIS)

    Friend, D.B.; Jones, A.

    1980-03-01

    A stereoscopic television system suitable for reactor inspection has been developed. Right and left eye views, obtained from two conventional black and white cameras, are displayed by the anaglyph technique and observers wear appropriately coloured viewing spectacles. All camera functions, such as zoom, focus and toe-in are remotely controlled. A laboratory experiment is described which demonstrates the increase in spatial awareness afforded by the use of stereo television and illustrates its potential in the supervision of remote handling tasks. Typical depth resolutions of 3mm at 1m and 10mm at 2m have been achieved with the reactor instrument. Trials undertaken during routine inspection at Oldbury Power Station in June 1978 are described. They demonstrate that stereoscopic television can indeed improve the convenience of remote handling and that the added display realism is beneficial in visual inspection. (author)

  10. Nuclear reactor insulation and preheat system

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    An insulation and preheat system is disclosed for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the ocmpartment. An external surface of the compartment of enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair

  11. Refueling system for a nuclear reactor

    International Nuclear Information System (INIS)

    Koschkin, J.N.; Ordynskij, G.V.; Schchijan, C.G.; Schapkin, A.F.; Fadeev, A.I.; Laptev, F.V.; Batjukov, V.I.; Korolkov, K.I.; Borodin, I.V.; Tschernomordik, E.N.

    1979-01-01

    With the refueling system fuel elements are transferred from the intermediate distributing chamber within the fast breeder reactor vessel to the storage tanks for new and irradiated fuel elements outside of the reactor vessel and vice versa. It consists of a hermetic chamber, filled with inert gas, within which the refueling machine, having got a vertical swing pipe, is placed. On the swing pipe there is mounted by means of a bracket a hanging support tube for a tube manipulator that can be moved over the openings to the fuel elements. At the end of the tube manipulator there is a gripping device whose drive mechanism is arranged within the support tube. The swing pipe is moved by means of a drive mechanism outside of the chamber. (DG) [de

  12. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    Adinolfi, R.; Noviello, L.

    1990-01-01

    The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

  13. The Liquid Annular Reactor System (LARS) propulsion

    International Nuclear Information System (INIS)

    Powell, J.; Ludewig, H.; Horn, F.; Lenard, R.

    1990-01-01

    A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5)

  14. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  15. Study of the kinetics parameters for subcritical media driven by source

    International Nuclear Information System (INIS)

    Lee, S.M.; Maiorino, J.R.

    2009-01-01

    This paper presents a theoretical study of reactor kinetics focusing on the methodology of calculation and the experimental measurements of the so-called kinetic parameters. A comparison between the methodology based on the Dulla's formalism and the classical method is made. The objective is to exhibit the dependence of the parameters on sub criticality level and perturbation. Two different slab type systems were considered: thermal one and fast one, both with homogeneous media. One group diffusion model was used for the fast reactor, and for the thermal system, two group diffusion model, considering, in both case only one precursor's family. For reason of simplicity, several additional assumptions were made for calculation of two group method: no up-scattering, fission reaction occurring only in thermal group, etc. The solutions for subcritical systems were obtained using the expansion method, and for critical systems, the methods presented in classical textbooks of reactor physics were applied. The numerical results presented their dependence on sub criticality level and perturbation. (author)

  16. Present status of Monte Carlo seminar for sub-criticality safety analysis in Japan

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi

    2003-01-01

    This paper provides overview of the methods and results of a series of sub-criticality safety analysis seminars for nuclear fuel cycle facility with the Monte Carlo method held in Japan from July 2000 to July 2003. In these seminars, MCNP-4C2 system (MS-DOS version) was installed in note-type personal computers for participants. Fundamental theory of reactor physics and Monte Carlo simulation as well as the contents of the MCNP manual were lectured. Effective neutron multiplication factors and neutron spectra were calculated for some examples such as JCO deposit tank, JNC uranium solution storage tank, JNC plutonium solution storage tank and JAERI TCA core. Management for safety of nuclear fuel cycle facilities was discussed in order to prevent criticality accidents in some of the seminars. (author)

  17. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

    Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

  18. Decision aid systems for nuclear reactors

    International Nuclear Information System (INIS)

    Evrard, J.M.; Martinez, J.M.

    1992-01-01

    The development of new techniques, especially in the field of artificial intelligence, makes it possible to design more powerful computerized systems, supporting tasks related to the design and operation of nuclear power plants. The potential contribution and perspectives for the integration of such systems depend upon whether the improvement of existing plants, the design of next generation reactors or future projects are concerned. We present four systems which show the state-of-the-art as regards knowledge-based systems. The first system is related to the automatic generation of procedures dealing with loss of electrical sources. The second one aims at assisting the power plant utility in following the technical specifications during maintenance operations. Finally, the last two are designed to help an emergency team evaluate and forecast the evolution of an accidental situation in a nuclear reactor. Perspectives for on-line operator assistance are then discussed, as well as the main technical themes which will make it possible to design such systems. We conclude with the difficulties which are encountered upon the integration of these tools: their validation and task sharing between man and machine

  19. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  20. Passive cooling systems in power reactors

    International Nuclear Information System (INIS)

    Aharon, J.; Harrari, R.; Weiss, Y.; Barnea, Y.; Katz, M.; Szanto, M.

    1996-01-01

    This paper reviews several R and D activities associated with the subject of passive cooling systems, conducted by the N.R.C.Negev thermohydraulic group. A short introduction considering different types of thermosyphons and their applications is followed by a detailed description of the experimental work, its results and conclusions. An ongoing research project is focused on the evaluation of the external dry air passive containment cooling system (PCCS) in the AP-600 (Westinghouse advanced pressurized water reactor). In this context some preliminary theoretical results and planned experimental research are for the fature described

  1. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  2. Introduction to nuclear power reactors and their health physics systems

    International Nuclear Information System (INIS)

    Brtis, J.S.

    1982-01-01

    This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

  3. Benchmarking criticality safety calculations with subcritical experiments

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1984-06-01

    Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments

  4. Analysis on burn-up behaviors for accelerator-driven sub-critical facility

    International Nuclear Information System (INIS)

    Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qinbiao; Ding Dazhao

    2000-01-01

    An analysis is performed on burn-up behaviors for accelerator-driven sub-critical reactor by means of the code PASC-1 for neutronics calculation, the code CBURN for burn-up calculation and 44 group constants is processed by CENDL-2 and ENDF/B-6 using NJOY-91.91

  5. The unique safety challenges of space reactor systems

    International Nuclear Information System (INIS)

    Lanes, S.J.; Marshall, A.C.

    1991-01-01

    Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

  6. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  7. Passive safety systems for integral reactors

    International Nuclear Information System (INIS)

    Kuul, V.S.; Samoilov, O.B.

    1996-01-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs

  8. Passive safety systems for integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuul, V S; Samoilov, O B [OKB Mechanical Engineering (Russian Federation)

    1996-12-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs.

  9. HYPER (hybrid power extraction reactor): a system for clean nuclear energy

    International Nuclear Information System (INIS)

    Park, W.S.; Shin, U.; Han, S.-J.; Song, T.Y.; Choi, B.H.; Park, C.K.

    2000-01-01

    The Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development (RID) called HYPER (hybrid power extraction reactor) for the transmutation of nuclear waste and energy production through the transmutation process. HYPER program is within the frame work of the national mid and long-term nuclear research plan. KAERI is aiming to develop the elemental technologies for the subcritical transmutation system by the year of 2001 and build a small bench scale test facility (∝5 MW) by the year of 2006. Some major features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead-bismuth (Pb-Bi) is adopted as a coolant and spallation target material. 1 GeV 16 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MW of power. The support ratio of HYPER for LWR units producing the same power is believed to be 5∝6. (orig.)

  10. ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski

    2014-01-01

    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  11. Advanced thermionic reactor systems design code

    International Nuclear Information System (INIS)

    Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

    1991-01-01

    An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

  12. Nuclear power reactors and hydrogen storage systems

    International Nuclear Information System (INIS)

    Ibrahim Aly Mahmoud El Osery.

    1980-01-01

    Among conclusions and results come by, a nuclear-electric-hydrogen integrated power system was suggested as a way to prevent the energy crisis. It was shown that the hydrogen power system using nuclear power as a leading energy resource would hold an advantage in the current international situation as well as for the long-term future. Results reported provide designers of integrated nuclear-electric-hydrogen systems with computation models and routines which will allow them to explore the optimal solution in coupling power reactors to hydrogen producing systems, taking into account the specific characters of hydrogen storage systems. The models were meant for average computers of a type easily available in developing countries. (author)

  13. Nuclear reactor cavity floor passive heat removal system

    Science.gov (United States)

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    2018-03-06

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

  14. Nuclear reactor system study for NASA/JPL

    Science.gov (United States)

    Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.

    1982-01-01

    Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

  15. Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

    International Nuclear Information System (INIS)

    1986-11-01

    This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

  16. Development of Vibration Diagnostic System in Research Reactors

    International Nuclear Information System (INIS)

    EL-Kafas, A. A.

    1999-01-01

    Early failure detection and diagnosis system are an important group with increasing interest with the operating support system. Already existing system to monitor integrity of primary system components are vibration and acoustic monitoring system (2,3). The development of vibration diagnostic system for MARIA reactor (30 MW)-the second research reactor in Poland -was made. The new system is applied for the Egypt research reactor (ETRR-1). This paper represents the result obtained during the operation of this activity that carried out at MARIA and ETRR-1 reactors

  17. Subcriticality determination in ADS: Valina-Booster experiments

    International Nuclear Information System (INIS)

    Persson, C. M.; Gudowski, W.; Fokau, A.; Bournos, V.; Fokov, Y.; Routkovskaia, C.; Serafimovich, I.; Kiyavitskaya, H.

    2007-01-01

    A major problem in operating a full-scale subcritical accelerator-driven system (ADS) is to ensure sufficient margin to criticality. Therefore, reliable techniques for subcriticality monitoring are required. In order to develop such techniques, a full understanding of existing reactivity determination methods is essential. In this work, reactivity determination methods, such as pulsed neutron source methods and noise methods, are studied experimentally in the subcritical facility YALINA-Booster. YALINA-Booster: The subcritical assembly YALINA-Booster: recently constructed at the Joint Institute for Power and Nuclear Research - Sosny, consists of a subcritical core driven by an external neutron source. The neutron source is a powerful neutron generator consisting of a deuteron accelerator and a target of deuterium or tritium embedded in titanium. Through (d, d) - or (d, t)-reactions neutrons are created with energy around 2.5 MeV and 14.1 MeV respectively. Neutrons are born in the centre of the core and multiply through a lead matrix fuelled with highly enriched uranium (90% and 36%). This zone is referred to as the booster zone and is surrounded by a thermal zone, moderated by polyethylene. In order to reach sufficient high effective multiplication factor, the thermal zone is fuelled by approximately one thousand rods of 10% enriched uranium dioxide in cylindrical geometry. To prevent thermal neutrons from diffusing into the fast booster zone, an interface, consisting of boron carbide and natural uranium rods, is located between the zones. YALINA-Booster has a radial graphite reflector of thickness 24 cm. Experiments: Experiments using the neutron source in pulsed mode will be presented, relying on methods such as the area method and the method of prompt neutron decay rate determination. Moreover, results from noise analysis using for instance the Feynman-α method will be presented

  18. Subcriticality monitoring method based on the exponential technique usable for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Suzaki, T.

    1987-01-01

    Buckling measurement methods in subcritical nuclear fuel systems (negative buckling measurements in small systems are well-known as the exponential experiment) were discussed from the viewpoint of the applicability to on-site monitorings of subcriticality and fuel characteristics of interest. From demonstration experiments using the TCA, it was revealed that the method is quite promising. Applicability of the method to the critical approach in critical assemblies was also discussed. (author)

  19. Cable handling system for use in a nuclear reactor

    International Nuclear Information System (INIS)

    Crosgrove, R.O.; Larson, E.M.; Moody, E.

    1982-01-01

    A cable handling system for use in an installation such as a nuclear reactor is disclosed herein along with relevant portions of the reactor which, in a preferred embodiment, is a liquid metal fast breeder reactor. The cable handling system provides a specific way of interconnecting certain internal reactor components with certain external components, through an assembly of rotatable plugs. Moreover, this is done without having to disconnect these components from one another during rotation of the plugs and yet without interfering with other reactor components in the vicinity of the rotating plugs and cable handling system

  20. Radiolytic production of chemical fuels in fusion reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fish, J D

    1977-06-01

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered.